Mechanical Properties of Hot Mix Crumb Rubber Modified Asphalt Concrete Using Waste Tire

Energy Technology Data Exchange (ETDEWEB)

Kim, Nak Seok; Lee, Woo Yeol [Kyonggi University, Suwon (Korea)

1998-06-30

Wheel tracking and ravelling tests were conducted on the hot mix crumb rubber modified asphalt concrete using waste tire to evaluate the mechanical properties in comparison with conventional asphalt concrete. According to the test results, the modified product was superior to the conventional one by 50% in the resistance of permanent deformation and by 15% in the resistance of durability. The experimental results should recommend that the waste tire is positively recycled for asphalt concrete. (author). 11 refs., 6 tabs., 2 figs.

Evaluation of Colemanite Waste as Aggregate Hot Mix Asphalt Concrete

Directory of Open Access Journals (Sweden)

Nihat MOROVA

2015-09-01

Full Text Available In this study usability of waste colemanite which is obtained after cutting block colemanite for giving proper shape to blocks as an aggregate in hot mix asphalt. For this aim asphalt concrete samples were prepared with four different aggregate groups and optimum bitumen content was determined. First of all only limestone was used as an aggregate. After that, only colemanite aggregate was used with same aggregate gradation. Then, the next step of the study, Marshall samples were produced by changing coarse and fine aggregate gradation as limestone and colemanite and Marshall test were conducted. When evaluated the results samples which produced with only limestone aggregate gave the maximum Marshall Stability value. When handled other mixture groups (Only colemanite, colemanite as coarse aggregate-limestone as fine aggregate, colemanite as fine aggregate-limestone as coarse aggregate all groups were verified specification limits. As a result, especially in areas where there is widespread colemanite waste, if transportation costs did not exceed the cost of limestone, colemanite stone waste could be used instead of limestone in asphalt concrete mixtures as fine aggregate

Fundamental Aspects of Zeolite Waste Form Production by Hot Isostatic Pressing

Energy Technology Data Exchange (ETDEWEB)

Jubin, Robert Thomas [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bruffey, Stephanie H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jordan, Jacob A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-02-01

The direct conversion of iodine-bearing sorbents into a stable waste form is a research topic of interest to the US Department of Energy. The removal of volatile radioactive 129I from the off-gas of a nuclear fuel reprocessing facility will be necessary in order to comply with the regulatory requirements that apply to facilities sited within the United States (Jubin et al., 2012a), and any iodine-containing media or solid sorbents generated by this process would contain ¹²⁹I and would be destined for eventual geological disposal. While recovery of iodine from some sorbents is possible, a method to directly convert iodineloaded sorbents to a durable waste form with little or no additional waste materials being formed and a potentially reduced volume would be beneficial. To this end, recent studies have investigated the conversion of iodine-loaded silver mordenite (I-AgZ) directly to a waste form by hot isostatic pressing (HIPing) (Bruffey and Jubin, 2015). Silver mordenite (AgZ), of the zeolite class of minerals, is under consideration for use in adsorbing iodine from nuclear reprocessing off-gas streams. Direct conversion of I-AgZ by HIPing may provide the following benefits: (1) a waste form of high density that is tolerant to high temperatures, (2) a waste form that is not significantly chemically hazardous, and (3) a robust conversion process that requires no pretreatment.

Use of a CO2 pellet non-destructive cleaning system to decontaminate radiological waste and equipment in shielded hot cells at the Bettis Atomic Power Laboratory

International Nuclear Information System (INIS)

Bench, T.R.

1997-01-01

This paper details how the Bettis Atomic Power Laboratory modified and utilized a commercially available, solid carbon dioxide (CO 2 ) pellet, non-destructive cleaning system to support the disposition and disposal of radioactive waste from shielded hot cells. Some waste materials and equipment accumulated in the shielded hot cells cannot be disposed directly because they are contaminated with transuranic materials (elements with atomic numbers greater than that of uranium) above waste disposal site regulatory limits. A commercially available CO 2 pellet non-destructive cleaning system was extensively modified for remote operation inside a shielded hot cell to remove the transuranic contaminants from the waste and equipment without generating any secondary waste in the process. The removed transuranic contaminants are simultaneously captured, consolidated, and retained for later disposal at a transuranic waste facility

The production of pig iron from crushing plant waste using hot blast cupola

Directory of Open Access Journals (Sweden)

Kusno Isnugroho

2018-03-01

Full Text Available A production of pig iron has been conducted from crushing plant waste. The process of preparing pig iron was using hot blast cupola (HBC furnace which was injected with charcoal powder to improve temperature process and reduction zone in the furnace. The process was started by washing process and magnetic separation of raw material as an effort to improve iron content degree from crushing plant waste. The next process was preparing the composite pellet with the particle size of −80 + 100 mesh and with the composition of 80% iron ore, 15% wood charcoal, and 5% bentonite. The result of pellet size was 2.5–4.0 mm. The experiment was continued to reduce pellet composite in the HBC furnace. The pig iron produced from this process contained of 93.62%Fe, 3.5%C, 1.55%Si, 0.87%Mn, 0.05%P, and 0.087%S.With this result, the pig iron produced already fulfill the metallurgical specification to be used in smelting industry. Keywords: Pig iron, Pellet, Injection, Charcoal, Hot blast cupola

Hot Cell Liners Category of Transuranic Waste Stored Below Ground within Area G

Energy Technology Data Exchange (ETDEWEB)

Jones, Robert Wesley [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Hargis, Kenneth Marshall [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2014-09-01

A large wildfire called the Las Conchas Fire burned large areas near Los Alamos National Laboratory (LANL) in 2011 and heightened public concern and news media attention over transuranic (TRU) waste stored at LANL’s Technical Area 54 (TA-54) Area G waste management facility. The removal of TRU waste from Area G had been placed at a lower priority in budget decisions for environmental cleanup at LANL because TRU waste removal is not included in the March 2005 Compliance Order on Consent (Reference 1) that is the primary regulatory driver for environmental cleanup at LANL. The Consent Order is an agreement between LANL and the New Mexico Environment Department (NMED) that contains specific requirements and schedules for cleaning up historical contamination at the LANL site. After the Las Conchas Fire, discussions were held by the U.S. Department of Energy (DOE) with the NMED on accelerating TRU waste removal from LANL and disposing it at the Waste Isolation Pilot Plant (WIPP). This report summarizes available information on the origin, configuration, and composition of the waste containers within the Hot Cell Liners category; their physical and radiological characteristics; the results of the radioassays; and the justification to reclassify the five containers as LLW rather than TRU waste.

Hot Spot Removal System: System description

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-09-01

Hazardous wastes contaminated with radionuclides, chemicals, and explosives exist across the Department of Energy complex and need to be remediated due to environmental concerns. Currently, an opportunity is being developed to dramatically reduce remediation costs and to assist in the acceleration of schedules associated with these wastes by deploying a Hot Spot Removal System. Removing the hot spot from the waste site will remove risk driver(s) and enable another, more cost effective process/option/remedial alternative (i.e., capping) to be applied to the remainder of the site. The Hot Spot Removal System consists of a suite of technologies that will be utilized to locate and remove source terms. Components of the system can also be used in a variety of other cleanup activities. This Hot Spot Removal System Description document presents technologies that were considered for possible inclusion in the Hot Spot Removal System, technologies made available to the Hot Spot Removal System, industrial interest in the Hot Spot Removal System`s subsystems, the schedule required for the Hot Spot Removal System, the evaluation of the relevant technologies, and the recommendations for equipment and technologies as stated in the Plan section.

Hot Spot Removal System: System description

International Nuclear Information System (INIS)

1997-09-01

Hazardous wastes contaminated with radionuclides, chemicals, and explosives exist across the Department of Energy complex and need to be remediated due to environmental concerns. Currently, an opportunity is being developed to dramatically reduce remediation costs and to assist in the acceleration of schedules associated with these wastes by deploying a Hot Spot Removal System. Removing the hot spot from the waste site will remove risk driver(s) and enable another, more cost effective process/option/remedial alternative (i.e., capping) to be applied to the remainder of the site. The Hot Spot Removal System consists of a suite of technologies that will be utilized to locate and remove source terms. Components of the system can also be used in a variety of other cleanup activities. This Hot Spot Removal System Description document presents technologies that were considered for possible inclusion in the Hot Spot Removal System, technologies made available to the Hot Spot Removal System, industrial interest in the Hot Spot Removal System''s subsystems, the schedule required for the Hot Spot Removal System, the evaluation of the relevant technologies, and the recommendations for equipment and technologies as stated in the Plan section

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

International Nuclear Information System (INIS)

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

2001-01-01

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

Mobile hot cell transition design phase study for radioactive waste treatment on the Hanford reservation site

International Nuclear Information System (INIS)

Pons, Y.

2010-01-01

Full text of publication follows: At the US Department of Energy's Hanford Reservation site, 4 caissons in under ground storage contain approximately 23 cubic meters of Transuranic (TRU) waste, in over 5,000 small packages. The retrieval of these wastes presents a number of very difficult issues, including the configuration of the vaults, approximately 50,000 curies of activity, high dose rates, and damaged/degraded waste packages. The waste will require remote retrieval and processing sufficient to produce certifiable RH-TRU waste packages. This RH-TRU will be packaged for staging on site until certification by CCP is completed to authorize shipment to the Waste Isolation Pilot Plant (WIPP). The project has introduced AREVA' s innovative Hot Mobile Cell (HMC) technology to perform size reduction, sorting, characterization, and packaging of the RH waste stream at the point of generation, the retrieval site in the field. This approach minimizes dose and hazard exposure to workers that is usually associated with this operation. The HMC can also be used to provide employee protection, weather protection, and capacity improvements similar to those realized in general burial ground. AREVA TA and his partner AFS will provide this technology based on the existing HMCs developed and operated in France: - ERFB (Bituminized Waste Drum Retrieval Facility): ERFB was built specifically for retrieving the bituminized waste drums (approximately 6,000 stored in trenches in the North zone on the Marcoule site (in operation since 2001). - ERCF (Waste Drum Recovery and Packaging Facility): The ERCF was built specifically to retrieve bituminized waste drums stored in 35 pits located in the south area on Marcoule site (in operation) - FOSSEA (Legacy Waste Removal and Trench Cleanup): The FOSSEA project consists of the retrieval of waste stored on the Basic Nuclear Facility. Waste from the 56 trenches will be inspected, characterised, and if necessary processed or repackaged, and

Hot dewatering and resin encapsulation of intermediate level radioactive waste

International Nuclear Information System (INIS)

Rickman, J.; Birch, D.

1985-01-01

The chemistry of the processes involved in the hot dewatering and encapsulation of alumino-ferric hydroxide floc in epoxide resin have been studied. Pretreatment of the floc to reduce resin attack and hydrolysis and to increase the dimensional stability of the solidified wasteform has been evaluated. It has been demonstrated that removal of ammonium nitrate from the floc and control of the residual water in the resin are important factors in ensuring dimensional stability of the solidified resin. Resin systems have been identified which, together with the appropriate waste pretreatment have successfully encapsulated a simulated magnox sludge producing a stable wasteform having mechanical and physical properties comparable with the basic resin. (author)

Energy saving type area hot water supply system using heat of hot waste water from the sludge center as hot source for hot water; New energy rokko airando CITY. Surajjisenta karano onhaisuinetsu wo kyuyuyo netsugen ni riyosuru sho energy gata chiiki onsui kyokyu system

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-01-01

Heat source of area hot water supply system in Rokko island City (man-made island) is heat of combustion at the sludge center (sludge incineration plant) in this island. Dehydrated sludge cakes (230ton/day) brought from seven sewage disposal plants in Kobe City is combusted (850degC) in the fluid bed hearth. Combustion gas washed in the scrubber, hot waste water after the washing give heat into heat transfer water in the first heat exchanger. Temperature being 64degC in summer and about 50degC in winter, this heat transfer water is sent into the second heat exchanger at every condominium building throughout the pipe line system circulating in the area. At each home, gas heater and hot water supply devices fitted, additional combustion is not necessary in summer but is used according to demand in other seasons. This hot water supply service has been carried out since 1988 and at present has been used by 3600 homes. Amount of supplying hot water being about 3000cu.m/day, saving is calculated roughly as 60% of gas for hot water supply. Fee for this system is 1500/yen/month uniformly for each home. 14 figs.

Ferrocyanide Safety Program: Waste tank sludge rheology within a hot spot or during draining

International Nuclear Information System (INIS)

Fauske, H.K.; Cash, R.J.

1993-11-01

The conditions under which ferrocyanide waste sludge flows as a homogeneous non-Newtonian two-phase (solid precipitate-liquid) mixture rather than as a liquid through a porous medium (of stationary precipitate) are examined theoretically, based on the notion that the preferred rheological behavior of the sludge is the one which imposes the least resistance to the sludge flow. The homogeneous two-phase mixture is modeled as a power-law fluid and simple criteria are derived that show that the homogeneous power-law sludge-flow is a much more likely flow situation than the porous medium model of sludge flow. The implication of this finding is that the formation of a hot spot or the drainage of sludge from a waste tank are not likely to result in the uncovering (drying) and subsequent potential overheating of the reactive-solid component of the sludge

In situ stabilization wall for containment and hot spot retrieval

International Nuclear Information System (INIS)

Loomis, G.G.

1996-01-01

This paper presents the results of a full scale field demonstration of a in situ stabilization technology applicable to buried transuranic waste. The technology involves creating a jet grouted wall around selected regions or hot spots within a buried waste site. The resulting wall provides a barrier against further horizontal migration of the contaminants and allows vertical digging of material inside the wall, thus minimizing waste during a hot spot removal action. The demonstration involved creating a open-quotes Uclose quotes shaped wall in the interior of a full sized, simulated waste pit. The wall simulated the main features of a four sided wall. The demonstration also involved a destructive examination and a stability test for a hot spot retrieval scenario

Potential of Waste Oyster Shells as a Novel Biofiller for Hot-Mix Asphalt

Directory of Open Access Journals (Sweden)

Nader Nciri

2018-03-01

Full Text Available This paper reports the use of waste oyster shells as a novel biofiller for hot-mix asphalt (HMA pavement applications. The effects of different fractions (e.g., 0, 5, 10, 15 wt % of oyster shell powder (OSP on the bitumen performance were investigated. The chemical properties of unfilled and OSP-filled asphalts were characterized by means of thin layer chromatography-ionization detection (TLC-FID, Fourier transform-infrared spectroscopy (FT-IR, X-ray diffraction (XRD, and scanning electron microscopy (SEM. Thermal characteristics were examined by thermogravimetric analysis (TGA and differential scanning calorimetry (DSC. Physical and rheological properties were assessed through penetration, softening point, ductility, and dynamic shear rheometer (DSR tests. Results showed that OSP addition increased the resins content, as well as the stiffness of blends. No obvious reactions have occurred between the filler and the asphalt. A higher dose of OSP altered the morphology of the binder, whereas lower and intermediate doses improved its thermal stability and enhanced its low-temperature, rutting, and fatigue performances with respect to the plain asphalt. Overall, the waste oyster shells could be used as filler substitute, not only to improve the quality of road pavements but also to reduce the cost of their construction and solve the waste disposal problems.

Development and demonstration of a stabilization system for buried mixed waste tanks: Initital results of the tank V-9 hot demonstration

International Nuclear Information System (INIS)

Matthern, G.E.; Kuhns, D.J.; Meservey, R.H.; Farnsworth, R.K.

1996-01-01

This paper describes a systematic approach for the stabilization of buried mixed waste tanks and presents the status of an application of this approach to a specific hot waste tank demonstration to be performed in FY-96. The approach uses the cradle-to-grave concept and includes technical, health and safety, and regulatory considerations and requirements. It starts with the identification of the tank and continues to the final disposition and monitoring of the tank

Hot Laboratories and Remote Handling

International Nuclear Information System (INIS)

Bart, G.; Blanc, J.Y.; Duwe, R.

2003-01-01

The European Working Group on ' Hot Laboratories and Remote Handling' is firmly established as the major contact forum for the nuclear R and D facilities at the European scale. The yearly plenary meetings intend to: - Exchange experience on analytical methods, their implementation in hot cells, the methodologies used and their application in nuclear research; - Share experience on common infrastructure exploitation matters such as remote handling techniques, safety features, QA-certification, waste handling; - Promote normalization and co-operation, e.g., by looking at mutual complementarities; - Prospect present and future demands from the nuclear industry and to draw strategic conclusions regarding further needs. The 41. plenary meeting was held in CEA Saclay from September 22 to 24, 2003 in the premises and with the technical support of the INSTN (National Institute for Nuclear Science and Technology). The Nuclear Energy Division of CEA sponsored it. The Saclay meeting was divided in three topical oral sessions covering: - Post irradiation examination: new analysis methods and methodologies, small specimen technology, programmes and results; - Hot laboratory infrastructure: decommissioning, refurbishment, waste, safety, nuclear transports; - Prospective research on materials for future applications: innovative fuels (Generation IV, HTR, transmutation, ADS), spallation source materials, and candidate materials for fusion reactor. A poster session was opened to transport companies and laboratory suppliers. The meeting addressed in three sessions the following items: Session 1 - Post Irradiation Examinations. Out of 12 papers (including 1 poster) 7 dealt with surface and solid state micro analysis, another one with an equally complex wet chemical instrumental analytical technique, while the other four papers (including the poster) presented new concepts for digital x-ray image analysis; Session 2 - Hot laboratory infrastructure (including waste theme) which was

Avoiding Carbon Bed Hot Spots in Thermal Process Off-Gas Systems

International Nuclear Information System (INIS)

Soelberg, Nick; Enneking, Joe

2011-01-01

Mercury has had various uses in nuclear fuel reprocessing and other nuclear processes, and so is often present in radioactive and mixed (radioactive and hazardous) wastes. Test programs performed in recent years have shown that mercury in off-gas streams from processes that treat radioactive wastes can be controlled using fixed beds of activated sulfur-impregnated carbon, to levels low enough to comply with air emission regulations such as the Hazardous Waste Combustor (HWC) Maximum Achievable Control Technology (MACT) standards. Carbon bed hot spots or fires have occurred several times during these tests, and also during a remediation of tanks that contained mixed waste. Hot spots occur when localized areas in a carbon bed become heated to temperatures where oxidation occurs. This heating typically occurs due to heat of absorption of gas species onto the carbon, but it can also be caused through external means such as external heaters used to heat the carbon bed vessel. Hot spots, if not promptly mitigated, can grow into bed fires. Carbon bed hot spots and fires must be avoided in processes that treat radioactive and mixed waste. Hot spots are detected by (a) monitoring in-bed and bed outlet gas temperatures, and (b) more important, monitoring of bed outlet gas CO concentrations. Hot spots are mitigated by (a) designing for appropriate in-bed gas velocity, for avoiding gas flow maldistribution, and for sufficient but not excessive bed depth, (b) appropriate monitoring and control of gas and bed temperatures and compositions, and (c) prompt implementation of corrective actions if bed hot spots are detected. Corrective actions must be implemented quickly if bed hot spots are detected, using a graded approach and sequence starting with corrective actions that are simple, quick, cause the least impact to the process, and are easiest to recover from.

An Applied Study on the Decontamination and Decommissioning of Hot Cell Facilities in the United States and Comparison with the Studsvik Facility for Solid and Liquid Waste

International Nuclear Information System (INIS)

Varley, Geoff; Rusch, Chris

2006-07-01

This report presents the plans, processes and results of the decontamination and decommissioning of the Hot Cell Facility in Building 23 at the General Atomics Torrey Pines Mesa Facility (HCF) and compares the program and cost of decommissioning HCF with the Swedish cost estimate for decontamination and decommissioning of the HM hot cell and wastes treatment facility at Studsvik in Sweden. The HCF had three main hot cells and was licensed to: Receive, handle and ship radioactive materials; Remotely handle, examine and store irradiated fuel materials; Extract tritium (engineering scale); Support new reactor production development; Develop, fabricate and inspect UO 2 - BeO fuel materials. The HM facility in Studsvik was constructed to handle and package medium-active solid and liquid wastes, prior to disposal. Central to the facility is a conventional hot cell including three work stations, serviced by master slave manipulators. Other parts of the facility include holding tanks for liquid wastes and slurries, a centrifuge room, as well as an encapsulation station where drummed wastes can be encapsulated in cement, offices, laboratories and workshops and so on, as well as building and cell ventilation systems. Decontamination and decommissioning of the HCF took place during 1993 through 2001. The objective was to obtain regulatory release of the site so that it could be used on an unrestricted basis. Based on data from extensive hazardous and radiological materials characterization, GA evaluated four decommissioning options and selected dismantling as the only option that would satisfy the decommissioning objective. The decontamination and decommissioning scope included the following actions. 1. Remove the legacy waste that consisted of radioactive wastes stored at the HCF consisting of 21,434 kg of irradiated fuel material (IFM) that was owned by the US DoE and store the waste in temporary storage set up at the GA site. 2. Actual Decontamination and Dismantlement

An Applied Study on the Decontamination and Decommissioning of Hot Cell Facilities in the United States and Comparison with the Studsvik Facility for Solid and Liquid Waste

Energy Technology Data Exchange (ETDEWEB)

Varley, Geoff; Rusch, Chris [NAC International, Atlanta, GA (United States)

2006-07-15

This report presents the plans, processes and results of the decontamination and decommissioning of the Hot Cell Facility in Building 23 at the General Atomics Torrey Pines Mesa Facility (HCF) and compares the program and cost of decommissioning HCF with the Swedish cost estimate for decontamination and decommissioning of the HM hot cell and wastes treatment facility at Studsvik in Sweden. The HCF had three main hot cells and was licensed to: Receive, handle and ship radioactive materials; Remotely handle, examine and store irradiated fuel materials; Extract tritium (engineering scale); Support new reactor production development; Develop, fabricate and inspect UO{sub 2} - BeO fuel materials. The HM facility in Studsvik was constructed to handle and package medium-active solid and liquid wastes, prior to disposal. Central to the facility is a conventional hot cell including three work stations, serviced by master slave manipulators. Other parts of the facility include holding tanks for liquid wastes and slurries, a centrifuge room, as well as an encapsulation station where drummed wastes can be encapsulated in cement, offices, laboratories and workshops and so on, as well as building and cell ventilation systems. Decontamination and decommissioning of the HCF took place during 1993 through 2001. The objective was to obtain regulatory release of the site so that it could be used on an unrestricted basis. Based on data from extensive hazardous and radiological materials characterization, GA evaluated four decommissioning options and selected dismantling as the only option that would satisfy the decommissioning objective. The decontamination and decommissioning scope included the following actions. 1. Remove the legacy waste that consisted of radioactive wastes stored at the HCF consisting of 21,434 kg of irradiated fuel material (IFM) that was owned by the US DoE and store the waste in temporary storage set up at the GA site. 2. Actual Decontamination and

Geostatistical risk estimation at waste disposal sites in the presence of hot spots

International Nuclear Information System (INIS)

Komnitsas, Kostas; Modis, Kostas

2009-01-01

The present paper aims to estimate risk by using geostatistics at the wider coal mining/waste disposal site of Belkovskaya, Tula region, in Russia. In this area the presence of hot spots causes a spatial trend in the mean value of the random field and a non-Gaussian data distribution. Prior to application of geostatistics, subtraction of trend and appropriate smoothing and transformation of the data into a Gaussian form were carried out; risk maps were then generated for the wider study area in order to assess the probability of exceeding risk thresholds. Finally, the present paper discusses the need for homogenization of soil risk thresholds regarding hazardous elements that will enhance reliability of risk estimation and enable application of appropriate rehabilitation actions in contaminated areas.

Residential hot water distribution systems: Roundtablesession

Energy Technology Data Exchange (ETDEWEB)

Lutz, James D.; Klein, Gary; Springer, David; Howard, Bion D.

2002-08-01

Residential building practice currently ignores the lossesof energy and water caused by the poor design of hot water systems. Theselosses include: combustion and standby losses from water heaters, thewaste of water (and energy) while waiting for hot water to get to thepoint of use; the wasted heat as water cools down in the distributionsystem after a draw; heat losses from recirculation systems and thediscarded warmth of waste water as it runs down the drain. Severaltechnologies are available that save energy (and water) by reducing theselosses or by passively recovering heat from wastewater streams and othersources. Energy savings from some individual technologies are reported tobe as much as 30 percent. Savings calculations of prototype systemsincluding bundles of technologies have been reported above 50 percent.This roundtable session will describe the current practices, summarizethe results of past and ongoing studies, discuss ways to think about hotwater system efficiency, and point to areas of future study. We will alsorecommend further steps to reduce unnecessary losses from hot waterdistribution systems.

Waste canister for storage of nuclear wastes

Science.gov (United States)

Duffy, James B.

1977-01-01

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

M3FT-17OR0301070211 - Preparation of Hot Isostatically Pressed AgZ Waste Form Samples

Energy Technology Data Exchange (ETDEWEB)

Jubin, Robert Thomas [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bruffey, Stephanie H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jordan, Jacob A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-10-01

The production of radioactive iodine-bearing waste forms that exhibit long-term stability and are suitable for permanent geologic disposal has been the subject of substantial research interest. One potential method of iodine waste form production is hot isostatic pressing (HIP). Recent studies at Oak Ridge National Laboratory (ORNL) have investigated the conversion of iodine-loaded silver mordenite (I-AgZ) directly to a waste form by HIP. ORNL has performed HIP with a variety of sample compositions and pressing conditions. The base mineral has varied among AgZ (in pure and engineered forms), silver-exchanged faujasite, and silverexchanged zeolite A. Two iodine loading methods, occlusion and chemisorption, have been explored. Additionally, the effects of variations in temperature and pressure of the process have been examined, with temperature ranges of 525°C–1,100°C and pressure ranges of 100–300 MPa. All of these samples remain available to collaborators upon request. The sample preparation detailed in this document is an extension of that work. In addition to previously prepared samples, this report documents the preparation of additional samples to support stability testing. These samples include chemisorbed I-AgZ and pure AgI. Following sample preparation, each sample was processed by HIP by American Isostatic Presses Inc. and returned to ORNL for storage. ORNL will store the samples until they are requested by collaborators for durability testing. The sample set reported here will support waste form durability testing across the national laboratories and will provide insight into the effects of varied iodine content on iodine retention by the produced waste form and on potential improvements in waste form durability provided by the zeolite matrix.

Fiscal 1999 report on result of the model project for waste heat recovery in hot blast stove; 1999 nendo netsufuro hainetsu kaishu model jigyo seika hokokusho

Energy Technology Data Exchange (ETDEWEB)

NONE

2000-07-01

For the purpose of curtailing energy consumption of the steel industry, a heavy energy consuming industry in China, a model project was carried out for waste heat recovery in a hot blast stove, with the fiscal 1999 results reported. In the process of this project, a heat exchanger for recovering heat is installed in the exhaust gas flue of a hot blast stove in ironworks, with sensible heat recovered through a heating medium. The heat exchanger for recovering heat and the preheating heat exchanger, which was installed in the main pipe for blast furnace gas and for combustion air, were connected by pressure piping, with the blast furnace gas and the combustion air preheated. In addition, a heating medium circulating pump for transporting the heating medium is installed, as are an expansion tank for absorbing expansion/contraction due to change in temperature, a heating medium storage tank for accepting the entire heating medium in the system for the maintenance of the equipment, and heating medium feeding pump, for example. This year, on the basis of the 'Agreement Annex', basic designs and detailed designs were performed for each equipment in the waste heat recovering equipment for the hot blast stove. Further, procurement and manufacturing were implemented for various component parts and devices of the waste heat recovering equipment. (NEDO)

Waste canister for storage of nuclear wastes

International Nuclear Information System (INIS)

Duffy, J.B.

1977-01-01

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

Waste reduction efforts through the evaluation and procurement of a digital camera system for the Alpha-Gamma Hot Cell Facility at Argonne National Laboratory-East

International Nuclear Information System (INIS)

Bray, T. S.; Cohen, A. B.; Tsai, H.; Kettman, W. C.; Trychta, K.

1999-01-01

The Alpha-Gamma Hot Cell Facility (AGHCF) at Argonne National Laboratory-East is a research facility where sample examinations involve traditional photography. The AGHCF documents samples with photographs (both Polaroid self-developing and negative film). Wastes generated include developing chemicals. The AGHCF evaluated, procured, and installed a digital camera system for the Leitz metallograph to significantly reduce labor, supplies, and wastes associated with traditional photography with a return on investment of less than two years

Process and equipment development for hot isostatic pressing treatability study

Energy Technology Data Exchange (ETDEWEB)

Bateman, Ken; Wahlquist, Dennis; Malewitz, Tim

2015-03-01

Battelle Energy Alliance (BEA), LLC, has developed processes and equipment for a pilot-scale hot isostatic pressing (HIP) treatability study to stabilize and volume reduce radioactive calcine stored at Idaho National Laboratory (INL). In 2009, the U. S. Department of Energy signed a Record of Decision with the state of Idaho selecting HIP technology as the method to treat 5,800 yd^3 (4,400 m^3) of granular zirconia and alumina calcine produced between 1953 and 1992 as a waste byproduct of spent nuclear fuel reprocessing. Since the 1990s, a variety of radioactive and hazardous waste forms have been remotely treated using HIP within INL hot cells. To execute the remote process at INL, waste is loaded into a stainless-steel or aluminum can, which is evacuated, sealed, and placed into a HIP furnace. The HIP simultaneously heats and pressurizes the waste, reducing its volume and increasing its durability. Two 1 gal cans of calcine waste currently stored in a shielded cask were identified as candidate materials for a treatability study involving the HIP process. Equipment and materials for cask-handling and calcine transfer into INL hot cells, as well as remotely operated equipment for waste can opening, particle sizing, material blending, and HIP can loading have been designed and successfully tested. These results demonstrate BEA’s readiness for treatment of INL calcine.

The development of synthetic test procedure for hot cell equipment systems in IMEF

International Nuclear Information System (INIS)

Ahn, Sang Bok; Lee, Key Soon; Park, Dae Kyu; Hong, Kwon Pyo; Choo, Yong Sun

1998-04-01

Hot cell facility should be confirmed to operation safety through pre-commissioning test after construction. In this report, the detailed procedure of hot cell equipment are described. The contents are as follows: 1. Entrance equipment of hot cell 2. Specimen transportation equipment between hot cells 3. Waste discharge equipment in hot cell 4. Specimen loading equipment to hot cell 5. Interlinking equipment in hot cell. (author). 4 tabs

A new waste minimization method for the determination of total nonhalogenated volatile organic compounds in TRU wastes

International Nuclear Information System (INIS)

Sandoval, W.; Quintana, B.D.; Ortega, L.

1997-01-01

As part of the technical support CST-12 provides for a wide variety of defense and nondefense programs within Los Alamos National Laboratory (LANL) and the Department of Energy (DOE) complex, new waste minimization technique is under development for radiological volatile organic analysis (Hot VOA). Currently all HOT VOA must be run in a glovebox. Several types of sample contain TRU radiological waste in the form of particulates. By prefiltering the samples through a 1.2 micron syringe and counting the radioactivity, it has been found that many of the samples can be analyzed outside a glovebox. In the present investigation, the types of Hot VOA samples that can take advantage of this new technique, the volume and types of waste reduced and the experimental parameters will be discussed. Overall, the radioactive waste generated is minimized

Effect of hot pressing additives on the leachability of hot pressed sodium hydrous titanium oxide

International Nuclear Information System (INIS)

Valentine, T.M.; Sambell, R.A.J.

1980-01-01

Sodium hydrous titanium oxide is an ion exchange resin which can be used for immobilizing medium level waste (MLW) liquors. When hot pressed, it undergoes conversion to a ceramic. Three low melting point materials (borax, bismuth trioxide, and a mixture of PbO/CuO) were added to the (Na)HTiO and the effect that each of these had on aiding densification was assessed. Hot pressing temperature, applied pressure, and percentage addition of hot pressing aid were varied. Percentage open porosity, flexural strength, and leachability were measured. There was a linear relationship between the percentage open porosity and the logarithm of the leach rate for a constant percentage addition of each additive

Hot sample archiving. Revision 3

International Nuclear Information System (INIS)

McVey, C.B.

1995-01-01

This Engineering Study revision evaluated the alternatives to provide tank waste characterization analytical samples for a time period as recommended by the Tank Waste Remediation Systems Program. The recommendation of storing 40 ml segment samples for a period of approximately 18 months (6 months past the approval date of the Tank Characterization Report) and then composite the core segment material in 125 ml containers for a period of five years. The study considers storage at 222-S facility. It was determined that the critical storage problem was in the hot cell area. The 40 ml sample container has enough material for approximately 3 times the required amount for a complete laboratory re-analysis. The final result is that 222-S can meet the sample archive storage requirements. During the 100% capture rate the capacity is exceeded in the hot cell area, but quick, inexpensive options are available to meet the requirements

Waste incineration

International Nuclear Information System (INIS)

Rumplmayr, A.; Sammer, G.

2001-01-01

Waste incineration can be defined as the thermal conversion processing of solid waste by chemical oxidation. The types of wastes range from solid household waste and infectious hospital waste through to toxic solid, liquid and gaseous chemical wastes. End products include hot incineration gases, composed primarily of nitrogen, carbon dioxide, water vapor and to a smaller extend of non-combustible residue (ash) and air pollutants (e. g. NO x ). Energy can be recovered by heat exchange from the hot incineration gases, thus lowering fossil fuel consumption that in turn can reduce emissions of greenhouse gases. Burning of solid waste can fulfil up to four distinctive objectives (Pera, 2000): 1. Volume reduction: volume reduction of about 90 %, weight reduction of about 70 %; 2. Stabilization of waste: oxidation of organic input; 3. Recovery of energy from waste; 4. Sanitization of waste: destruction of pathogens. Waste incineration is not a means to make waste disappear. It does entail emissions into air as well as water and soil. The generated solid residues are the topic of this task force. Unlike other industrial processes discussed in this platform, waste incineration is not a production process, and is therefore not generating by-products, only residues. Residues that are isolated from e. g. flue gas, are concentrated in another place and form (e. g. air pollution control residues). Hence, there are generally two groups of residues that have to be taken into consideration: residues generated in the actual incineration process and others generated in the flue gas cleaning system. Should waste incineration finally gain public acceptance, it will be necessary to find consistent regulations for both sorts of residues. In some countries waste incineration is seen as the best option for the treatment of waste, whereas in other countries it is seen very negative. (author)

Comparative waste forms study

International Nuclear Information System (INIS)

Wald, J.W.; Lokken, R.O.; Shade, J.W.; Rusin, J.M.

1980-12-01

A number of alternative process and waste form options exist for the immobilization of nuclear wastes. Although data exists on the characterization of these alternative waste forms, a straightforward comparison of product properties is difficult, due to the lack of standardized testing procedures. The characterization study described in this report involved the application of the same volatility, mechanical strength and leach tests to ten alternative waste forms, to assess product durability. Bulk property, phase analysis and microstructural examination of the simulated products, whose waste loading varied from 5% to 100% was also conducted. The specific waste forms investigated were as follows: Cold Pressed and Sintered PW-9 Calcine; Hot Pressed PW-9 Calcine; Hot Isostatic Pressed PW-9 Calcine; Cold Pressed and Sintered SPC-5B Supercalcine; Hot Isostatic pressed SPC-5B Supercalcine; Sintered PW-9 and 50% Glass Frit; Glass 76-68; Celsian Glass Ceramic; Type II Portland Cement and 10% PW-9 Calcine; and Type II Portland Cement and 10% SPC-5B Supercalcine. Bulk property data were used to calculate and compare the relative quantities of waste form volume produced at a spent fuel processing rate of 5 metric ton uranium/day. This quantity ranged from 3173 L/day (5280 Kg/day) for 10% SPC-5B supercalcine in cement to 83 L/day (294 Kg/day) for 100% calcine. Mechanical strength, volatility, and leach resistance tests provide data related to waste form durability. Glass, glass-ceramic and supercalcine ranked high in waste form durability where as the 100% PW-9 calcine ranked low. All other materials ranked between these two groupings

Thermoelectric Power Generation System for Future Hybrid Vehicles Using Hot Exhaust Gas

Science.gov (United States)

Kim, Sun-Kook; Won, Byeong-Cheol; Rhi, Seok-Ho; Kim, Shi-Ho; Yoo, Jeong-Ho; Jang, Ju-Chan

2011-05-01

The present experimental and computational study investigates a new exhaust gas waste heat recovery system for hybrid vehicles, using a thermoelectric module (TEM) and heat pipes to produce electric power. It proposes a new thermoelectric generation (TEG) system, working with heat pipes to produce electricity from a limited hot surface area. The current TEG system is directly connected to the exhaust pipe, and the amount of electricity generated by the TEMs is directly proportional to their heated area. Current exhaust pipes fail to offer a sufficiently large hot surface area for the high-efficiency waste heat recovery required. To overcome this, a new TEG system has been designed to have an enlarged hot surface area by the addition of ten heat pipes, which act as highly efficient heat transfer devices and can transmit the heat to many TEMs. As designed, this new waste heat recovery system produces a maximum 350 W when the hot exhaust gas heats the evaporator surface of the heat pipe to 170°C; this promises great possibilities for application of this technology in future energy-efficient hybrid vehicles.

High-level radioactive waste incorporation into (special) cements

International Nuclear Information System (INIS)

Roy, D.M.; Gouda, G.R.

1978-01-01

A feasibility study has demonstrated that very strong, durable, relatively impermeable cylinders may be prepared by hot pressing combinations of cements with simulated radioactive waste solids. While the properties have not been studied exhaustively, the results suggest an optional method for immobilization and isolation of radioactive waste. Samples prepared with calcium aluminate cements appeared to have properties superior to those with Portland cements. Four simulated radioactive waste compositions having high rare-earth oxide contents, and some containing a large excess of NaNO 3 , were studied. Modest temperatures [423 to 673 K (150 to 400 0 C)] were used for hot pressing at pressures from 178 to 345 MPa. Dense strong very low porosity specimens resulted when mixtures containing from 10 to 50% waste were hot pressed, incorporating also a small percentage of water. In addition, high-strength cement cylinders were prepared with the waste solid (approximately 20 wt% waste) in a separate core and were very resistant to leaching by water near its boiling point. With this configuration, even the NaNO 3 -containing wastes were resistant to leaching by water

Performance Monitoring of Residential Hot Water Distribution Systems

Energy Technology Data Exchange (ETDEWEB)

Liao, Anna; Lanzisera, Steven; Lutz, Jim; Fitting, Christian; Kloss, Margarita; Stiles, Christopher

2014-08-11

Current water distribution systems are designed such that users need to run the water for some time to achieve the desired temperature, wasting energy and water in the process. We developed a wireless sensor network for large-scale, long time-series monitoring of residential water end use. Our system consists of flow meters connected to wireless motes transmitting data to a central manager mote, which in turn posts data to our server via the internet. This project also demonstrates a reliable and flexible data collection system that could be configured for various other forms of end use metering in buildings. The purpose of this study was to determine water and energy use and waste in hot water distribution systems in California residences. We installed meters at every end use point and the water heater in 20 homes and collected 1s flow and temperature data over an 8 month period. For a typical shower and dishwasher events, approximately half the energy is wasted. This relatively low efficiency highlights the importance of further examining the energy and water waste in hot water distribution systems.

Waste package characterisation

Energy Technology Data Exchange (ETDEWEB)

Sannen, L.; Bruggeman, M.; Wannijn, J.P

1998-09-01

Radioactive wastes originating from the hot labs of the Belgian Nuclear Research Centre SCK-CEN contain a wide variety of radiotoxic substances. The accurate characterisation of the short- and long-term radiotoxic components is extremely difficult but required in view of geological disposal. This paper describes the methodology which was developed and adopted to characterise the high- and medium-level waste packages at the SCK-CEN hot laboratories. The proposed method is based on the estimation of the fuel inventory evacuated in a particular waste package; a calculation of the relative fission product contribution on the fuel fabrication and irradiation footing; a comparison of the calculated, as expected, dose rate and the real measured dose rate of the waste package. To cope with the daily practice an appropriate fuel inventory estimation route, a user friendly computer programme for fission product and corresponding dose rate calculation, and a simple dose rate measurement method have been developed and implemented.

Waste package characterisation

International Nuclear Information System (INIS)

Sannen, L.; Bruggeman, M.; Wannijn, J.P.

1998-09-01

Radioactive wastes originating from the hot labs of the Belgian Nuclear Research Centre SCK-CEN contain a wide variety of radiotoxic substances. The accurate characterisation of the short- and long-term radiotoxic components is extremely difficult but required in view of geological disposal. This paper describes the methodology which was developed and adopted to characterise the high- and medium-level waste packages at the SCK-CEN hot laboratories. The proposed method is based on the estimation of the fuel inventory evacuated in a particular waste package; a calculation of the relative fission product contribution on the fuel fabrication and irradiation footing; a comparison of the calculated, as expected, dose rate and the real measured dose rate of the waste package. To cope with the daily practice an appropriate fuel inventory estimation route, a user friendly computer programme for fission product and corresponding dose rate calculation, and a simple dose rate measurement method have been developed and implemented

Application of annular centrifugal contactors in the hot test of the improved total partitioning process for high level liquid waste.

Science.gov (United States)

Duan, Wuhua; Chen, Jing; Wang, Jianchen; Wang, Shuwei; Feng, Xiaogui; Wang, Xinghai; Li, Shaowei; Xu, Chao

2014-08-15

High level liquid waste (HLLW) produced from the reprocessing of the spent nuclear fuel still contains moderate amounts of uranium, transuranium (TRU) actinides, (90)Sr, (137)Cs, etc., and thus constitutes a permanent hazard to the environment. The partitioning and transmutation (P&T) strategy has increasingly attracted interest for the safe treatment and disposal of HLLW, in which the partitioning of HLLW is one of the critical technical issues. An improved total partitioning process, including a TRPO (tri-alkylphosphine oxide) process for the removal of actinides, a CESE (crown ether strontium extraction) process for the removal of Sr, and a CECE (calixcrown ether cesium extraction) process for the removal of Cs, has been developed to treat Chinese HLLW. A 160-hour hot test of the improved total partitioning process was carried out using 72-stage 10-mm-dia annular centrifugal contactors (ACCs) and genuine HLLW. The hot test results showed that the average DFs of total α activity, Sr and Cs were 3.57 × 10(3), 2.25 × 10(4) and 1.68 × 10(4) after the hot test reached equilibrium, respectively. During the hot test, 72-stage 10-mm-dia ACCs worked stable, continuously with no stage failing or interruption of the operation. Copyright © 2014 Elsevier B.V. All rights reserved.

Hot Isostatic Press (HIP) vitrification of radwaste concretes

International Nuclear Information System (INIS)

Siemer, D.D.; Scheetz, B.; Gougar, M.L.D.

1995-01-01

Properly formulated and properly ''canned'' radwaste concretes can be readily hot-isostatically-pressed (HIPed) into materials that exhibit performance equivalent to typical radwaste-type glasses. The HIPing conditions (temperature/pressure) required to turn a concrete waste form into a ''vitrified'' waste form are quite mild and therefore consistent with both safety and high productivity. This paper describes the process and its products with reference to its potential application to Idaho Chemical Processing Plant (ICPP) reprocessing wastes

Mechanical properties of hot bituminous mixes manufactured with recycled aggregate of Silestone® waste

Directory of Open Access Journals (Sweden)

Rubio, M. C.

2011-03-01

Full Text Available This article presents the results of a research project which analyzes the viability of incorporating waste material from decorative quartz solid surfacing in the manufacture of hot bituminous mixes. For this purpose, various bituminous mixes were manufactured with waste aggregate in different percentage. A set of tests were carried out that permitted the characterization of the mechanical behavior of these mixes. The results of these tests were similar to the results obtained when conventional mixes made from virgin quarry aggregate were tested. The results for moisture sensitivity as well as for wheel track rutting on mixes showed the optimal performance of this waste material even under very demanding traffic conditions. Laboratory studies showed that the use of this waste material in the manufacture of bituminous mixes is technically viable and can provide high-quality recycled aggregates at a very low cost, which can be used in the pavement of road.

Este artículo muestra los resultados de un proyecto de investigación que tiene por objeto analizar la viabilidad de incorporación de residuos procedentes de piedra decorativa en la fabricación de mezclas bituminosas en caliente. Para ello se fabricaron mezclas con áridos reciclados de residuos procedentes de piedra decorativa en distintos porcentajes, realizando un conjunto de ensayos que posibilitaron caracterizar el comportamiento mecánico de dichas mezclas. Los resultados obtenidos fueron similares a los de las mezclas convencionales fabricadas con áridos vírgenes procedentes de cantera. Los valores de resistencia conservada ante la acción del agua y deformación en pista, pusieron de manifiesto la aptitud del residuo incluso ante las condiciones de tráfico más exigentes. Los trabajos realizados en el laboratorio indican que la utilización del residuo en la fabricación de mezclas bituminosas es técnicamente viable, pudiendo obtenerse áridos reciclados de gran calidad y

Construction of concrete hot cells

International Nuclear Information System (INIS)

1981-12-01

The standard is to be applied to rooms (hot cells) which are enclosed by a concrete shield and in which radioactive material is handled by remote control. The rooms may be in facilities for experimental purposes (e.g. development of fuel elements and materials or of chemical processes) or in facilities for production purposes (e.g. reprocessing of nuclear fuel or treatment of radioactive wastes). The standard is to give a design hasis for concrete hot cells and their installations which is to be applied by designers, constructors, future users and competent authorities as well as independent experts. (orig.) [de

Construction of concrete hot cells

International Nuclear Information System (INIS)

1980-09-01

The standard is to be applied to rooms (hot cells) which are enclosed by a concrete shield and in which radioactive material is handled by remote control. The rooms may be in facilities for experimental purposes (e.g. development of fuel elements and materials or of chemical processes) or in facilities for production purposes (e.g. reprocessing of nuclear fuel or treatment of radioactive wastes). The standard is to give a design basis for concrete hot cells and their installations which is to be applied by designers, constructors, future users and competent authorities as well as independent experts. (orig.) [de

Discussion of Carbon Emissions for Charging Hot Metal in EAF Steelmaking Process

Science.gov (United States)

Yang, Ling-zhi; Jiang, Tao; Li, Guang-hui; Guo, Yu-feng

2017-07-01

As the cost of hot metal is reduced for iron ore prices are falling in the international market, more and more electric arc furnace (EAF) steelmaking enterprises use partial hot metal instead of scrap as raw materials to reduce costs and the power consumption. In this paper, carbon emissions based on 1,000 kg molten steel by charging hot metal in EAF steelmaking is studied. Based on the analysis of material and energy balance calculation in EAF, the results show that 146.9, 142.2, 137.0, and 130.8 kg/t of carbon emissions are produced at a hot metal ratio of 0 %, 30 %, 50 %, and 70 %, while 143.4, 98.5, 65.81, and 31.5 kg/t of carbon emissions are produced at a hot metal ratio of 0 %, 30 %, 50 %, and 70 % by using gas waste heat utilization (coal gas production) for EAF steelmaking unit process. However, carbon emissions are increased by charging hot metal for the whole blast furnace-electric arc furnace (BF-EAF) steelmaking process. In the condition that the hot metal produced by BF is surplus, as carbon monoxide in gas increased by charging hot metal, the way of coal gas production can be used for waste heat utilization, which reduces carbon emissions in EAF steelmaking unit process.

Current status of JAERI Tokai hot cell facilities

International Nuclear Information System (INIS)

Itami, Hiroharu; Morozumi, Minoru; Yamahara, Takeshi

1992-01-01

JAERI has 4 hot cell facilities in order to examine high radioactive materials. Three of them, the Research Hot Laboratory, the Reactor Fuel Examination Facility and the Waste Safety Testing Facility are located in the JAERI Tokai site, and the rest is the JMTR Hot Laboratory in the Oarai site. The Research Hot Laboratory (RHL) was constructed for post-irradiation examination (PIE), especially nuclear related basic research experiment, such as metallurgical, chemical and mechanical examination on fuels and materials irradiated in research and test reactors. This facility has 10 large dimension concrete and 38 lead cells. At present the RHL is used for various kinds of examinations of high radioactive samples such as fuels of research and test reactors, power reactors and high temperature testing reactor (HTTR), and structural materials. The Reactor Fuel Examination Facility (RFEF) was designed and constructed for carrying out PIE of irradiated full-size fuel assemblies of light water reactors (LWRs). This facility has a storage pool, 8 concrete and 5 lead cells. They are currently used for safety evaluation on high burnup and advanced lWR fuels as part of the national program. The Waste Safety Testing Facility (WASTEF) was designed and constructed for safety research on long-term storage and disposal of high level radioactive wastes, generated by fuel reprocessing. The WASTEF has 5 concrete cells and 1 lead cell. Examinations on the behavior of various long-lived fission products in a glass form and in a canister and, releasing behavior of them out of a canister are carrying out under the condition at storage. (author)

Preparation techniques for ceramic waste form powder

International Nuclear Information System (INIS)

Hash, M.C.; Pereira, C.; Lewis, M.A.

1997-01-01

The electrometallurgical treatment of spent nuclear fuels result in a chloride waste salt requiring geologic disposal. Argonne National Laboratory (ANL) is developing ceramic waste forms which can incorporate this waste. Currently, zeolite- or sodalite-glass composites are produced by hot isostatic pressing (HIP) techniques. Powder preparations include dehydration of the raw zeolite powders, hot blending of these zeolite powders and secondary additives. Various approaches are being pursued to achieve adequate mixing, and the resulting powders have been HIPed and characterized for leach resistance, phase equilibria, and physical integrity

Testing and development strategy for the tank waste remediation system

International Nuclear Information System (INIS)

Reddick, G.W.

1994-12-01

This document provides a strategy for performing radioactive (hot) and nonradioactive testing to support processing tank waste. It evaluates the need for hot pilot plant(s) to support pretreatment and other processing functions and presents a strategy for performing hot test work. A strategy also is provided for nonradioactive process and equipment testing. The testing strategy supports design, construction, startup, and operation of Tank Waste Remediation System (TWRS) facilities

Testing and development strategy for the tank waste remediation system

International Nuclear Information System (INIS)

Reddick, G.W.

1995-01-01

This document provides a strategy for performing radioactive (hot) and nonradioactive testing to support processing tank waste. It evaluates the need for hot pilot plant(s) to support pretreatment and other processing functions and presents a strategy for performing hot test work. A strategy also is provided for nonradioactive process and equipment testing. The testing strategy supports design, construction, startup, and operation of Tank Waste Remediation System (TWRS) facilities

Talc-silicon glass-ceramic waste forms for immobilization of high- level calcined waste

International Nuclear Information System (INIS)

Vinjamuri, K.

1993-06-01

Talc-silicon glass-ceramic waste forms are being evaluated as candidates for immobilization of the high level calcined waste stored onsite at the Idaho Chemical Processing Plant. These glass-ceramic waste forms were prepared by hot isostatically pressing a mixture of simulated nonradioactive high level calcined waste, talc, silicon and aluminum metal additives. The waste forms were characterized for density, chemical durability, and glass and crystalline phase compositions. The results indicate improved density and chemical durability as the silicon content is increased

Radioactive spent resins conditioning by the hot super-compaction process

International Nuclear Information System (INIS)

Roth, Andreas; Centner, Baudouin; Lemmens, Alain

2007-01-01

Spent ion exchanger media are considered to be problematic waste that, in many cases, requires special approaches and precautions during its immobilization to meet the acceptance criteria for disposal. The waste acceptance criteria define, among others, the quality of waste forms for disposal, and therefore will sometimes define appropriate treatment options. The selection of treatment options for spent ion exchange materials must consider their physical and chemical characteristics. Basically, the main methods for the treatment of spent organic ion exchange materials, following to pretreatment methods are: - Direct immobilization, producing a stable end product by using Cement, Bitumen, Polymer or High Integrity Containers, - The destruction of the organic compounds by using Thermochemical processes or Oxidation to produce an inorganic intermediate product that may or may not be further conditioned for storage and/or disposal, - The complete removal of the resin inner structural water by a thermal process. After a thorough technical economical analysis, Tractebel Engineering selected the Resin Hot Compaction Process to be installed at Tihange Nuclear Power Plant. The Resin Hot Compaction Process is used to make dense homogenous organic blocks from a wide range of particulate waste. In this process spent resins are first dewatered and dried to remove the inner structural water content. The drying takes place in a drying vessel that holds the contents of two 200 L drums (Figure). In the oil heated drying and mixing unit, the resins are heated to the necessary process temperature for the hot pressing step and then placed into special metal drums, which are automatically lidded and immediately transferred to a high force compactor. After high force compaction the pellets are transferred to a measuring unit, where the dose rate, height and weight are automatically measured and recorded. A volume reduction factor of approximately up to four (depending on the type of

Development of a test system for high level liquid waste partitioning

Directory of Open Access Journals (Sweden)

Duan Wu H.

2015-01-01

Full Text Available The partitioning and transmutation strategy has increasingly attracted interest for the safe treatment and disposal of high level liquid waste, in which the partitioning of high level liquid waste is one of the critical technical issues. An improved total partitioning process, including a tri-alkylphosphine oxide process for the removal of actinides, a crown ether strontium extraction process for the removal of strontium, and a calixcrown ether cesium extraction process for the removal of cesium, has been developed to treat Chinese high level liquid waste. A test system containing 72-stage 10-mm-diam annular centrifugal contactors, a remote sampling system, a rotor speed acquisition-monitoring system, a feeding system, and a video camera-surveillance system was successfully developed to carry out the hot test for verifying the improved total partitioning process. The test system has been successfully used in a 160 hour hot test using genuine high level liquid waste. During the hot test, the test system was stable, which demonstrated it was reliable for the hot test of the high level liquid waste partitioning.

Prototype Development of Remote Operated Hot Uniaxial Press (ROHUP) to Fabricate Advanced Tc-99 Bearing Ceramic Waste Forms - 13381

Energy Technology Data Exchange (ETDEWEB)

Alaniz, Ariana J.; Delgado, Luc R.; Werbick, Brett M. [University of Nevada - Las Vegas, Howard R. Hughes College of Engineering, 4505 S. Maryland Parkway, Box 454009, Las Vegas, NV 89154-4009 (United States); Hartmann, Thomas [University of Nevada - Las Vegas, Harry Reid Canter, 4505 S. Maryland Parkway, Box 454009, Las Vegas, NV 89154-4009 (United States)

2013-07-01

The objective of this senior student project is to design and build a prototype construction of a machine that simultaneously provides the proper pressure and temperature parameters to sinter ceramic powders in-situ to create pellets of rather high densities of above 90% (theoretical). This ROHUP (Remote Operated Hot Uniaxial Press) device is designed specifically to fabricate advanced ceramic Tc-99 bearing waste forms and therefore radiological barriers have been included in the system. The HUP features electronic control and feedback systems to set and monitor pressure, load, and temperature parameters. This device operates wirelessly via portable computer using Bluetooth{sup R} technology. The HUP device is designed to fit in a standard atmosphere controlled glove box to further allow sintering under inert conditions (e.g. under Ar, He, N{sub 2}). This will further allow utilizing this HUP for other potential applications, including radioactive samples, novel ceramic waste forms, advanced oxide fuels, air-sensitive samples, metallic systems, advanced powder metallurgy, diffusion experiments and more. (authors)

A Bayesian sampling strategy for hazardous waste site characterization

International Nuclear Information System (INIS)

Skalski, J.R.

1987-12-01

Prior knowledge based on historical records or physical evidence often suggests the existence of a hazardous waste site. Initial surveys may provide additional or even conflicting evidence of site contamination. This article presents a Bayes sampling strategy that allocates sampling at a site using this prior knowledge. This sampling strategy minimizes the environmental risks of missing chemical or radionuclide hot spots at a waste site. The environmental risk is shown to be proportional to the size of the undetected hot spot or inversely proportional to the probability of hot spot detection. 12 refs., 2 figs

Conserving energy by eliminating waste

Energy Technology Data Exchange (ETDEWEB)

Jones, N. H.

1979-07-01

Some ways in which energy is wasted in industry are discussed and the losses involved are quantified. Reference is made to a particular loss in annealing furnaces; wasted energy in factory and lighting systems; heat generated by motors and lighting and by such processes as welding; unlagged hot pipework and most hot processes; and poor building envelope features. It is concluded that an industry should declare its intention of conservation at the highest possible level, identify conservation as a manufacturing target, and invest the responsibility in people for whom it is a full-time activity. (MCW)

Encapsulation of tritiated wastes in hot laboratories

International Nuclear Information System (INIS)

Hayet, L.; Bourdinaud, P.

1982-06-01

From 1962 to 1976, one of the laboratories called ''Cellule 4'', was equipped to produce large quantities of tritium. The decisions taken in June 1976 to redirect the activities of the Radioisotopes Department included the suspension of the tritium activities of ''Cellule 4'' at the Saclay Nuclear Research Center. From 1976 to 1981, many CEA departments contributed to the design and implementation of a packaging procedure for tritiated wastes resulting from the dismantling of ''Cellule 4''. A classification into three groups was adopted for this purpose. (1) Packaging in a welded stainless steel container. (2) Packaging in a reinforced concrete shell lined internally with a thick coat of epoxy resin charged with sand. (3) Packaging in a reinforced concrete shell lined internally with a thin coat of epoxy resin. The dismantling operations were carried out in three phases. (1) (2 1/2 months): the T activity remained unchanged during this period and waste selection was carried out. (2) (2 1/2 months): waste dismantling and packaging of groups (2) and (3) were carried out in this phase. Activity decayed rapidly. (3) (2 months): the work performed included the loosening and cutting of the general structures [fr

Application and research of special waste plasma disposal technology

International Nuclear Information System (INIS)

Lan Wei

2007-12-01

The basic concept of plasma and the principle of waste hot plasma disposal technology are simply introduced. Several sides of application and research of solid waste plasma disposal technology are sumed up. Compared to the common technology, the advantages of waste hot plasma disposal technology manifest further. It becomes one of the most prospective and the most attended high tech disposal technology in particular kind of waste disposal field. The article also simply introduces some experiment results in Southwest Institute of Physics and some work on the side of importation, absorption, digestion, development of foreign plasma torch technology and researching new power sources for plasma torch. (authors)

Potential of district cooling in hot and humid climates

DEFF Research Database (Denmark)

Dominkovic, Dominik Franjo; Rashid, K. A. Bin Abdul; Romagnoli, A.

2017-01-01

Efficiently utilizing energy that is currently being wasted can significantly increase energy efficiency of the system, as well as reduce the carbon footprint. In hot climates with large cooling demands, excess waste heat can be utilized via absorption chillers to generate cold. Moreover, cold from...... liquefied natural gas gasification process can further provide energy source for meeting the cold demand. In order to connect the large sources of waste heat and cold energy with customers demanding the cold, a significant investment in district cooling grid is a necessity. In order to deal...

Treatment of rod shaped intermediate active waste

International Nuclear Information System (INIS)

Graf, A.; Blase, F.; Dirks, F.; Valencia, L.

2002-01-01

The Central Decontamination Operation Department (HDB) of the Research Center Karlsruhe operates facilities for the disposal of radioactive waste. In general, their objective is to reduce the volume of the radioactive waste and to obtain waste products suitable for repository storage. One of the central facilities of the HDB is the intermediate level waste (ILW) scrapping facility which processes intermediate level waste. Since the ILW scrapping facility was not large enough to handle radioactive waste coming from the dismantling and operating of nuclear facilities, HDB expanded and built a larger hot cell. It contains a hydraulically driven metal cutter with a guiding channel and a high pressure compactor. A major task in the hot cell of the ILW scrapping facility is disposing of fuel boxes. These are cut in pieces and scrapped, which is a unique technique in Germany for fuel box disposal. HDB's experiences in disposing of radioactive waste in the ILW scrapping facility will described in detail, with special emphasis on the handling of rod shaped components. (author)

Dismantling of an alpha contaminated hot cell at the Marcoule Pilot Plant

International Nuclear Information System (INIS)

Tachon, M.

1988-01-01

For the remodeling of Marcoule Pilot Plant, the cell 82: old unit for plutonium solution purification by extraction, was dismantled. About 42 tons of wastes were evacuated. Some wastes wen decontaminated by mechanical means other wastes with higher residual activity were stored for subsequent processing. The operation shows that dismantling of a hot cell is possible even if incorporated in an operating plant [fr

An approach for the reasonable TRU waste management in NUCEF

International Nuclear Information System (INIS)

Mineo, H.; Dojiri, S.; Takeshita, I.; Tsujino, T.; Matsumura, T.; Nishizawa, I.; Sugikawa, S.

1995-01-01

The Nuclear Fuel Cycle Safety Engineering Research Facility (NUCEF) has started its hot operation at the beginning of 1995, where TRU (transuranic) elements are used. The management of TRU waste arisen in the facility is very important issue. Liquid and solid wastes containing TRU elements are generated mainly from the Fuel Treatment System for critical experiments and from the researches of reprocessing process and TRU waste management for reprocessing plants using hot cells and glove-boxes. The TRU waste management in NUCEF is based on the classification of waste, and is to maximize the recycle of reagents and the reuse of TRU elements separated from the waste, as well as to reduce the waste volume and to lower the risk of waste by advanced separation and solidification. In the future, the separation and solidification of TRU elements in the tanks of liquid waste, and the classification and discrimination of solid wastes, will be carried out applying the outcomes of the development by the researches in NUCEF. (authors)

Aspen Plus® and economic modeling of equine waste utilization for localized hot water heating via fast pyrolysis.

Science.gov (United States)

Hammer, Nicole L; Boateng, Akwasi A; Mullen, Charles A; Wheeler, M Clayton

2013-10-15

Aspen Plus(®) based simulation models have been developed to design a pyrolysis process for on-site production and utilization of pyrolysis oil from equine waste at the Equine Rehabilitation Center at Morrisville State College (MSC). The results indicate that utilization of all the available waste from the site's 41 horses requires a 6 oven dry metric ton per day (ODMTPD) pyrolysis system but it will require a 15 ODMTPD system for waste generated by an additional 150 horses at the expanded area including the College and its vicinity. For this a dual fluidized bed combustion reduction integrated pyrolysis system (CRIPS) developed at USDA's Agricultural Research Service (ARS) was identified as the technology of choice for pyrolysis oil production. The Aspen Plus(®) model was further used to consider the combustion of the produced pyrolysis oil (bio-oil) in the existing boilers that generate hot water for space heating at the Equine Center. The model results show the potential for both the equine facility and the College to displace diesel fuel (fossil) with renewable pyrolysis oil and alleviate a costly waste disposal problem. We predict that all the heat required to operate the pyrolyzer could be supplied by non-condensable gas and about 40% of the biochar co-produced with bio-oil. Techno-economic Analysis shows neither design is economical at current market conditions; however the 15 ODMTPD CRIPS design would break even when diesel prices reach $11.40/gal. This can be further improved to $7.50/gal if the design capacity is maintained at 6 ODMTPD but operated at 4950 h per annum. Published by Elsevier Ltd.

Giving waste a hot time [incineration technology

International Nuclear Information System (INIS)

Cruickshank, Andrew.

1986-01-01

High temperature incineration technology, as an effective way of managing both solid wastes and sludges, is described. The process, developed by the Belgian Nuclear Research Centre, is detailed. (U.K.)

HOTLAB: European hot laboratories research capacities and needs. Plenary meeting 2004

International Nuclear Information System (INIS)

Oberlaender, B.C.; Jenssen, H.K.

2005-01-01

The report presents proceedings from the 2004 annual HOTLAB plenary meeting at Halden and Kjeller, Norway. The goal of the yearly plenary meeting was to: Exchange experience on analytical methods, their implementation in hot cells, the methodologies used and their application in nuclear research. Share experience on common infrastructure exploitation matters such as remote handling techniques, safety features, QA-certification, waste handling, etc. Promote normalisation and co-operation, e.g. by looking at mutual complementarities. Prospect present and future demands from the nuclear industry and to draw strategic conclusions regarding further needs. The main themes of the five topical oral sessions of the Halden plenary meeting cover: Work package leaders report and specific papers, presentation of PIE facility databases, i.e. one worldwide (IAEA) and one inside the European communities. Reports from present and future needs and on nuclear transports. Refabrication and instrumentation: Available equipment, technical characteristics such as fabrication procedures, hot-cell compatibility, and practical experiences. Post irradiation examination: Updated and new remote techniques and methodologies, new materials such as inert matrix fuels, spallation sources and neutron absorber materials. Refurbishment and decommissioning: reports on refurbishment and decommissioning of PIE facilities. Waste and transport: Hot laboratory waste characteristics and handling, spent fuel research. Several posters are presented

Hot cell design in the vitrification plant China

International Nuclear Information System (INIS)

Jiang Yubo; Wang Guangkai; Zhang Wei; Liang Runan; Dou Yuan

2015-01-01

In the area of reprocessing and radioactive waste management, gloveboxes and cells are a kind of non-standard equipments providing an isolated room to operate radioactive material inside, while the operator outside with essential biological shield and protection. The hot cell is a typical one, which could handle high radioactive material with various operating means and tight enclosure. The dissertation is based on Vitrification Plant China, a cooperation project between China and Germany. For the sino-western difference in design philosophy, it was presented how to draft an acceptable design proposal of applicable huge hot cells by analysing the design requirements, such as radioprotection, observation, illumination, remote handling, transportation, maintenance and decontamination. The construction feasibility of hot cells was also approved. Thanks to 3D software Autodesk Inventor, digital hot cell was built to integrate all the interfaces inside, which validated the design by checking the mechanical interference. (author)

Remote characterization of mixed waste by infrared spectroscopy: Fiscal year 1995 report

International Nuclear Information System (INIS)

Rebagay, T.V.; Reich, F.R.; Dodd, D.A.; Lopez, T.; Watts, J.K.; Cash, R.J.

1995-09-01

This report summarizes development work completed in FY 1995 on near-infrared (NIR) and Fourier Transform Infrared (FTIR) spectroscopy of determining the moisture content and solid species present in Hanford Site high-level waste. In prior fiscal years, the main emphasis was on identification of cyanide species that might be present in the ferrocyanide waste tanks, but the present thrust has been more focused on determining the moisture content of the waste. If sufficient moisture is present in the waste, propagating reactions in reactive waste are precluded, regardless of its fuel content. A prototype hot cell NIR moisture probe is now ready for hot cell deployment to sense moisture contents and homogeneity in tank waste samples

On the road to WIPP: Or remote packaging of transuranic waste

International Nuclear Information System (INIS)

Ledbetter, J.M.; Field, L.R.

1994-01-01

At the Los Alamos National Laboratory (LANL) Hot Cell facility, highly productive programs in reactor research spanning three decades have generated appreciable quantities of legacy waste. Hot cell capability had become virtually useless due to the storage of this waste. As a result of concentrated efforts by LANL staff, in cooperation with Westinghouse Waste Isolation Pilot Plant (WIPP), a solution was arrived at that allowed the facility to become productive once again. Equipment has been designed and fabricated to remotely handle 55-gal. waste drums, load waste canisters, perform canister weld closure, leak test welds, grapple the waste canister and transport the canister to an interim storage site. It is our contention that the technology and acquired equipment produced from this effort should be used to further benefit other DOE sites

Assessment of PCBs and exposure risk to infants in breast milk of primiparae and multiparae mothers in an electronic waste hot spot and non-hot spot areas in Ghana.

Science.gov (United States)

Asamoah, Anita; Essumang, David Kofi; Muff, Jens; Kucheryavskiy, Sergey V; Søgaard, Erik Gydesen

2018-01-15

The aim of the study was to assess the levels of PCBs in the breast milk of some Ghanaian women at suspected hotspot and relatively non-hotspot areas and to find out if the levels of these PCBs pose any risk to the breastfed infants. A total of 128 individual human breast milk were sampled from both primiparae and multiparae mothers. The levels of PCBs in the milk samples were compared. Some of these mothers (105 individuals) work or reside in and around Agbogbloshie (hot-spot), the largest electric and electronic waste dump and recycling site in Accra, Ghana. Others (23 donor mothers) also reside in and around Kwabenya (non-hotspot) which is a mainly residential area without any industrial activities. Samples were analyzed using GC-MS/MS. The total mean levels and range of Σ 7 PCBs were 3.64ng/glipidwt and ˂LOD-29.20ng/glipidwt, respectively. Mean concentrations from Agbogbloshie (hot-spot area) and Kwabenya (non-hotspot areas) were 4.43ng/glipidwt and 0.03ng/glipidwt, respectively. PCB-28 contributed the highest of 29.5% of the total PCBs in the milk samples, and PCB-101 contributed the lowest of 1.74%. The estimated daily intake of PCBs and total PCBs concentrations in this work were found to be lower as compared to similar studies across the world. The estimated hazard quotient using Health Canada's guidelines threshold limit of 1μg/kgbw/day showed no potential health risk to babies. However, considering minimum tolerable value of 0.03μg/kgbw/day defined by the Agency for Toxic Substances and Disease Registry (ATSDR), the values of some mothers were found to be at the threshold limit. This may indicate a potential health risk to their babies. Mothers with values at the threshold levels of the minimum tolerable limits are those who work or reside in and around the Agbogbloshie e-waste site. Copyright © 2017 Elsevier B.V. All rights reserved.

HOTLAB: European hot laboratories research and capacities and needs. Plenary meeting 2004

Energy Technology Data Exchange (ETDEWEB)

Oberlaender, B.C.; Jenssen, H.K. (ed.)

2005-01-01

The report presents proceedings from the 2004 annual HOTLAB plenary meeting at Halden and Kjeller, Norway. The goal of the yearly plenary meeting was to: Exchange experience on analytical methods, their implementation in hot cells, the methodologies used and their application in nuclear research. Share experience on common infrastructure exploitation matters such as remote handling techniques, safety features, QA-certification, waste handling, etc. Promote normalisation and co-operation, e.g. by looking at mutual complementarities. Prospect present and future demands from the nuclear industry and to draw strategic conclusions regarding further needs. The main themes of the five topical oral sessions of the Halden plenary meeting cover: Work package leaders report and specific papers, presentation of PIE facility databases, i.e. one worldwide (IAEA) and one inside the European communities. Reports from present and future needs and on nuclear transports. Refabrication and instrumentation: Available equipment, technical characteristics such as fabrication procedures, hot-cell compatibility, and practical experiences. Post irradiation examination: Updated and new remote techniques and methodologies, new materials such as inert matrix fuels, spallation sources and neutron absorber materials. Refurbishment and decommissioning: reports on refurbishment and decommissioning of PIE facilities. Waste and transport: Hot laboratory waste characteristics and handling, spent fuel research. Several posters are presented.

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.)

WIPP conceptual design report. Addendum J. Support equipment in the high level waste facility of the Waste Isolation Pilot Plant

International Nuclear Information System (INIS)

Rieb, M.J.; Foley, R.S.

1977-04-01

The Aerojet Manufacturing Company (AMCO) received a contract in November 1976 to provide consulting services in assisting Holmes and Narver, Incorporated with the conceptual designs, cost estimates, and schedules of equipment used to handle waste casks, to decontaminate waste canisters and to overpack damaged or highly contaminated waste canisters for the Waste Isolation Pilot Plant (WIPP). Also, the layout of the hot cell in which canister handling, overpack and decontamination takes place was to be reviewed along with the time and motion study of the cell operations. This report has been prepared to present the results of the efforts and contains all technical and planning data developed during the program. The contents of this report are presented in three sections: (1) comments on the existing design criteria, equipment conceptual designs, hot cell design and time and motion studies of projected hot cell activities; (2) design descriptions of the equipment concepts and justification for varying from the existing concept (if a variation occurred). Drawings of each concept are provided in Appendix A. These design descriptions and drawings were used as the basis for the cost estimates; and (3) schedule projections and cost estimates for the equipment described in Section 2. Detail cost estimate backup data is provided in Appendix B

A summary of available information on ferrocyanide tank wastes

International Nuclear Information System (INIS)

Burger, L.L.; Strachan, D.M.; Reynolds, D.A.; Schulz, W.W.

1991-10-01

Ferrocyanide wastes were generated at the Hanford site during the mid to late 1950s to make more tank space available for the storage of high level nuclear waste. The ferrocyanide process was developed as a method of removing 137 Cs from existing waste solutions and from process solutions that resulted from the recovery of valuable uranium in waste tanks. During the coarse of the research associated with the ferrocyanide process, it was discovered that ferrocyanide materials when mixed with NaNO 3 and/or NaNO 2 exploded. This chemical reactivity became an issue in the 1980s when the safety associated with the storage of ferrocyanide wastes in Hanford tanks became prominent. These safety issues heightened in the late 1980s and led to the current scrutiny of the safety associated with these wastes and the current research and waste management programs. Over the past three years, numerous explosive test have been carried out using milligram quantities of cyanide compounds. These tests provide information on the nature of possible tank reactions. On heating a mixture of ferrocyanide and nitrate or nitrite, an explosive reaction normally begins at about 240 degrees C, but may occur well below 200 degrees C in the presence of catalysts or organic compounds that may act as initiators. The energy released is highly dependent on the course of the reaction. Three attempts to model hot spots in local areas of the tanks indicate a very low probability of having a hot spot large enough and hot enough to be of concern. The main purpose of this document is to inform the members of the Tank Waste Science Panel of the background and issues associated with the ferrocyanide wastes. Hopefully, this document fulfills similar needs outside of the framework of the Tank Waste Science Panel. 50 refs., 9 figs., 7 tabs

EDF experience with open-quotes hot spotclose quotes management

International Nuclear Information System (INIS)

Guio, J.M. de

1995-01-01

During the past few years, open-quotes hot spotsclose quotes due to the presence of particles of metal activated during their migration through the reactor core, have been detected at several French pressurized water reactor (PWR) units. These open-quotes hot spots,close quotes which generate very high dose rates (from about 10 Gy/h to 200 G/h) are a significant factor in increase occupational exposures during outrates. Of particular concern are the difficult cases which prolong outage duration and increase the volume of radiological waste. Confronted with this situation, Electricite de France (EDF) has set up a national research group, as part of its ALARA program, to establish procedures and techniques to avoid, detect, and eliminate of hot spots. In particular, specific processes have been developed to eliminate these hot spots which are most costly in terms of occupational exposure due to the need for reactor maintenance. This paper sets out the general approach adopted at EDF so far to cope with the problem of hot spots, illustrated by experience at Blayais 3 and 4

Waste Treatment Plant - 12508

Energy Technology Data Exchange (ETDEWEB)

Harp, Benton; Olds, Erik [US DOE (United States)

2012-07-01

The Waste Treatment Plant (WTP) will immobilize millions of gallons of Hanford's tank waste into solid glass using a proven technology called vitrification. The vitrification process will turn the waste into a stable glass form that is safe for long-term storage. Our discussion of the WTP will include a description of the ongoing design and construction of this large, complex, first-of-a-kind project. The concept for the operation of the WTP is to separate high-level and low-activity waste fractions, and immobilize those fractions in glass using vitrification. The WTP includes four major nuclear facilities and various support facilities. Waste from the Tank Farms is first pumped to the Pretreatment Facility at the WTP through an underground pipe-in-pipe system. When construction is complete, the Pretreatment Facility will be 12 stories high, 540 feet long and 215 feet wide, making it the largest of the four major nuclear facilities that compose the WTP. The total size of this facility will be more than 490,000 square feet. More than 8.2 million craft hours are required to construct this facility. Currently, the Pretreatment Facility is 51 percent complete. At the Pretreatment Facility the waste is pumped to the interior waste feed receipt vessels. Each of these four vessels is 55-feet tall and has a 375,000 gallon capacity, which makes them the largest vessels inside the Pretreatment Facility. These vessels contain a series of internal pulse-jet mixers to keep incoming waste properly mixed. The vessels are inside the black-cell areas, completely enclosed behind thick steel-laced, high strength concrete walls. The black cells are designed to be maintenance free with no moving parts. Once hot operations commence the black-cell area will be inaccessible. Surrounded by black cells, is the 'hot cell canyon'. The hot cell contains all the moving and replaceable components to remove solids and extract liquids. In this area, there is ultrafiltration

Remote Robotic Cleaning System for Contaminated Hot-Cell Floor

International Nuclear Information System (INIS)

Kim, Ki Ho; Park, Jang Jin; Yang, Myung S.; Kwon, Hyo Kjo

2005-01-01

The M6 hot-cell of the Irradiated Material Examination Facility at the Korea Atomic Energy Research Institute (KAERI) has been contaminated with spent fuel debris and other radioactive waste due to the DUPIC nuclear fuel development processes. As the hot-cell is active, direct human workers' access, even with protection, to the in-cell is not possible because of the nature of the high radiation level of the spent PWR fuel. A remote robotic cleaning system has been developed for use in a highly radioactive environment of the M6 hot-cell. The remote robotic cleaning system was designed to completely eliminate human interaction with hazardous radioactive contaminants. This robotic cleaning system was also designed to remove contaminants or contaminated smears placed or fixed on the floor of the M6 hot-cell by mopping it in a remote manner. The environmental, functional and mechanical design considerations, control system and capabilities of the developed remote robotic cleaning system are presented

Closed bioregenerative life support systems: Applicability to hot deserts

Science.gov (United States)

Polyakov, Yuriy S.; Musaev, Ibrahim; Polyakov, Sergey V.

2010-09-01

Water scarcity in hot deserts, which cover about one-fifth of the Earth's land area, along with rapid expansion of hot deserts into arable lands is one of the key global environmental problems. As hot deserts are extreme habitats characterized by the availability of solar energy with a nearly complete absence of organic life and water, space technology achievements in designing closed ecological systems may be applicable to the design of sustainable settlements in the deserts. This review discusses the key space technology findings for closed biogenerative life support systems (CBLSS), which can simultaneously produce food, water, nutrients, fertilizers, process wastes, and revitalize air, that can be applied to hot deserts. Among them are the closed cycle of water and the acceleration of the cycling times of carbon, biogenic compounds, and nutrients by adjusting the levels of light intensity, temperature, carbon dioxide, and air velocity over plant canopies. Enhanced growth of algae and duckweed at higher levels of carbon dioxide and light intensity can be important to provide complete water recycling and augment biomass production. The production of fertilizers and nutrients can be enhanced by applying the subsurface flow wetland technology and hyper-thermophilic aerobic bacteria for treating liquid and solid wastes. The mathematical models, optimization techniques, and non-invasive measuring techniques developed for CBLSS make it possible to monitor and optimize the performance of such closed ecological systems. The results of long-duration experiments performed in BIOS-3, Biosphere 2, Laboratory Biosphere, and other ground-based closed test facilities suggest that closed water cycle can be achieved in hot-desert bioregenerative systems using the pathways of evapotranspiration, condensation, and biological wastewater treatment technologies. We suggest that the state of the art in the CBLSS design along with the possibility of using direct sunlight for

Production of an impermeable composite of irradiated graphite and glass by hot isostatic pressing as a long term leach resistant waste form

International Nuclear Information System (INIS)

Fachinger, Johannes; Muller, Walter; Marsat, Eric; Grosse, Karl-Heinz; Seemann, Richard; Scales, Charlie; Easton, Michael Mark; Anthony Banford

2013-01-01

Around 250,000 tons of irradiated graphite (i-graphite) exists worldwide and can be considered as a current waste or future waste stream. The largest national i-graphite inventory is located in UK (∼ 100,000 tons) with significant quantities also in Russia and France [5]. Most of the i-graphite remains in the cores of shutdown nuclear reactors including the MAGNOX type in UK and the UNGG in France. Whilst there are still operational power reactors with graphite cores, such as the Russian RBMKs and the AGRs in UK, all of them will reach their end of life during the next two decades. The most common reference waste management option of i-graphite is a wet or dry retrieval of the graphite blocks from the reactor core and the grouting of these blocks in a container without further conditioning. This produces large waste package volumes because the encapsulation capacity of the grout is limited and large cavities in the graphite blocks could reduce the packing densities. Packing densities from 0.5 to 1 tons per cubic meter have been assumed for grouting solutions. Furthermore the grout is permeable. This could over time allow the penetration of aqueous phases into the waste block and a potential dissolution and release of radionuclides. As a result particularly highly soluble radionuclides may not be retained by the grout. Vitrification could present an alternative, however a similar waste package volume increase may be expected since the encapsulation capacity of glass is potentially similar to or worse than that of grout. FNAG has developed a process for the production of a graphite-glass composite material called Impermeable Graphite Matrix (IGM) [3]. This process is also applicable to irradiated graphite which allows the manufacturing of an impermeable material without volume increase. Crushed i-graphite is mixed with 20 vol.% of glass and then pressed under vacuum at an elevated temperature in an axial hot vacuum press (HVP). The obtained product has zero or

Production of an impermeable composite of irradiated graphite and glass by hot isostatic pressing as a long term leach resistant waste form

Energy Technology Data Exchange (ETDEWEB)

Fachinger, Johannes; Muller, Walter [FNAG ZU Hanau, Hanau (Germany); Marsat, Eric [FNAG SAS Le Pont de Claix (France); Grosse, Karl-Heinz; Seemann, Richard [ALD Hanau (Germany); Scales, Charlie; Easton, Michael Mark [NNL, Workington (United Kingdom); Anthony Banford [NNL, Warrington (United Kingdom); University of Manchester, Manchester (United Kingdom)

2013-07-01

Around 250,000 tons of irradiated graphite (i-graphite) exists worldwide and can be considered as a current waste or future waste stream. The largest national i-graphite inventory is located in UK (∼ 100,000 tons) with significant quantities also in Russia and France [5]. Most of the i-graphite remains in the cores of shutdown nuclear reactors including the MAGNOX type in UK and the UNGG in France. Whilst there are still operational power reactors with graphite cores, such as the Russian RBMKs and the AGRs in UK, all of them will reach their end of life during the next two decades. The most common reference waste management option of i-graphite is a wet or dry retrieval of the graphite blocks from the reactor core and the grouting of these blocks in a container without further conditioning. This produces large waste package volumes because the encapsulation capacity of the grout is limited and large cavities in the graphite blocks could reduce the packing densities. Packing densities from 0.5 to 1 tons per cubic meter have been assumed for grouting solutions. Furthermore the grout is permeable. This could over time allow the penetration of aqueous phases into the waste block and a potential dissolution and release of radionuclides. As a result particularly highly soluble radionuclides may not be retained by the grout. Vitrification could present an alternative, however a similar waste package volume increase may be expected since the encapsulation capacity of glass is potentially similar to or worse than that of grout. FNAG has developed a process for the production of a graphite-glass composite material called Impermeable Graphite Matrix (IGM) [3]. This process is also applicable to irradiated graphite which allows the manufacturing of an impermeable material without volume increase. Crushed i-graphite is mixed with 20 vol.% of glass and then pressed under vacuum at an elevated temperature in an axial hot vacuum press (HVP). The obtained product has zero or

Method of processing radioactive metal wastes

International Nuclear Information System (INIS)

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

1980-01-01

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

Prolixe-prototype reprocessing unit for irradiating wastes contamined with alpha emitters

International Nuclear Information System (INIS)

Madic, C.; Sontag, R.

1987-01-01

A large number of hot cells are employed for research on nuclear fuel reprocessing and the production of isotope of transuranium elements. These activities generate solid wastes highly contaminated with alpha, beta, gamma emitters. The Prolixe hot cell was built in order to: 1/ reprocess the solid wastes contaminated with alpha, beta, gamma emitters produced in the Radiochemistry building: 2/ produce package wastes storable in shallow-ground disposal sites: 3/ develop a process sufficiently flexible to make it applicable to waste produced in other installations. The process is based on waste leaching after grinding. Depending on the type of wastes the leaching reactant will have a different composition 1/ nitric acid solution for cellulose waste: 2/ nitric solutions containing Ag(II) for other material. The complete process should achieve: 1/ a high waste volume reduction factor: 2/ the production of immobilized waste packages storage in shallow-ground disposal sites: 3/ the recycling of transuranium elements: 4/ the generation of a minimal volume of effluents. This process can be considered as an alternative process to incineration for the reprocessing of solid wastes highly contaminated with alpha, beta, gamma emitters

Application of SYNROC to high-level defense wastes

International Nuclear Information System (INIS)

Tewhey, J.D.; Hoenig, C.L.; Newkirk, H.W.; Rozsa, R.B.; Coles, D.G.; Ryerson, F.J.

1981-01-01

The SYNROC method for immobilization of high-level nuclear reactor wastes is currently being applied to US defense wastes in tank storage at Savannah River, South Carolina. The minerals zirconolite, perovskite, and hollandite are used in SYNROC D formulations to immobilize fission products and actinides that comprise up to 10% of defense waste sludges and coexisting solutions. Additional phase in SYNROC D are nepheline, the host phase for sodium; and spinel, the host for excess aluminum and iron. Up to 70 wt % of calcined sludge can be incorporated with 30 wt % of SYNROC additives to produce a waste form consisting of 10% nepheline, 30% spinel, and approximately 20% each of the radioactive waste-bearing phases. Urea coprecipitation and spray drying/calcining methods have been used in the laboratory to produce homogeneous, reactive ceramic powders. Hot pressing and sintering at temperatures from 1000 to 1100 0 C result in waste form products with greater than 97% of theoretical density. Hot isostatic pressing has recently been implemented as a processing alternative. Characterization of waste-form mineralogy has been done by means of XRD, SEM, and electron microprobe. Leaching of SYNROC D samples is currently being carried out. Assessment of radiation damage effects and physical properties of SYNROC D will commence in FY81

The design of hot laboratories

International Nuclear Information System (INIS)

1976-01-01

The need for specialized laboratories to handle radioactive substances of high activity has increased greatly due to the expansion of the nuclear power industry and the widespread use of radioisotopes in scientific research and technology. Such laboratories, which are called hot laboratories, are specially designed and equipped to handle radioactive materials of high activity, including plutonium and transplutonium elements. The handling of plutonium and transplutonium elements presents special radiation-protection and safety problems because of their high specific activity and high radiotoxicity. Therefore, the planning, design, construction and operation of hot laboratories must meet the stringent safety, containment, ventilation, shielding, criticality control and fire-protection requirements. The IAEA has published two manuals in its Safety Series, one on the safety aspects of design and equipment of hot laboratories (SS No.30) and the other on the safe handling of plutonium (SS No.39). The purpose of the symposium in Otaniemi was to collect information on recent developments in the safety features of hot laboratories and to review the present state of knowledge. A number of new developments have taken place as the result of growing sophistication in the philosophy of radiation protection as given in the ICRP recommendations (Report No.22) and in the Agency's basic safety standards (No.9). The topics discussed were safety features of planning and design, air cleaning, transfer and transport systems, criticality control, fire protection, radiological protection, waste management, administrative arrangements and operating experience

Immobilisation of Higher Activity Wastes from Nuclear Reactor Production of 99Mo

Directory of Open Access Journals (Sweden)

Martin W. A. Stewart

2013-01-01

Full Text Available A variety of intermediate- and low-level liquid and solid wastes are produced from reactor production of 99Mo using UAl alloy or UO2 targets and in principle can be collectively or individually converted into waste forms. At ANSTO, we have legacy acidic uranyl-nitrate-rich intermediate level waste (ILW from the latter, and an alkaline liquid ILW, a U-rich filter cake, plus a shorter lived liquid stream that rapidly decays to low-level waste (LLW standards, from the former. The options considered consist of cementitious products, glasses, glass-ceramics, or ceramics produced by vitrification or hot isostatic pressing for intermediate-level wastes. This paper discusses the progress in waste form development and processing to treat ANSTO’s ILW streams arising from 99Mo. The various waste forms and the reason for the process option chosen will be reviewed. We also address the concerns over adapting our chosen process for use in a hot-cell environment.

EDF experience with {open_quotes}hot spot{close_quotes} management

Energy Technology Data Exchange (ETDEWEB)

Guio, J.M. de [Blayais Nuclear Power Plant, St. Ciers (France)

1995-03-01

During the past few years, {open_quotes}hot spots{close_quotes} due to the presence of particles of metal activated during their migration through the reactor core, have been detected at several French pressurized water reactor (PWR) units. These {open_quotes}hot spots,{close_quotes} which generate very high dose rates (from about 10 Gy/h to 200 G/h) are a significant factor in increase occupational exposures during outrates. Of particular concern are the difficult cases which prolong outage duration and increase the volume of radiological waste. Confronted with this situation, Electricite de France (EDF) has set up a national research group, as part of its ALARA program, to establish procedures and techniques to avoid, detect, and eliminate of hot spots. In particular, specific processes have been developed to eliminate these hot spots which are most costly in terms of occupational exposure due to the need for reactor maintenance. This paper sets out the general approach adopted at EDF so far to cope with the problem of hot spots, illustrated by experience at Blayais 3 and 4.

Effect of Hot water and dilute acid pretreatment on the chemical properties of liquorice root

Directory of Open Access Journals (Sweden)

zahra takzare

2016-06-01

Full Text Available Abstract In this study, the liquorice root (Glycyrrhiza glabra that was extracted in the factory in Kerman province, pre-hydrolyzed and then chemical compositions (Extractives, Lignin content, Holocellulose percent, the hydrolysis process yield and weight loss of the waste was measured. Pre-hydrolysis process was done on the above mentioned waste by hot water, hot water followed by 0.5 percent sulfuric acid and also alone sulfuric acid with different concentrations (0.5, 1, 1.5 and 2 percent The samples were pre-hydrolyzed in hot water at 150 °C and 30, 60 and 90 minutes as well as in the mixture of hot water and 0.5 % sulfuric acid at 150 °C and 60 minutes and also in pure sulfuric acid, at 130 °C and at 60 minutes. The results showed that the pre-hydrolyzed treatment with hot water in 60 minutes had been favorable performance in the respect of weight loss, lignin content and holocellulose percent. Also, in the case of pre-treatment including sulfuric acid, 2% dose can be good selected option in term of maximum holocellulose percent and minimum lignin content so that it can be suggested to produce higher value-added products such as bioethanol from licorice root bid.

Characterization of 618-11 solid waste burial ground, disposed waste, and description of the waste generating facilities

International Nuclear Information System (INIS)

Hladek, K.L.

1997-01-01

) buried in three rows in the northeast comer. In addition, five eight-foot diameter caissons are located at the west end of the center row of the drum storage units. Initially, wastes disposed to the caissons and drum storage units were from the 325 and 327 building hot cells. Later, a small amount of remote-handled (RH) waste from the 309 building Plutonium Recycle Test Reactor (PRTR) cells, and the newly built 324 building hot cells, was disposed at the site

Characterization of 618-11 solid waste burial ground, disposed waste, and description of the waste generating facilities

Energy Technology Data Exchange (ETDEWEB)

Hladek, K.L.

1997-10-07

) buried in three rows in the northeast comer. In addition, five eight-foot diameter caissons are located at the west end of the center row of the drum storage units. Initially, wastes disposed to the caissons and drum storage units were from the 325 and 327 building hot cells. Later, a small amount of remote-handled (RH) waste from the 309 building Plutonium Recycle Test Reactor (PRTR) cells, and the newly built 324 building hot cells, was disposed at the site.

Commercial and ERDA waste packaging criteria: possible similarities and differences

International Nuclear Information System (INIS)

Lowrie, B.

1977-01-01

The schedule calls for hot operation of two waste repositories by the end of 1985, and these facilities will have to be licensed. This licensing requirement sets the commercial program apart from the ERDA defense waste program. Packaging criteria are discussed for commercial and ERDA wastes. The different NRC, DOT, and EPA criteria are considered

Hot laboratory in Saclay. Equipment and radio-metallurgy technique of the hot lab in Saclay. Description of hot cell for handling of plutonium salts. Installation of an hot cell

International Nuclear Information System (INIS)

Bazire, R.; Blin, J.; Cherel, G.; Duvaux, Y.; Cherel, G.; Mustelier, J.P.; Bussy, P.; Gondal, G.; Bloch, J.; Faugeras, P.; Raggenbass, A.; Raggenbass, P.; Fufresne, J.

1959-01-01

Describes the conception and installation of the hot laboratory in Saclay (CEA, France). The construction ended in 1958. The main aim of this laboratory is to examine fuel rods of EL2 and EL3 as well as nuclear fuel studies. It is placed in between both reactors. In a first part, the functioning and specifications of the hot lab are given. The different hot cells are described with details of the ventilation and filtration system as well as the waste material and effluents disposal. The different safety measures are explained: description of the radiation protection, decontamination room and personnel monitoring. The remote handling equipment is composed of cutting and welding machine controlled with manipulators. Periscopes are used for sight control of the operation. In a second part, it describes the equipment of the hot lab. The unit for an accurate measurement of the density of irradiated uranium is equipped with an high precision balance and a thermostat. The equipment used for the working of irradiated uranium is described and the time length of each operation is given. There is also an installation for metallographic studies which is equipped with a manipulation bench for polishing and cleaning surfaces and a metallographic microscope. X-ray examination of uranium pellets will also be made and results will be compared with those of metallography. The last part describes the hot cells used for the manipulation of plutonium salts. The plutonium comes from the reprocessing plant and arrived as a nitric solution. Thus these cells are used to study the preparation of plutonium fluorides from nitric solution. The successive operations needed are explained: filtration, decontamination and extraction with TBP, purification on ion exchangers and finally formation of the plutonium fluorides. Particular attention has been given to the description of the specifications of the different gloveboxes and remote handling equipment used in the different reaction steps and

Method of processing radioactive wastes

International Nuclear Information System (INIS)

Nomura, Ichiro; Hashimoto, Yasuo.

1984-01-01

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

Small-scale hot facility for reprocessing and alpha research

International Nuclear Information System (INIS)

Abdel-Rassoul, A.A.

1976-01-01

The experimental hot facility at Inchas is planned for research activities related to the decontamination of radioactive wastes, analytical chemistry of alpha emitters and chemical treatment of spent UO 2 -Mg fuel samples. The design concept permits safe handling of source materials with radioactivity levels up to 10000Ci. The laboratory includes a reception area, process hall, a number of research laboratories and other facilities for chemical and physical analysis, nuclear measurements and health physics control. The radioactive waste management plant allows for control and decontamination of intermediate- and low-level laboratory effluents. Fixation of radioactive residues will be carried out in the sludge immobilization plant. High-level fission-product waste liquors are subject to preconcentration and transformation to a glassy matrix before ultimate storage. (author)

Design and management of hot-laboratories

International Nuclear Information System (INIS)

1976-09-01

This document is a manual for the design and management of hot-laboratories. It is composed of three parts. The first part is devoted to the design of hot-laboratories. Items included here are; conceptual design; many regulations which must be considered at design stage; design of cave and its shielding; and the design of building, ventilation, and draining. Many examples of specific designs are presented by figures and photographs. The second part is concerned with the methods of operation management. Organizational structure, scheduling of operation, process management, and regulatory problems are discussed with some examples. Technological problems associated with the operation of a hot laboratory (e.g., manipulator, transfer machine, maintenance, and decontamination) are also discussed based on the authors' experiences. An example of the operation manual is presented for reference. The third part is devoted to the safety management and the training of personnel. The regulations by law are briefly explained. Most of this part is devoted to the problem of monitoring radio-activity. Monitoring of control areas, radio-active wastes, and personal dosage is discussed together with many other specific monitoring problems. As for training, the purpose and the present status are explained. (Aoki, K.)

Analysis of TRU waste for RCRA-listed elements

International Nuclear Information System (INIS)

Mahan, C.; Gerth, D.; Yoshida, T.

1996-01-01

Analytical methods for RCRA listed elements on Portland cement type waste have been employed using both microwave and open hot plate digestions with subsequent analysis by inductively coupled plasma atomic emission spectroscopy (ICP-AES), inductively coupled plasma mass spectrometry (ICP-AES), inductively coupled plasma mass spectrometry (ICP-MS), graphite furnace atomic absorption (GFAA) and cold vapor atomic absorption and fluorescence (CVAA/CVAFS). Four different digestion procedures were evaluated including an open hot plate nitric acid digestion, EPA SW-846 Method 3051, and 2 methods using modifications to Method 3051. The open hot plate and the modified Method 3051, which used aqua regia for dissolution, were the only methods which resulted in acceptable data quality for all 14 RCRA-listed elements. Results for the nitric acid open hot plate digestion were used to qualify the analytical methods for TRU waste characterization, and resulted in a 99% passing score. Direct chemical analysis of TRU waste is being developed at Los Alamos National Laboratory in an attempt to circumvent the problems associated with strong acid digestion methods. Technology development includes laser induced breakdown spectroscopy (LIBS), laser ablation inductively coupled plasma mass spectrometry (LA-ICPMS), dc arc CID atomic emission spectroscopy (DC-AES), and glow discharge mass spectrometry (GDMS). Analytical methods using the Portland cement matrix are currently being developed for each of the listed techniques. Upon completion of the development stage, blind samples will be distributed to each of the technology developers for RCRA metals characterization

E-waste recycling: where does it go from here?

Science.gov (United States)

Zhang, Kai; Schnoor, Jerald L; Zeng, Eddy Y

2012-10-16

E-waste recycling has become a hotly debated global issue. This study, using China as a case study, analyzes the environmental, economic, and social implications of e-waste recycling in the developing world. More practical approaches, taking into account local economic and social conditions and the principles of Extended Producer Responsibility, are recommended to alleviate the increasing environmental disruption from improper e-waste disposal.

Hot Deformation Behavior of Hot-Extruded AA7175 Through Hot Torsion Tests.

Science.gov (United States)

Lee, Se-Yeon; Jung, Taek-Kyun; Son, Hyeon-Woo; Kim, Sang-Wook; Son, Kwang-Tae; Choi, Ho-Joon; Oh, Sang-Ho; Lee, Ji-Woon; Hyun, Soong-Keun

2018-03-01

The hot deformation behavior of hot-extruded AA7175 was investigated with flow curves and processing maps through hot torsion tests. The flow curves and the deformed microstructures revealed that dynamic recrystallization (DRX) occurred in the hot-extruded AA7175 during hot working. The failure strain was highest at medium temperature. This was mainly influenced by the dynamic precipitation of fine rod-shaped MgZn2. The processing map determined the optimal deformation condition for the alloy during hot working.

Decontamination of hot cells K-1, K-3, M-1, M-3, and A-1, M-Wing, Building 200: Project final report Argonne National Laboratory-East

International Nuclear Information System (INIS)

Cheever, C.L.; Rose, R.W.

1996-09-01

The purpose of this project was to remove radioactively contaminated materials and equipment from the hot cells, to decontaminate the hot cells, and to dispose of the radioactive waste. The goal was to reduce stack releases of Rn-220 and to place the hot cells in an emptied, decontaminated condition with less than 10 microSv/h (1 mrem/h) general radiation background. The following actions were needed: organize and mobilize a decontamination team; prepare decontamination plans and procedures; perform safety analyses to ensure protection of the workers, public, and environment; remotely size-reduce, package, and remove radioactive materials and equipment for waste disposal; remotely decontaminate surfaces to reduce hot cell radiation background levels to allow personnel entries using supplied air and full protective suits; disassemble and package the remaining radioactive materials and equipment using hands-on techniques; decontaminate hot cell surfaces to remove loose radioactive contaminants and to attain a less than 10 microSv/h (1 mrem/h) general background level; document and dispose of the radioactive and mixed waste; and conduct a final radiological survey

Conditioning of uranium-containing technological radioactive waste

International Nuclear Information System (INIS)

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

2006-01-01

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

Glass-Ceramic Waste Forms for Uranium and Plutonium Residues Wastes - 13164

International Nuclear Information System (INIS)

Stewart, Martin W.A.; Moricca, Sam A.; Zhang, Yingjie; Day, R. Arthur; Begg, Bruce D.; Scales, Charlie R.; Maddrell, Ewan R.; Hobbs, Jeff

2013-01-01

A program of work has been undertaken to treat plutonium-residues wastes at Sellafield. These have arisen from past fuel development work and are highly variable in both physical and chemical composition. The principal radiological elements present are U and Pu, with small amounts of Th. The waste packages contain Pu in amounts that are too low to be economically recycled as fuel and too high to be disposed of as lower level Pu contaminated material. NNL and ANSTO have developed full-ceramic and glass-ceramic waste forms in which hot-isostatic pressing is used as the consolidation step to safely immobilize the waste into a form suitable for long-term disposition. We discuss development work on the glass-ceramic developed for impure waste streams, in particular the effect of variations in the waste feed chemistry glass-ceramic. The waste chemistry was categorized into actinides, impurity cations, glass formers and anions. Variations of the relative amounts of these on the properties and chemistry of the waste form were investigated and the waste form was found to be largely unaffected by these changes. This work mainly discusses the initial trials with Th and U. Later trials with larger variations and work with Pu-doped samples further confirmed the flexibility of the glass-ceramic. (authors)

Fourteenth annual U.S. Department of Energy low-level radioactive waste management conference: Proceedings

Energy Technology Data Exchange (ETDEWEB)

NONE

1993-03-01

This document contains 69 papers that were presented at the Fourteenth Annual U.S. Department of Energy Low-Level Radioactive Waste Management Conference, November 18--20, 1992, in Phoenix, Arizona. The papers address the following general topics: (a) Performance Management Track: risk assessment; waste characterization; site characterization; facility design; groundwater modeling; monitoring and modeling; and regulatory requirements; (b) Technical Track: waste minimization; new technologies; international perspectives; licensing issues; hot topics; commercial storage; DOE storage; treatment technologies; and mixed waste; and (c) Institutional Track: status report; changes in orders, regulations, and guidance; regulatory compliance issues; communicating risk; hot topics; and storage impacts. Papers have been processed separately for inclusion on the data base.

Thermal Shielding of the Shock Absorber to a Seal of a Hot-cell Cask

Energy Technology Data Exchange (ETDEWEB)

Bang, K. S.; Lee, J. C.; Kim, K. Y.; Seo, C. S.; Seo, K. S. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2010-05-15

In order to safely transport the radioactive waste arising from the hot test of ACP(Advanced Spent Fuel Conditioning Process) a shipping package is required. Therefore KAERI is developing a shipping package to transport the radioactive waste arising in the ACPF during a hot test. Regulatory requirements for a Type B package are specified in the Korea MOST Act 2008-69, IAEA Safety Standard Series No. TS-R-1, and US 10 CFR Part. These regulatory guidelines classify the hot cell cask as a Type B package, and state that the Type B package for transporting radioactive materials should be able to withstand a test sequence consisting of a 9 m drop onto an unyielding surface, a 1 m drop onto a puncture bar, and a 30 minute fully engulfing fire. Greiner et al. performed a research on the thermal protection provided by shock absorbers by using CAFE computer code. This paper discusses the experimental approach used to simulate the response of the hot cell cask to fire in a furnace with chamber dimensions of 300 cm(W) x 400 cm(L) x 200 cm(H) by using a 1/2 scale model which was damaged by both a 9 m drop test and a 1 m puncture test

Treatment of radioactive wastes

International Nuclear Information System (INIS)

Machida, Chuji

1976-01-01

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

25th-year jubilee edition of the Hot Laboratory's annual report

International Nuclear Information System (INIS)

Bart, G.

1989-07-01

The topics discussed in this report concern the history of PSI's Hot Laboratory, nuclear fuel development performed there during the last twenty years, post irradiation examination work, basic research for a final repository of radioactive wastes, and neutron embrittlement experiments. 44 figs., 1 tab., 82 refs

Documentation associated with the shipping of Hot-Cell Waste from WESF 225-B to the 200W (218-W-3AE) burial grounds under shipment number RSR-37338

International Nuclear Information System (INIS)

PAWLAK, M.W.

1998-01-01

The purpose of this report is to compile the records generated during the Packaging and Shipping of WESF Hot-Cell Waste from the 225-B Facility to 200W (218-W-3AE) burial grounds. A total of six 55-gallon drums were packaged and shipped using the Chem-Nuc Cask in accordance with WHC-SD-TP-SARP-025, Rev.0 ''Safety Analysis Report for Packaging (Onsite) for Type B Material in the CNS-14-215H Cask''

Waste characterization for radioactive liquid waste evaporators at Argonne National Laboratory - West

International Nuclear Information System (INIS)

Christensen, B. D.

1999-01-01

Several facilities at Argonne National Laboratory - West (ANL-W) generate many thousand gallons of radioactive liquid waste per year. These waste streams are sent to the AFL-W Radioactive Liquid Waste Treatment Facility (RLWTF) where they are processed through hot air evaporators. These evaporators remove the liquid portion of the waste and leave a relatively small volume of solids in a shielded container. The ANL-W sampling, characterization and tracking programs ensure that these solids ultimately meet the disposal requirements of a low-level radioactive waste landfill. One set of evaporators will process an average 25,000 gallons of radioactive liquid waste, provide shielding, and reduce it to a volume of six cubic meters (container volume) for disposal. Waste characterization of the shielded evaporators poses some challenges. The process of evaporating the liquid and reducing the volume of waste increases the concentrations of RCIU regulated metals and radionuclides in the final waste form. Also, once the liquid waste has been processed through the evaporators it is not possible to obtain sample material for characterization. The process for tracking and assessing the final radioactive waste concentrations is described in this paper, The structural components of the evaporator are an approved and integral part of the final waste stream and they are included in the final waste characterization

Mechanistic evaluation of the effect of calcium carbide waste on ...

African Journals Online (AJOL)

Calcium Carbide Waste (CCW) was used as an alternative to traditional Portland cement mineral filler in hot mix asphalt concrete to rid its disposal problem. Its effect on mechanical properties of hot mix asphalt was assessed using the Marshall method of mix design. Using the optimum bitumen content determined from ...

Experiences from Refurbishment of Metallography Hot Cells and Application of a New Preparation Concept for Materialography Samples

International Nuclear Information System (INIS)

Oberlander, B. C.; Espeland, M.; Solum, N. O.

2001-01-01

After more than 30 years of operation the lead shielded metallography hot cells needed a basic renewal and modernisation not least of the specimen preparation equipment. Preparation in hot cells of radioactive samples for metallography and ceramography is challenging and time consuming. It demands a special design and quality of all in-cell equipment and skill and patience from the operator. Essentials in the preparation process are: simplicity and reliability of the machines, and a good quality, reproducibility and efficiency in performance. Desirable is process automation, flexibility and an alara amounto of radioactive waste produced per sample prepared. State of the art preparation equipment for materialography seems to meet most of the demands, however, it cannot be used in hot cells without modifications. Therefore. IFE and Struers in Copenhagen modified a standard model of a Strues precision cutting machine and a microprocessor controlled grinding and polishing machine for Hot Cell application. Hot cell utilisation of the microcomputer controlled grinding and polishing machine and the existing automatic dosing equipment made the task of preparing radioactive samples more attractive. The new grinding and polishing system for hot cells provides good sample preparation quality and reproductibility at reduced preparation time and reduced amount of contaminated waste produced per sample prepared. the sample materials examined were irradiated cladding materials and fuels

Nuclear waste: our radioactive hot potato

International Nuclear Information System (INIS)

Conselman, F.B.

1984-01-01

Nuclear industry inevitably produces nuclear waste, whose prudent, prompt and economic disposal is important to the national welfare. Technological problems of containment and isolation have apparently been solved. Underground or geologic disposal sites have the potential form permanent isolation, with salt, basalt, granite, shale, and tuff currently receiving principal attention as repository host rocks. Bedded salt deposits may offer the principal mechanical advantages, but in the northwestern United States the abundance of basalt at existing test sites has made it the subject of experimentation. However, psychological, political, and allegedly environmental obstructionism have stalled the process and virtually immobilized current construction. A program is suggested with the purpose of satisfying technical requirements for public protection while allaying the exaggerated fears of anti-nuclear factions

A systematic method to customize an efficient organic Rankine cycle (ORC) to recover waste heat in refineries

International Nuclear Information System (INIS)

Yu, Haoshui; Feng, Xiao; Wang, Yufei; Biegler, Lorenz T.; Eason, John

2016-01-01

Highlights: • Multiple waste heat streams in refinery are recovered for an ORC using a hot water intermediate. • WHCC and GCC are used to identify opportunities to save utility and/or upgrade waste heat. • The methods consider the interaction between the HEN and ORC in an integrated manner. - Abstract: Organic Rankine cycles (ORCs) convert low temperature waste heat into power. When there are multiple waste heat sources in a refinery, operability and safety considerations may make it more practical to use hot water as the medium to recover waste heat. The hot water stream can then release the heat to the organic working fluid in an ORC system. In this paper, how to customize an efficient ORC for a heat exchanger network (HEN) to optimally recover multiple strands of waste heat is investigated. Because the heat exchanger network structure, the hot water loop, and ORC system interact with each other, the coordination and synthesis of these systems ought to be considered simultaneously to maximize the energy performance. A methodology is proposed using the waste heat composite curve (WHCC) and grand composite curve (GCC) to diagnose inefficiencies in an existing heat exchanger network. In addition, the WHCC can be used to solve the problem of the tradeoff between waste heat quality and quantity recovered with an intermediate stream. WHCCs are classified into two types, and procedures for designing the recovery network for each type are presented while considering the interaction with working fluid selection. The methods proposed in this paper can help engineers diagnose problems with the original heat exchanger network, and determine the flowrate of hot water, the structure of the waste heat recovery network, the best working fluid and the operating conditions of ORC system in an integrated manner. The ideas are applied to an illustrative case study in collaboration with Sinopec. The case study shows the effectiveness of this method and compares different

DOE model conference on waste management and environmental restoration

International Nuclear Information System (INIS)

1990-01-01

Reports dealing with current topics in waste management and environmental restoration were presented at this conference in six sessions. Session 1 covered the ''Hot Topics'' including regulations and risk assessment. Session 2 dealt with waste reduction and minimization; session 3 dealt with waste treatment and disposal. Session 4 covered site characterization and analysis. Environmental restoration and associated technologies wee discussed in session 5 and 6. Individual papers have been cataloged separately

The development of a mobile hot cell facility for the conditioning of spent high activity radioactive sources

International Nuclear Information System (INIS)

Liebenberg, G.R.; Al-Mughrabi, M.

2010-01-01

The International Atomic Energy Agency (IAEA) Waste Technology Section with additional support from the U.S. National Nuclear Security Administration (NNSA) through the IAEA Nuclear Security Fund has funded the design, fabrication, evaluation, and testing of a portable hot cell intended to address the problem of disused Spent High Activity Radioactive Sources (SHARS) in obsolete irradiation devices such as teletherapy heads and dry irradiators. The project is initially targeting the African continent but expected soon to expand to Latin America and Asia. This hot cell allows source removal, characterization, consolidation, repackaging in modern storage shields, and secure storage of high risk SHARS at national radioactive waste storage facilities. (authors)

Hot isostatically-pressed aluminosilicate glass-ceramic with natural crystalline analogues for immobilizing the calcined high-level nuclear waste at the Idaho Chemical Processing Plant

International Nuclear Information System (INIS)

Raman, S.

1993-12-01

The additives Si, Al, MgO, P 2 O 5 were mechanically blended with fluorinelsodium calcine in varying proportions. The batches were vacuum sealed in stainless steel canisters and hot isostatically pressed at 20,000 PSI and 1000 C for 4 hours. The resulting suite of glass-ceramic waste forms parallels the natural rocks in microstructural and compositional heterogeneity. Several crystalline phases ar analogous in composition and structure to naturally occurring minerals. Additional crystalline phases are zirconia and Ca-Mg borate. The glasses are enriched in silica and alumina. Approximately 7% calcine elements occur dissolved in this glass and the total glass content in the waste forms averages 20 wt%. The remainder of the calcine elements are partitioned into crystalline phases at 75 wt% calcine waste loading. The waste forms were tested for chemical durability in accordance with the MCC1-test procedure. The leach rates are a function of the relative proportions of additives and calcine, which in turn influence the composition and abundances of the glass and crystalline phases. The DOE leach rate criterion of less than 1 g/m 2 -day is met by all the elements B, Cs and Na are increased by lowering the melt viscosity. This is related to increased crystallization or devitrification with increases in MgO addition. This exploratory work has shown that the increases in waste loading occur by preferred partitioning of the calcine components among crystalline and glass phases. The determination of optimum processing parameters in the form of additive concentration levels, homogeneous blending among the components, and pressure-temperature stabilities of phases must be continued to eliminate undesirable effects of chemical composition, microstructure and glass devitrification

Survey of matrix materials for solidified radioactive high-level waste

Energy Technology Data Exchange (ETDEWEB)

Gurwell, W.E.

1981-09-01

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

Survey of matrix materials for solidified radioactive high-level waste

International Nuclear Information System (INIS)

Gurwell, W.E.

1981-09-01

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

Waste processing to support 99Mo production at Sandia National Laboratories

International Nuclear Information System (INIS)

Longley, S.; Carson, S.; McDonald, M.

1997-01-01

As part of the isotope production program at Sandia National Laboratories (SNL), procedures are being finalized for the production of 99 Mo from the irradiation of 235 U-coated stainless-steel targets at the Technical Area (TA) V reactor and hot-cell facilities. Methods have been identified and tested for the management of the nonproduct (waste) material as the final step in the production process. These methods were developed utilizing the waste material from a series of cold and hot tests, beginning with depleted uranium powder and culminating with a test involving an irradiated 235 U target with an initial fission product inventory of ∼18000 Ci at the end of the irradiation cycle. This paper describes the radioactive waste management from the isotope production

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

International Nuclear Information System (INIS)

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

1987-01-01

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

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

International Nuclear Information System (INIS)

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

1987-01-01

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

TMI abnormal wastes disposal options

International Nuclear Information System (INIS)

Ayers, A.L. Jr.

1984-03-01

A substantial quantity of high beta-gamma/high-TRU contaminated wastes are expected from cleanup activities of Unit 2 of the Three Mile Island Nuclear Power Station. Those wastes are not disposable because of present regulatory constraints. Therefore, they must be stored temporarily. This paper discusses three options for storage of those wastes at the Idaho National Engineering Laboratory: (1) storage in temporary storage casks; (2) underground storage in vaults; and (3) storage in silos at a hot shop. Each option is analyzed and evaluated. Also included is a discussion of future disposal strategies, which might be pursued when a suitable federal or commercial repository is built

Radioactive waste management of experimental DUPIC fuel fabrication process

International Nuclear Information System (INIS)

Lee, H. H.; Park, J. J.; Shin, J. M.; Yang, M. S.; Hong, K. P.

2001-01-01

The concept of DUPIC(Direct Use of Spent PWR Fuel in CANDU Reactors) is a dry processing technology to manufacture CANDU compatible DUPIC fuel from spent PWR fuel material. Real spent PWR fuel was used in IMEF M6 hot cell to carry out DUPIC experiment. Afterwards, about 200 kg-U of spent PWR fuel is supposed to be used till 2006. This study has been conducted in some hot cells of PIEF and M6 cell of IMEF. There are various forms of nuclear material such as rod cut, powder, green pellet, sintered pellet, fabrication debris, fuel rod, fuel bundle, sample, and process waste produced from various manufacturing experiment of DUPIC fuel. After completing test, the above nuclear wastes and test equipment etc. will be classified as radioactive waste, transferred to storage facility and managed rigorously according to domestic and international laws until the final management policy is determined. It is desirable to review management options in advance for radioactive waste generated from manufacturing experiment of DUPIC nuclear fuel as well as residual nuclear material and dismantled equipment. This paper includes basic plan for DUPIC radwaste, arising source and estimated amount of radioactive waste, waste classification and packing, transport cask, transport procedures

Too hot to handle. Social and policy issues in the management of radioactive wastes

International Nuclear Information System (INIS)

Walker, C.A.; Gould, L.C.; Woodhouse, E.J.

1983-01-01

Information about the management of radioactive wastes is provided in this book. Specifically, the book attempts to supply information to further the understanding of the history of radioactive waste management in this country and the role of nuclear energy in the future of the US; the science and technology of the processes that produce radioactive wastes and of the methods proposed for managing them; the biological effects of radiation; the public attitudes about nuclear power; the nature of risks resulting from technological developments and ways of managing them; and the political institutions and processes that govern radioactive waste management. The authors have attempted to present an objective view of nuclear waste management taking a stand neither for nor against nuclear power but placing special emphasis on radioactive waste management rather than nuclear power, because they feel that the latter aspect of the subject has received much more extensive coverage elsewhere. The contents of the book are divided into 7 chapters entitled: The Radioactive Waste Management Problem, Science and Technology of the Sources and Management of Radioactive Wastes, Nuclear Waste Management and Risks to Human Health, Public Attitudes toward Radioactive Wastes, How Safe Is Safe Enough; Determinants of Perceived and Acceptable Risk, The Politics of Nuclear Waste Management, and Value Issues in Radioactive Waste Management

DOE model conference on waste management and environmental restoration

Energy Technology Data Exchange (ETDEWEB)

1990-01-01

Reports dealing with current topics in waste management and environmental restoration were presented at this conference in six sessions. Session 1 covered the Hot Topics'' including regulations and risk assessment. Session 2 dealt with waste reduction and minimization; session 3 dealt with waste treatment and disposal. Session 4 covered site characterization and analysis. Environmental restoration and associated technologies wee discussed in session 5 and 6. Individual papers have been cataloged separately.

Addendum to the Safety Analysis Report for the Steel Waste Packaging. Revision 1

International Nuclear Information System (INIS)

Crow, S.R.

1996-01-01

The Battelle Pacific Northwest National Laboratory Safety Analysis Report (SAR) for the Steel Waste Package requires additional analyses to support the shipment of remote-handled radioactive waste and special-case waste from the 324 building hot cells to PUREX for interim storage. This addendum provides the analyses required to show that this waste can be safely shipped onsite in the configuration shown

Functional design criteria radioactive liquid waste line replacement, Project W-087. Revision 3

International Nuclear Information System (INIS)

McVey, C.B.

1994-01-01

This document provides the functional design criteria for the 222-S Laboratory radioactive waste drain piping and transfer pipeline replacement. The project will replace the radioactive waste drain piping from the hot cells in 222-S to the 219-S Waste Handling Facility and provide a new waste transfer route from 219-S to the 244-S Catch Station in Tank Farms

Hot cells of the Osiris reactor

International Nuclear Information System (INIS)

Jourdain, Jean

1969-10-01

Hot cells of the Osiris reactor are β and γ type cells. Their main functions are: the extraction of irradiated samples from experimental assemblies (irradiation loops, experimental devices) used to irradiate them, the reinstallation of experimental setups with irradiated samples, the fractioning of unrecoverable experimental devices, and the removal of irradiated samples and active wastes. Each cell is therefore equipped with means for remote handling, for observation and for removal, and a venting. Each cell may also receive additional equipment, notably for the dismantling of experimental setups. This report presents the cell implantation in the reactor, elements to be handled in cells, the path followed by elements to be handled (arrival, departure, conveyors). It describes the cells (capacity and protection, design and construction, external and internal arrangements) and the cell equipment (remote handling devices, windows, lighting, lifting unit, sound system), and the installed electric power. A realisation planning is provided. An appendix indicates the cost of these hot cells

The management of steel industry by-products and waste

Energy Technology Data Exchange (ETDEWEB)

1987-01-01

The report considers the management of solid and semi-solid wastes that are reused or disposed of outside steelworks. Headings are: introduction; ironmaking slags (including generation, properties, processing, uses and disposal); (steelmaking slag from hot metal pretreatment, and primary and secondary steelmaking); ironmaking dust and sludges; steelmaking dust and sludges; millscale and sludge from continuous casting and rolling mills; treatment and handling of used oils and greases; refractory waste from refining of metallurgical furnaces and vessels; by-products, waste and wastewater arising from coke oven batteries; treatment of stainless steel waste; characterisation of waste by leaching tests; dumping technology; and conclusions.

Hot isostatic pressing of glass-zeolite composites

International Nuclear Information System (INIS)

Hash, M.C.; Pereira, C.; Lewis, M.A.

1996-01-01

Glass-zeolite waste forms are being developed for immobilizing the chloride waste salt generated from the electrometallurgical treatment of spent fuel. Glass-zeolite composites with high densities were made using hot isostatic pressing (HIP) techniques. Processing parameters were investigated to yield desirable structural ceramic properties such as mechanical, chemical, and thermal stability. Limits for these parameters were determined by differential thermal and thermogravimetric analysis. The resulting ceramic properties such as bulk density, open or apparent porosity, and leach resistance were determined. In addition, phase equilibria and particle-size distribution were observed by optical light and electron microscopy. Pre-HIP processing techniques were also studied to ensure intimate mixing of the glass and zeolite powders. Particle size distributions resulting from dry blending procedure are appropriate for needed flow and packing characteristics

Biodiversity hot spot on a hot spot: novel extremophile diversity in Hawaiian fumaroles.

Science.gov (United States)

Wall, Kate; Cornell, Jennifer; Bizzoco, Richard W; Kelley, Scott T

2015-01-06

Fumaroles (steam vents) are the most common, yet least understood, microbial habitat in terrestrial geothermal settings. Long believed too extreme for life, recent advances in sample collection and DNA extraction methods have found that fumarole deposits and subsurface waters harbor a considerable diversity of viable microbes. In this study, we applied culture-independent molecular methods to explore fumarole deposit microbial assemblages in 15 different fumaroles in four geographic locations on the Big Island of Hawai'i. Just over half of the vents yielded sufficient high-quality DNA for the construction of 16S ribosomal RNA gene sequence clone libraries. The bacterial clone libraries contained sequences belonging to 11 recognized bacterial divisions and seven other division-level phylogenetic groups. Archaeal sequences were less numerous, but similarly diverse. The taxonomic composition among fumarole deposits was highly heterogeneous. Phylogenetic analysis found cloned fumarole sequences were related to microbes identified from a broad array of globally distributed ecotypes, including hot springs, terrestrial soils, and industrial waste sites. Our results suggest that fumarole deposits function as an "extremophile collector" and may be a hot spot of novel extremophile biodiversity. © 2015 The Authors. MicrobiologyOpen published by John Wiley & Sons Ltd.

Conceptual design report of hot cell modification and process for fission Mo-99 production

International Nuclear Information System (INIS)

Park, Jin Ho; Choung, W. M.; Lee, K. I.; Hwang, D. S.; Kim, Y. K.; Park, K. B.; Jung, Y. J.; Kim, D. S.; Park, Y. C.

2001-05-01

In this conceptual design report, the basic data and design guides for detail design of fission Mo-99 production process and hot cell modification are included.The basic data and design guides for detail design of fission Mo-99 production process contains following contents. -design capacity, the basic process, process flow diagram, process material balance, process data. The basic data and design guides for modification of existing hot cell contains following contents. - plot plan of hot cell facility, the plan for shield reinforcement of hot cell, the plan for management and storage of high level liquid wastes, the plan of ventilation system, the plan for modification of auxiliary facilities. And also, the results of preliminary safety analysis(normal operation and accidents) and criticality analysis are included in this conceptual design report

Conceptual design report of hot cell modification and process for fission Mo-99 production

Energy Technology Data Exchange (ETDEWEB)

Park, Jin Ho; Choung, W. M.; Lee, K. I.; Hwang, D. S.; Kim, Y. K.; Park, K. B.; Jung, Y. J.; Kim, D. S.; Park, Y. C

2001-05-01

In this conceptual design report, the basic data and design guides for detail design of fission Mo-99 production process and hot cell modification are included.The basic data and design guides for detail design of fission Mo-99 production process contains following contents. -design capacity, the basic process, process flow diagram, process material balance, process data. The basic data and design guides for modification of existing hot cell contains following contents. - plot plan of hot cell facility, the plan for shield reinforcement of hot cell, the plan for management and storage of high level liquid wastes, the plan of ventilation system, the plan for modification of auxiliary facilities. And also, the results of preliminary safety analysis(normal operation and accidents) and criticality analysis are included in this conceptual design report.

Solar 'hot spots' are still hot

Science.gov (United States)

Bai, Taeil

1990-01-01

Longitude distributions of solar flares are not random but show evidence for active zones (or hot spots) where flares are concentrated. According to a previous study, two hot spots in the northern hemisphere, which rotate with a synodic period of about 26.72 days, produced the majority of major flares, during solar cycles 20 and 21. The more prominent of these two hot spots is found to be still active during the rising part of cycle 22, producing the majority of northern hemisphere major flares. The synodic rotation period of this hot spot is 26.727 + or - 0.007 days. There is also evidence for hot spots in the southern hemisphere. Two hot spots separated by 180 deg are found to rotate with a period of 29.407 days, with one of them having persisted in the same locations during cycles 19-22 and the other, during cycles 20-22.

Investigation of using steel slag in hot mix asphalt for the surface course of flexible pavements

Science.gov (United States)

Nguyen, Hien Q.; Lu, Dai X.; Le, Son D.

2018-04-01

The rapid development of heavy industry in Vietnam leads to the establishments of steel industry. Steel slag, a by-product of steelwork industry, under Vietnamese’s law, was considered as a deleterious solid waste which needed to be processed and landfilled. However, this has changed recently, and steel slag is now seen as a normal or non-deleterious solid waste, and has been studied for reuse in the construction industry. In this study, steel slag was used, as a replacement for mineral aggregate, in hot mix asphalt. Two hot mix asphalt mixtures with an equivalent nominal aggregate size of 12.5 (C12.5) and 19 mm (C19) were produced using steel slag. In addition, one conventional hot mix asphalt mixture of C19 was produced using mineral aggregate for comparison purpose. Investigation in laboratory condition and trial sections was carried out on Marshall tests, surface roughness, skid resistance, and modulus of the pavement before and after applying a new surface course of hot mix asphalt. The study showed that all steel slag asphalt mixtures passed the Marshall stability and flow test requirements. The skid resistance of steel slag hot mix asphalt mixtures for the surface course satisfied the Vietnamese specification for asphalt. Moreover, the pavement sections with the surface course of steel slag hot mix asphalt showed a considerable higher modulus than that of the conventional one. Only the roughness of the surface course paved with C19 did not pass the requirement of the specification.

KfK Institute for Hot Atom Chemistry. Results report on research and development activities 1989

International Nuclear Information System (INIS)

1990-03-01

The Institute for Hot Atom Chemistry is concerned with R and D tasks in nuclear fuel reprocessing. The aim is to optimize reprocessing technology with a view to safety and economic efficiency. Work is carried out within the framework of the projects reprocessing and waste management, and fast breeder in contact with WAK. The Institute increasingly carries out work within the framework of the project pollution abatement in the environment; the emphasis lies on dioxin chemistry. After the Wackersdorf task, subjects of waste management, in particular special wastes, have been added. (orig.) [de

Materials design considerations and selection for a large rad waste incinerator

International Nuclear Information System (INIS)

Vormelker, P.R.; Jenkins, C.F.; Burns, H.H.

1997-01-01

A new incinerator has been built to process self-generated, low level radioactive wastes at the Department of Energy's Savannah River Site. Wastes include protective clothing and other solid materials used during the handling of radioactive materials, and liquid chemical wastes resulting from chemical and waste management operations. The basic design and materials of construction selected to solve the anticipated corrosion problems from hot acidic gases are reviewed. Problems surfacing during trial runs prior to radioactive operations are discussed

Solar hot spots are still hot

International Nuclear Information System (INIS)

Bai, T.

1990-01-01

Longitude distributions of solar flares are not random but show evidence for active zones (or hot spots) where flares are concentrated. According to a previous study, two hot spots in the northern hemisphere, which rotate with a synodic period of about 26.72 days, produced the majority of major flares, during solar cycles 20 and 21. The more prominent of these two hot spots is found to be still active during the rising part of cycle 22, producing the majority of northern hemisphere major flares. The synodic rotation period of this hot spot is 26.727 + or - 0.007 days. There is also evidence for hot spots in the southern hemisphere. Two hot spots separated by 180 deg are found to rotate with a period of 29.407 days, with one of them having persisted in the same locations during cycles 19-22 and the other, during cycles 20-22. 14 refs

Radioactive waste management, decommissioning, spent fuel storage. V. 1. Waste management principles, decommissioning, dismantling, operations in hot environment

International Nuclear Information System (INIS)

1985-01-01

This book deals mainly with decommissioning problems concerning more particularly dismantling and decontamination techniques, and radioactive waste processing. Radioactive waste management in France and the French regulation are tackled. Equipments developed for works in hostile environment are also presented [fr

HOT 2015

DEFF Research Database (Denmark)

Hannibal, Sara Stefansen

2016-01-01

HOT samler og formidler 21 literacykyndiges bud på, hvad der er hot, og hvad der bør være hot inden for literacy – og deres begrundelser for disse bud.......HOT samler og formidler 21 literacykyndiges bud på, hvad der er hot, og hvad der bør være hot inden for literacy – og deres begrundelser for disse bud....

Diamond drilling for nuclear waste QC

International Nuclear Information System (INIS)

Jennings, Martin.

1990-01-01

Specialised diamond core drilling equipment could soon have a role to play in the safe disposal of intermediate level radioactive waste (ILW). Equipment to core and extract samples for quality checking from cement-filled steel waste drums by techniques compatible with eventual remote-handling operations in a 'hot-cell' is being developed. All coring tests carried out to date have been on simulant waste: 200 litre drums containing mixtures of Ordinary Portland Cement, Ground Granulated Blast Furnace Slag and Pulverised Fuel Ash. No radioactive materials have yet been used for the coring trials. The coring equipment and the diamond coring bits are described. (author)

Performance of Recycled Porous Hot Mix Asphalt with Gilsonite Additive

Directory of Open Access Journals (Sweden)

Ludfi Djakfar

2015-01-01

Full Text Available The objective of the study is to evaluate the performance of porous asphalt using waste recycled concrete material and explore the effect of adding Gilsonite to the mixture. As many as 90 Marshall specimens were prepared with varied asphalt content, percentage of Gilsonite as an additive, and proportioned recycled and virgin coarse aggregate. The test includes permeability capability and Marshall characteristics. The results showed that recycled concrete materials seem to have a potential use as aggregate in the hot mix asphalt, particularly on porous hot mix asphalt. Adding Gilsonite at ranges 8–10% improves the Marshall characteristic of the mix, particularly its stability, without decreasing significantly the permeability capability of the mix. The use of recycled materials tends to increase the asphalt content of the mix at about 1 to 2% higher. With stability reaching 750 kg, the hot mix recycled porous asphalt may be suitable for use in the local roads with medium vehicle load.

Preliminary safety analysis report for the Auxiliary Hot Cell Facility, Sandia National Laboratories, Albuquerque, New Mexico

International Nuclear Information System (INIS)

OSCAR, DEBBY S.; WALKER, SHARON ANN; HUNTER, REGINA LEE; WALKER, CHERYL A.

1999-01-01

The Auxiliary Hot Cell Facility (AHCF) at Sandia National Laboratories, New Mexico (SNL/NM) will be a Hazard Category 3 nuclear facility used to characterize, treat, and repackage radioactive and mixed material and waste for reuse, recycling, or ultimate disposal. A significant upgrade to a previous facility, the Temporary Hot Cell, will be implemented to perform this mission. The following major features will be added: a permanent shield wall; eight floor silos; new roof portals in the hot-cell roof; an upgraded ventilation system; and upgraded hot-cell jib crane; and video cameras to record operations and facilitate remote-handled operations. No safety-class systems, structures, and components will be present in the AHCF. There will be five safety-significant SSCs: hot cell structure, permanent shield wall, shield plugs, ventilation system, and HEPA filters. The type and quantity of radionuclides that could be located in the AHCF are defined primarily by SNL/NM's legacy materials, which include radioactive, transuranic, and mixed waste. The risk to the public or the environment presented by the AHCF is minor due to the inventory limitations of the Hazard Category 3 classification. Potential doses at the exclusion boundary are well below the evaluation guidelines of 25 rem. Potential for worker exposure is limited by the passive design features incorporated in the AHCF and by SNL's radiation protection program. There is no potential for exposure of the public to chemical hazards above the Emergency Response Protection Guidelines Level 2

Preliminary safety analysis report for the Auxiliary Hot Cell Facility, Sandia National Laboratories, Albuquerque, New Mexico

Energy Technology Data Exchange (ETDEWEB)

OSCAR,DEBBY S.; WALKER,SHARON ANN; HUNTER,REGINA LEE; WALKER,CHERYL A.

1999-12-01

The Auxiliary Hot Cell Facility (AHCF) at Sandia National Laboratories, New Mexico (SNL/NM) will be a Hazard Category 3 nuclear facility used to characterize, treat, and repackage radioactive and mixed material and waste for reuse, recycling, or ultimate disposal. A significant upgrade to a previous facility, the Temporary Hot Cell, will be implemented to perform this mission. The following major features will be added: a permanent shield wall; eight floor silos; new roof portals in the hot-cell roof; an upgraded ventilation system; and upgraded hot-cell jib crane; and video cameras to record operations and facilitate remote-handled operations. No safety-class systems, structures, and components will be present in the AHCF. There will be five safety-significant SSCs: hot cell structure, permanent shield wall, shield plugs, ventilation system, and HEPA filters. The type and quantity of radionuclides that could be located in the AHCF are defined primarily by SNL/NM's legacy materials, which include radioactive, transuranic, and mixed waste. The risk to the public or the environment presented by the AHCF is minor due to the inventory limitations of the Hazard Category 3 classification. Potential doses at the exclusion boundary are well below the evaluation guidelines of 25 rem. Potential for worker exposure is limited by the passive design features incorporated in the AHCF and by SNL's radiation protection program. There is no potential for exposure of the public to chemical hazards above the Emergency Response Protection Guidelines Level 2.

Hot demonstration of proposed commercial nuclide removal technology

International Nuclear Information System (INIS)

Lee, D.

1996-01-01

This task covers the development and operation of an experimental test unit located in a Building 4501 hot cell within Building 4501 at Oak Ridge National Laboratory (ORNL). This equipment is designed to test radionuclides removal technologies under continuous operatoin on actual ORNL Melton Valley Storage Tank (MVST) supernatant, Savannah River high-level waste supernatant, and Hanford supernatant. The latter two may be simulated by adding the appropriate chemicals and/or nuclides to the MVST supernatant

Is waste problem solved

International Nuclear Information System (INIS)

Anon.

1977-01-01

In a short report, with flow diagram, a process being developed in Sweden by ASEA for the processing of radioactive wastes by hot isostatic pressing is briefly described. Three methods have been developed. In the first, calcinated waste is mixed with durable ceramic materials and sintered at high pressure to massive objects whose crystalline state is very stable. The second concentrates the waste materials in solution on inorganic ion exchangers, which, after drying, are converted to massive ceramic objects by the same process and with the same properties, as in the first method. In the third method, spent fuel elements are directly processed into objects with high strength and stability by the same processes as above. The final products may be permanently stored in ground water in geological formations. (JIW)

327 legacy waste processing plan

International Nuclear Information System (INIS)

Henderson, J.F.

1998-01-01

The B and W Hanford Company's (BWHC) 327 Facility [Postirradiation Testing Laboratory (PTL)] houses 10 hot cells in which a variety of postirradiation examinations have been performed since its construction in the mid 1950s. Over the years, the waste that was generated in these cells has been collected in one gallon buckets. These buckets are essentially one gallon cylindrical cans made of thin wall stainless steel with welded bottoms and slip fit lids. They contain assorted compactable waste (i.e., Wipe-Alls, Q-tips, towels, etc.) as well as non-compactable waste (i.e., small tools, pieces of metal tubing, etc.). There is a FY-98 BWHC Performance Agreement (PA) milestone in place to package 200 of these buckets in drums and ship them from the 327 facility to the Central Waste Complex (CWC) by September 30, 1998

Waste heat recovery system

International Nuclear Information System (INIS)

Phi Wah Tooi

2010-01-01

Full text: The Konzen in-house designed anaerobic digester system for the POME (Palm Oil Mill Effluent) treatment process is one of the registered Clean Development Mechanism (CDM) projects in Malaysia. It is an organic wastewater treatment process which achieves excellent co-benefits objectives through the prevention of water pollution and reduction of greenhouse gas emissions, which is estimated to be 40,000 to 50,000 t-CO 2 per year. The anaerobic digester was designed in mesophile mode with temperature ranging from 37 degree Celsius to 45 degree Celsius. A microorganisms growth is optimum under moderately warm temperature conditions. The operating temperature of the anaerobic digester needs to be maintained constantly. There are two waste heat recovery systems designed to make the treatment process self-sustaining. The heat recovered will be utilised as a clean energy source to heat up the anaerobic digester indirectly. The first design for the waste heat recovery system utilises heat generated from the flue gas of the biogas flaring system. A stainless steel water tank with an internal water layer is installed at the top level of the flare stack. The circulating water is heated by the methane enriched biogas combustion process. The second design utilizes heat generated during the compression process for the biogas compressor operation. The compressed biogas needs to be cooled before being recycled back into the digester tank for mixing purposes. Both the waste heat recovery systems use a design which applies a common water circulation loop and hot water tank to effectively become a closed loop. The hot water tank will perform both storage and temperature buffer functions. The hot water is then used to heat up recycled sludge from 30 degree Celsius to 45 degree Celsius with the maximum temperature setting at 50 degree Celsius. The recycled sludge line temperature will be measured and monitored by a temperature sensor and transmitter, which will activate the

HotRegion: a database of predicted hot spot clusters.

Science.gov (United States)

Cukuroglu, Engin; Gursoy, Attila; Keskin, Ozlem

2012-01-01

Hot spots are energetically important residues at protein interfaces and they are not randomly distributed across the interface but rather clustered. These clustered hot spots form hot regions. Hot regions are important for the stability of protein complexes, as well as providing specificity to binding sites. We propose a database called HotRegion, which provides the hot region information of the interfaces by using predicted hot spot residues, and structural properties of these interface residues such as pair potentials of interface residues, accessible surface area (ASA) and relative ASA values of interface residues of both monomer and complex forms of proteins. Also, the 3D visualization of the interface and interactions among hot spot residues are provided. HotRegion is accessible at http://prism.ccbb.ku.edu.tr/hotregion.

ANSTO's waste forms for the 31. century

International Nuclear Information System (INIS)

Vance, E.R.; Begg, B. D.; Day, R. A.; Moricca, S.; Perera, D. S.; Stewart, M. W. A.; Carter, M. L.; McGlinn, P. J.; Smith, K. L.; Walls, P. A.; Robina, M. La

2004-01-01

ANSTO waste form development for high-level radioactive waste is directed towards practical applications, particularly problematic niche wastes that do not readily lend themselves to direct vitrification. Integration of waste form chemistry and processing method is emphasised. Some longstanding misconceptions about titanate ceramics are dealt with. We have a range of titanate-bearing waste form products aimed at immobilisation of tank wastes and sludges, actinide-rich wastes, INEEL calcines and Na-bearing liquid wastes, Al-rich wastes arising from reprocessing of Al-clad fuels, Mo-rich wastes arising from reprocessing of U-Mo fuels, partitioned Cs-rich wastes, and 99 Tc. Waste form production techniques cover hot isostatic and uniaxial pressing, sintering, and cold-crucible melting, and these are strongly integrated into waste form design. Speciation and leach resistance of Cs and alkalis in cementitious products and geo-polymers are being studied. Recently we have embarked on studies of candidate inert matrix fuels for Pu burning. We also have a considerable program directed at basic understanding of the waste forms in regard to crystal chemistry, dissolution behaviour in aqueous media, radiation damage effects and optimum processing techniques. (authors)

ANSTO's waste forms for the 31. century

Energy Technology Data Exchange (ETDEWEB)

Vance, E R; Begg, B D; Day, R A; Moricca, S; Perera, D S; Stewart, M W. A.; Carter, M L; McGlinn, P J; Smith, K L; Walls, P A; Robina, M La

2004-07-01

ANSTO waste form development for high-level radioactive waste is directed towards practical applications, particularly problematic niche wastes that do not readily lend themselves to direct vitrification. Integration of waste form chemistry and processing method is emphasised. Some longstanding misconceptions about titanate ceramics are dealt with. We have a range of titanate-bearing waste form products aimed at immobilisation of tank wastes and sludges, actinide-rich wastes, INEEL calcines and Na-bearing liquid wastes, Al-rich wastes arising from reprocessing of Al-clad fuels, Mo-rich wastes arising from reprocessing of U-Mo fuels, partitioned Cs-rich wastes, and {sup 99}Tc. Waste form production techniques cover hot isostatic and uniaxial pressing, sintering, and cold-crucible melting, and these are strongly integrated into waste form design. Speciation and leach resistance of Cs and alkalis in cementitious products and geo-polymers are being studied. Recently we have embarked on studies of candidate inert matrix fuels for Pu burning. We also have a considerable program directed at basic understanding of the waste forms in regard to crystal chemistry, dissolution behaviour in aqueous media, radiation damage effects and optimum processing techniques. (authors)

Decommissioning of the Risoe Hot Cell facility

International Nuclear Information System (INIS)

Carlsen, H.

1994-06-01

Nuclear fuels have been handled and examined after irradiation by physical and chemical techniques, and radiotherapy sources, mainly 60 Co, have been produced at Risoe National Laboratory since 1964. The aims of decommissioning (during 1990-94, at IAEA Stage 2 level for reactors) were to obtain safe conditions for the remaining parts of the facility and to produce clean space areas for new projects. The facility comprises 6 concrete cells, several lead-shielded steel cells, glove boxes, shielded storage for waste, a remotely operated optical microscope, a frogman area for personnel access to the concrete cells, a decontamination facility, workshops and safety installations. All steel cells, glove boxes and the microscope were emptied and removed. The concrete cells were emptied of fissile material, scientific equipment, general tools and scrap. Decontamination should facilitate waste packing and reduce amount of waste to be stored temporarily at the Risoe waste treatment facility together with highly active waste. The concrete cells were cleaned remotely by wiping, hot spot removal, by mechanical means and vacuum cleaning. The interiors of 2 cells were decontaminated by high pressure water jetting. All master-slave manipulators and part of the contaminated ventilation system at the cells were removed. The cells are left in a non-ventilated state, connected to the atmosphere by an absolute filter. The main contaminants measured before cell closure were 60 Co, 137 Cs and alpha-emitters. Dismantling, decontamination waste disposal and received doses are described. Simple techniques involving low doses were found to be very effective. (AB)

Development and testing of matrices for the encapsulation of glass and ceramic nuclear waste forms

International Nuclear Information System (INIS)

Wald, J.W.; Brite, D.W.; Gurwell, W.E.; Buckwalter, C.Q.; Bunnell, L.R.; Gray, W.J.; Blair, H.T.; Rusin, J.M.

1982-02-01

This report details the results of research on the matrix encapsulation of high level wastes at PML over the past few years. The demonstrations and tests described were designed to illustrate how the waste materials are effected when encapsulated in an inert matrix. Candidate materials evaluated for potential use as matrices for encapslation of pelletized ceramics or glass marbles were categorized into four groups: metals, glasses, ceramics, and graphite. Two processing techniques, casting and hot pressing, were investigated as the most promising methods of formation or densification of the matrices. The major results reported deal with the development aspects. However, chemical durability tests (leach tests) of the matrix materials themselves and matrix-waste form composites are also reported. Matrix waste forms can provide a low porosity, waste-free barrier resulting in increased leach protection, higher impact strength and improved thermal conductivity compared to unencapsulated glass or ceramic waste materials. Glass marbles encapsulated in a lead matrix offer the most significant improvement in waste form stability of all combinations evaluated. This form represents a readily demonstrable process that provides high thermal conductivity, mechanical shock resistance, radiation shielding and increased chemical durability through both a chemical passivation mechanism and as a physical barrier. Other durable matrix waste forms evaluated, applicable primarily to ceramic pellets, involved hot-pressed titanium or TiO 2 materials. In the processing of these forms, near 100% dense matrices were obtained. The matrix materials had excellent compatibility with the waste materials and superior potential chemical durability. Cracking of the hot-pressed ceramic matrix forms, in general, prevented the realization of their optimum properties

A hot air driven thermoacoustic-Stirling engine

Energy Technology Data Exchange (ETDEWEB)

Tijani, M.E.H.; Spoelstra, S. [ECN Biomass and Energy Efficiency, Petten (Netherlands)

2012-09-15

Significant energy savings can be obtained by implementing a thermally driven heat pump into industrial or domestic applications. Such a thermally driven heat pump uses heat from a high-temperature source to drive the system which upgrades an abundantly available heat source (industrial waste heat, air, water, geothermal). A way to do this is by coupling a thermoacoustic engine with a thermoacoustic heat pump. The engine is driven by a burner and produces acoustic power and heat at the required temperature. The acoustic power is used to pump heat in the heat pump to the required temperature. This system is attractive since it uses a noble gas as working medium and has no moving mechanical parts. This paper deals with the first part of this system: the engine. In this study, hot air is used to simulate the flue gases originating from a gas burner. This is in contrast with a lot of other studies of thermoacoustic engines that use an electrical heater as heat source. Using hot air resembles to a larger extent the real world application. The engine produces about 300W of acoustic power with a performance of 41% of the Carnot efficiency at a hot air temperature of 620C.

Plasma technology for treatment of waste

Energy Technology Data Exchange (ETDEWEB)

Cohn, D [Massachusetts Inst. of Technology, Cambridge, MA (United States). Plasma Fusion Center

1997-12-31

Meeting goals for waste cleanup will require new technology with improved environmental attractiveness and reduced cost. Plasma technology appears promising because of the high degree of controllability; capability to process waste without the adverse effects of combustion; and a very wide temperature range of operation. At the Plasma Fusion Center at the Massachusetts Institute of Technology, a range of plasma technologies are being investigated. `Hot` thermal plasmas produced by DC arc technology are being examined for treatment of solid waste. In conjunction with this activity, new diagnostics are being developed for monitoring arc furnace operating parameters and gaseous emissions. Electron-beam generated plasma technology is being investigated as a means of producing non-thermal `cold` plasmas for selective processing of dilute concentrations of gaseous waste. (author). 4 figs., 5 refs.

Anaerobic bioconversion of organic waste into biogas by hot water treatment at near-critical conditions: application in bioregenerative life support.

Science.gov (United States)

Lissens, Geert; Verstraete, Willy; Albrecht, Tobias; Brunner, Gerd; Lasseur, Christophe

2003-01-01

The feasibility of nearly-complete conversion of lignocellulosic waste (70% food crops, 20% faecal matter and 10% green algae) into biogas was investigated in the context of a Life Support Project. The treatment comprised a series of processes, i.e. a mesophilic laboratory scale CSTR (continuously stirred tank reactor), an upflow biofilm reactor and a hydrothermolysis system in near-critical water. By the one-stage CSTR, a biogas yield of 75% with a specific biogas production of 0.37 l biogas g(-1) VSS (volatile suspended solids) added at a HRT (hydraulic retention time) of 20 d was obtained. Biogas yields further increased with 10-15% at HRT > 20 d, indicating the hydrolysis of lignocellulose to be the rate-limiting conversion step. The solids present in the CSTR-effluent were subsequently treated by hot water treatment (T approximately 310-350 degrees C, p approximately 240 bar), resulting in effective carbon liquefaction (50-60% without and 83% with carbon dioxide saturation) and complete hygienisation of the residue. Subsequent anaerobic digestion of the hydrolysate allowed further conversion of 48-60% on COD (chemical oxygen demand) basis. Thus, the total process yielded biogas corresponding with a COD conversion up to 90% of the original organic matter. It appears that mesophilic digestion in conjunction with hydrothermolysis at near-critical conditions offers interesting features for (nearly) complete, non-toxic and hygienic carbon and energy recovery from human waste in a bioregenerative life support context.

Waste volume reduction by acid digestion

International Nuclear Information System (INIS)

Lerch, R.E.; Divine, J.R.

1975-06-01

Acid digestion is a process being developed at the Hanford Engineering Development Laboratory (HEDL) in Richland, Washington, to reduce the volume of alpha-contaminated combustible waste by converting it into a non-combustible residue. Typical waste materials such as polyvinylchloride (PVC), polyethylene, paper and other cellulosic materials, ion exchange resin, all types of rubber, etc., are digested in hot (230 0 C--270 0 C) concentrated sulfuric acid containing nitric acid oxidant to form inert residues generally having less than four percent of their original volume and less than twenty-five percent of their original mass. The process is currently being tested using non-radioactive waste in an Acid Digestion Test Unit (ADTU) with all glass equipment. Engineering tests to date have shown acid digestion to be a potentially attractive method for treating combustible waste materials. Based on results of the engineering tests, an acid digestion pilot unit capable of treating radioactive wastes is being designed and constructed. Design capacity of the pilot unit for radioactive waste will be 100 kg of waste per day. (U.S.)

FRACTIONAL CRYSTALLIZATION FLOWSHEET TESTS WITH ACTUAL TANK WASTE

International Nuclear Information System (INIS)

HERTING, D.L.

2006-01-01

Laboratory-scale flowsheet tests of the fractional crystallization process were conducted with actual tank waste samples in a hot cell at the 222-S Laboratory. The process is designed to separate medium-curie liquid waste into a low-curie stream for feeding to supplemental treatment and a high-curie stream for double-shell tank storage. Separations criteria (for Cs-137 sulfate, and sodium) were exceeded in all three of the flowsheet tests that were performed

Mechanical and thermophysical properties of hot-pressed SYNROC B

International Nuclear Information System (INIS)

Hoenig, C.L.; Newkirk, H.W.; Otto, R.A.; Brady, R.L.; Brown, A.E.; Ulrich, A.R.; Lum, R.C.

1981-01-01

The optimal SYNROC compositons for use with commercial waste are reviewed. Large amounts of powder (about 2.5 kg) were prepared by convention al ceramic operations to test the SYNROC concept on a processing scale. Samples, 15.2 cm in diameter, were hot pressed in graphite, and representative samples were cut for microstructural evaluations. Measured mechanical and thermophysical properties did not vary significantly as a function of sample location and were typical of titanate ceramic materials

Ceramic waste form qualification using results from witness tubes

International Nuclear Information System (INIS)

O'Holleran, T.P.; Johnson, S.G.; Bateman, K.J.

2002-01-01

A ceramic waste form has been developed to immobilize the salt waste stream from electrometallurgical treatment of spent nuclear fuel. The ceramic waste form is prepared in a hot isostatic press (HIP). The use of small, easily fabricated HIP capsules called witness tubes has been proposed as a practical way to obtain representative samples of ceramic waste form material for process monitoring, waste form qualification, and archiving. Witness tubes are filled with the same material used to fill the corresponding HIP can, and are HIPed along with the HIP can. Relevant physical, chemical, and performance (leach test) data are analyzed and compared. Differences between witness tube and HIP can materials are shown to be statistically insignificant, demonstrating that witness tubes do provide ceramic waste form material representative of the material in the corresponding HIP can.

HOT 2012

DEFF Research Database (Denmark)

Lund, Henriette Romme

Undersøgelse af, hvad der er hot - og hvad der burde være hot på læseområdet med 21 læsekyndige. Undersøgelsen er gennemført siden 2010. HOT-undersøgelsen er foretaget af Nationalt Videncenter for Læsning - Professionshøjskolerne i samarb. med Dansklærerforeningen......Undersøgelse af, hvad der er hot - og hvad der burde være hot på læseområdet med 21 læsekyndige. Undersøgelsen er gennemført siden 2010. HOT-undersøgelsen er foretaget af Nationalt Videncenter for Læsning - Professionshøjskolerne i samarb. med Dansklærerforeningen...

The defense waste processing facility: A status report

International Nuclear Information System (INIS)

Cowan, S.P.; Fulmer, D.C.

1987-01-01

The Defense Waste Processing Fascility (DWPF) will be the nation's first production scale facility for immobilizing high-level waste for disposal. It will also be the largest facility of its kind in the world. The technology, design, and construction efforts are on schedule for ''hot'' operation in fiscal year 1990. This paper provides a status report on the DWPF technology, design, and construction, and describes some of the challenges that have arisen during design and construction

Removal of an acid fume system contaminated with perchlorates located within hot cell

International Nuclear Information System (INIS)

Rosenberg, K.E.; Henslee, S.P.; Vroman, W.R.; Krsul, J.R.; Michelbacher, J.A.; Knighton, G.C.

1992-09-01

An add scrubbing system located within the confines of a highly radioactive hot cell at Argonne National Laboratory-West (ANL-W) was remotely removed. The acid scrubbing system was routinely used for the dissolution of irradiated reactor fuel samples and structural materials. Perchloric acid was one of the acids used in the dissolution process and remained in the system with its inherent risks. Personnel could not enter the hot cell to perform the dismantling of the acid scabbing system due to the high radiation field and the explosion potential associated with the perchlorates. A robot was designed and built at ANL-W and used to dismantle the system without the need for personnel entry into the hot cell. The robot was also used for size reduction of removed components and loading of the removed components into waste containers

Decommissioning of the Risoe Hot Cell facility

International Nuclear Information System (INIS)

Carlsen, H.

1993-10-01

A concise description of the current status of the decommissioning of the hot cell capacity at Risoe National Laboratory is given in this 6th periodic report covering January 1st to June 30th, 1993. All registered and safeguarded fissile material has been removed and the task of cutting and packing scrap material and experimental equipment from the concrete cell line has been completed. Concrete cells 5 and 6 have been finally cleaned and the master slave manipulators removed from them. The major part of the contamination on the shutters and shutter houses were on their horizontal planes and the main contaminant was 137 Cs. Here the surfaces were cleaned by wiping with wet cloths. The method is described. Tables illustrating the resulting contamination levels are included, the density is now low on the shutters. The method of final inn-cell cleaning is explained, and here again tables represent the resulting contamination levels. The work on ''hot spot'' removal and remote cleaning by vacuuming continues on the remaining cells. A collective dose of ca. 16.3 man-mSv was ascribed to 18 persons in the first half of 1993, arising mainly from in-cell work and waste handling. To sum up, the main results from this period are successful removal of last waste from the cells, remote cleaning of cells 2 and 3, final condition for all shutters and shutter housings and final condition for cells 5 and 6. Tables illustrate measured dose rates in detail. (AB)

Report of the Institute for Hot Chemistry on research and development in 1982

International Nuclear Information System (INIS)

1983-02-01

The Institute for Hot Chemistry is concerned with research and development programmes in the field of re-processing nuclear fuels. The investigations are oriented towards the objectives of the planned waste disposal plant and are carried out within the frame-work of the Reprocessing, Waste Treatment and Fast Breeder Projects, with the cooperation of the firms DWK and WAK. The Institute can be divided up into the following subject areas: extraction chemistry and plant operation, analytical processing, chemical processing and apparatus development; solvent and waste gas treatment; process control and automation; organic analysis; and fundamental research. In the developmental stage, evaluations are carried out up to the kilogram and kilo-Curie level, at the technical level, however, up to a daily throughput in tonnes. (orig.) [de

33 Shafts Category of Transuranic Waste Stored Below Ground within Area G

Energy Technology Data Exchange (ETDEWEB)

Hargis, Kenneth Marshall [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Monk, Thomas H [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2017-05-22

This report compiles information to support the evaluation of alternatives and analysis of regulatory paths forward for the 33 shafts. The historical information includes a form completed by waste generators for each waste package (Reference 6) that included a waste description, estimates of Pu-239 and uranium-235 (U-235) based on an accounting technique, and calculations of mixed fission products (MFP) based on radiation measurements. A 1979 letter and questionnaire (Reference 7) provides information on waste packaging of hot cell waste and the configuration of disposal shafts as storage in the 33 Shafts was initiated. Tables of data by waste package were developed during a review of historical documents that was performed in 2005 (Reference 8). Radiological data was coupled with material-type data to estimate the initial isotopic content of each waste package and an Oak Ridge National Laboratory computer code was used to calculate 2009 decay levels. Other sources of information include a waste disposal logbook for the 33 shafts (Reference 9), reports that summarize remote-handled waste generated at the CMR facility (Reference 10) and placement of waste in the 33 shafts (Reference 11), a report on decommissioning of the LAMPRE reactor (Reference 12), interviews with an employee and manager involved in placing waste in the 33 shafts (References 13 and 14), an interview with a long-time LANL employee involved in waste operations (Reference 15), a 2002 plan for disposition of remote-handled TRU waste (Reference 16), and photographs obtained during field surveys of several shafts in 2007. The WIPP Central Characterization Project (CCP) completed an Acceptable Knowledge (AK) summary report for 16 canisters of remote-handled waste from the CMR Facility that contains information relevant to the 33 Shafts on hot-cell operations and timeline (Reference 17).

Bitumen modified with recycled polyurethane foam for employment in hot mix asphalt

Directory of Open Access Journals (Sweden)

Miguel Ángel Salas

2018-01-01

Full Text Available A wide variety of modifiers have been applied to bitumen in order to enhance their properties and performance. Among them, polymers have been mainly used. The aim of this paper is to assess the use of polyurethane foam waste as a bitumen modifier for hot mix asphalts. The polyurethane foam is a by-product of the manufacturing of polyurethane for thermal insulation. From a bitumen with a penetration grade of 50/70, various samples with percentages of waste material in weight ranging from 1% to 5% were produced and tested. Samples with 5% of waste material or more became rough and were refused due to their poor workability. A bituminous mixture with modified bitumen with a 4% of polyurethane was manufactured and compared with a sample with the same aggregates and original bitumen. Results in Marshall test showed that a mix with polymer modified bitumen yielded improvements in stability and a lower deformability. This result suggests that the employment of polyurethane foam waste is a promising bitumen modifier, contributing also to recycle waste materials.

Carbon Bed Mercury Emissions Control For Mixed Waste Treatment

International Nuclear Information System (INIS)

Soelberg, Nick; Enneking, Joe

2010-01-01

Mercury has had various uses in nuclear fuel reprocessing and other nuclear processes, and so is often present in radioactive and mixed (both radioactive and hazardous according to the Resource Conservation and Recovery Act) wastes. Depending on regulatory requirements, the mercury in the off-gas must be controlled with sometimes very high efficiencies. Compliance to the Hazardous Waste Combustor (HWC) Maximum Achievable Control Technology (MACT) standards can require off-gas mercury removal efficiencies up to 99.999% for thermally treating some mixed waste streams. Several test programs have demonstrated this level of off-gas mercury control using fixed beds of granular sulfur-impregnated activated carbon. Other results of these tests include: (a) The depth of the mercury control mass transfer zone was less than 15-30 cm for the operating conditions of these tests, (b) MERSORB(reg s ign) carbon can sorb Hg up to 19 wt% of the carbon mass, and (c) the spent carbon retained almost all (98-99.99%) of the Hg; but when even a small fraction of the total Hg dissolves, the spent carbon can fail the TCLP test when the spent carbon contains high Hg concentrations. Localized areas in a carbon bed that become heated through heat of adsorption, to temperatures where oxidation occurs, are referred to as 'bed hot spots.' Carbon bed hot spots must be avoided in processes that treat radioactive and mixed waste. Key to carbon bed hot spot mitigation are (a) designing for sufficient gas velocity, for avoiding gas flow maldistribution, and for sufficient but not excessive bed depth, (b) monitoring and control of inlet gas flowrate, temperature, and composition, (c) monitoring and control of in-bed and bed outlet gas temperatures, and (d) most important, monitoring of bed outlet CO concentrations. An increase of CO levels in the off-gas downstream of the carbon bed to levels about 50-100 ppm higher than the inlet CO concentration indicate CO formation in the bed, caused by carbon bed

ASEA reduces nuclear waste in Sweden

International Nuclear Information System (INIS)

Anon.

1977-01-01

ASEA, the worldwide manufacturer of electrical and mechanical equipment with headquarters in Vasteras, Sweden, is working on the development of a method using hot isostatic pressing for the safe containment and storage of nuclear waste. This development work is being carried out at the company's High-Pressure Laboratory at Robertsfors, in the north of Sweden

Radiation shielding design for a hot repair facility

International Nuclear Information System (INIS)

Courtney, J.C.; Dwight, C.C.

1991-01-01

A new repair and decontamination area is being built to support operations at the demonstration fuel cycle facility for the Integral Fast Reactor program at Argonne National Laboratory's site at the Idaho National Engineering Laboratory. Provisions are made for remote, glove wall, and contact maintenance on equipment removed from hot cells where spent fuel will be electrochemically processed and recycled to the Experimental Breeder Reactor-II. The source for the shielding design is contamination from a mix of fission and activation products present on items removed from the hot cells. The repair facility also serves as a transfer path for radioactive waste produced by processing operations. Radiation shields are designed to limit dose rates to no more than 5 microSv h-1 (0.5 mrem h-1) in normally occupied areas. Point kernel calculations with buildup factors have been used to design the shielding and to position radiation monitors within the area

Process and equipment qualification of the ceramic and metal waste forms for spent fuel treatment

International Nuclear Information System (INIS)

Marsden, Ken; Knight, Collin; Bateman, Kenneth; Westphal, Brian; Lind, Paul

2005-01-01

The electrometallurgical process for treating sodium-bonded spent metallic fuel at the Materials and Fuels Complex of the Idaho National Laboratory separates actinides and partitions fission products into two waste forms. The first is the metal waste form, which is primarily composed of stainless steel from the fuel cladding. This stainless steel is alloyed with 15w% zirconium to produce a very corrosion-resistant metal which binds noble metal fission products and residual actinides. The second is the ceramic waste form which stabilizes fission product-loaded chloride salts in a sodalite and glass composite. These two waste forms will be packaged together for disposal at the Yucca Mountain repository. Two production-scale metal waste furnaces have been constructed. The first is in a large argon-atmosphere glovebox and has been used for equipment qualification, process development, and process qualification - the demonstration of process reliability for production of the DOE-qualified metal waste form. The second furnace will be transferred into a hot cell for production of metal waste. Prototype production-scale ceramic waste equipment has been constructed or procured; some equipment has been qualified with fission product-loaded salt in the hot cell. Qualification of the remaining equipment with surrogate materials is underway. (author)

Literature survey on metal waste form for metallic waste from electrorefiners for the electrometallurgical treatment of spent metallic fuels

International Nuclear Information System (INIS)

Nishimura, Tomohiro

2003-01-01

This report summarizes the recent results of the metal waste form development activities at the Argonne National Laboratory in the USA for high-level radioactive metallic waste (stainless-steel (SS) cladding hulls, zirconium (Zr), noble-metal fission products (NMFPs), etc.) from electrorefiners for the electrometallurgical treatment of spent metallic fuels. Their main results are as follows: (1) SS- 15 wt.% Zr- ∼4 wt.% NMFPs alloy was selected as the metal waste form, (2) metallurgical data, properties, long-term corrosion data, etc. of the alloy have been collected, (3) 10-kg ingots have been produced in hot tests and a 60-kg production machine is under development. The following research should be made to show the feasibility of the metal waste form in Japan: (1) degradation assessment of the metal waste form in Japanese geological repository environments, and (2) clarification of the maximum allowable contents of NMFPs. (author)

Progress report on safety research of high-level waste management for the period April, 1981 to March, 1982

International Nuclear Information System (INIS)

Tashiro, Shingo

1982-10-01

Main results obtained on Safety Research of High-Level Waste Management in 1981 were edited. The research tjeme are following. (1) Characterization of vitrified waste. (2) Alternative waste form development. (3) Durability tests for HLW storage facility. (4) Safety evaluation of geologic disposal. (5) Preparation for hot test. (author)

2020 Vision for Tank Waste Cleanup (One System Integration) - 12506

Energy Technology Data Exchange (ETDEWEB)

Harp, Benton; Charboneau, Stacy; Olds, Erik [US DOE (United States)

2012-07-01

The mission of the Department of Energy's Office of River Protection (ORP) is to safely retrieve and treat the 56 million gallons of Hanford's tank waste and close the Tank Farms to protect the Columbia River. The millions of gallons of waste are a by-product of decades of plutonium production. After irradiated fuel rods were taken from the nuclear reactors to the processing facilities at Hanford they were exposed to a series of chemicals designed to dissolve away the rod, which enabled workers to retrieve the plutonium. Once those chemicals were exposed to the fuel rods they became radioactive and extremely hot. They also couldn't be used in this process more than once. Because the chemicals are caustic and extremely hazardous to humans and the environment, underground storage tanks were built to hold these chemicals until a more permanent solution could be found. The Cleanup of Hanford's 56 million gallons of radioactive and chemical waste stored in 177 large underground tanks represents the Department's largest and most complex environmental remediation project. Sixty percent by volume of the nation's high-level radioactive waste is stored in the underground tanks grouped into 18 'tank farms' on Hanford's central plateau. Hanford's mission to safely remove, treat and dispose of this waste includes the construction of a first-of-its-kind Waste Treatment Plant (WTP), ongoing retrieval of waste from single-shell tanks, and building or upgrading the waste feed delivery infrastructure that will deliver the waste to and support operations of the WTP beginning in 2019. Our discussion of the 2020 Vision for Hanford tank waste cleanup will address the significant progress made to date and ongoing activities to manage the operations of the tank farms and WTP as a single system capable of retrieving, delivering, treating and disposing Hanford's tank waste. The initiation of hot operations and subsequent full operations

Stripping in hot mix asphalt produced by aggregates from construction and demolition waste.

Science.gov (United States)

Pérez, I; Pasandín, A R; Gallego, J

2012-01-01

This paper analyses the effect of water on the durability of hot asphalt mixtures made with recycled aggregates from construction and demolition debris. Indirect tensile stress tests were carried out to evaluate stripping behaviour. The mixtures tested were fabricated with 0, 20, 40 and 60% recycled aggregates. Two types of natural aggregates were used: schist and calcite dolomite. An increase in the percentage of recycled aggregates was found to produce a decrease in the tensile stress ratio of the hot asphalt mixtures. To study this phenomenon, two and three factor analyses of variance (ANOVA) were performed with indirect tensile stress being used as the dependent variable. The factors studied were the percentage of recycled aggregates (0, 20, 40 and 60%), the moisture state (dry, wet) and the type of natural aggregate (schist, calcite). On the basis of the ANOVA results, it was found that the most important factor affecting resistance was the moisture state (dry, wet) of the specimens. The percentage of recycled aggregate also affected indirect tensile stress, especially in the dry state. The type of natural aggregate did not have a significant effect on indirect tensile stress. The hot asphalt mixture specimens made with different percentages of recycled aggregates from construction and demolition debris and of natural quarry aggregates showed poor stripping behaviour. This stripping behaviour can be related to both the poor adhesion of the recycled aggregates and the high absorption of the mortar of cement adhered to them.

DOE states reheat nuclear waste debate

International Nuclear Information System (INIS)

Crawford, M.

1985-01-01

After decades of struggling with the issue, Congress in late 1982 established a firm plan for burying growing volumes of nuclear reactor wastes. But 2 l/2 years later the waste disposal debate is as hot as ever. Utility companies, environmentalists, federal officials, and state governments are again clashing - this time over the way the program is proceeding. The Nuclear Waste Policy Act calls for the Department of Energy to start accepting wastes in 1998 at the first of two planned repositories. Selection of this first repository site was mandated for early 1987, but program delays at DOE have pushed the decision back to March 1991. Despite this postponement and other schedule slips, the Department still aims to meet Congress's 1998 deadline. But states, Indian tribes, and environmentalists fear the site selection process will be compromised and want the start up date rolled back

Evaluation of waste crate counter

International Nuclear Information System (INIS)

Wachter, J.R.; Shaw, S.W.

1994-01-01

A novel nondestructive measurement system has been developed to perform combined gamma-ray, passive neutron, and active neutron analyses of radioactive waste packaged in large crates. The system will be used to examine low level and transuranic waste at the Waste Receiving and Processing facility at Westinghouse-Hanford Corp. Prior to delivery of the system, an extensive evaluation of its performance characteristics will be conducted. The evaluation is to include an assessment of the mechanical properties of the system, gamma-ray attenuation correction algorithms, instrument response as a function of source positions, performance of the high resolution gamma-ray detector for ''hot spot'' and isotopic analyses, active and passive neutron counter response, instrument sensitivity, matrix effects, and packaging effects. This report will discuss the findings of the evaluation program, to date, and indicate future directions for the program

Hot Flashes

Science.gov (United States)

Hot flashes Overview Hot flashes are sudden feelings of warmth, which are usually most intense over the face, neck and chest. Your skin might redden, as if you're blushing. Hot flashes can also cause sweating, and if you ...

ANSTO's waste forms for the 31. century

Energy Technology Data Exchange (ETDEWEB)

Vance, E.R.; Begg, B. D.; Day, R. A.; Moricca, S.; Perera, D. S.; Stewart, M. W. A.; Carter, M. L.; McGlinn, P. J.; Smith, K. L.; Walls, P. A.; Robina, M. La

2004-07-01

ANSTO waste form development for high-level radioactive waste is directed towards practical applications, particularly problematic niche wastes that do not readily lend themselves to direct vitrification. Integration of waste form chemistry and processing method is emphasised. Some longstanding misconceptions about titanate ceramics are dealt with. We have a range of titanate-bearing waste form products aimed at immobilisation of tank wastes and sludges, actinide-rich wastes, INEEL calcines and Na-bearing liquid wastes, Al-rich wastes arising from reprocessing of Al-clad fuels, Mo-rich wastes arising from reprocessing of U-Mo fuels, partitioned Cs-rich wastes, and {sup 99}Tc. Waste form production techniques cover hot isostatic and uniaxial pressing, sintering, and cold-crucible melting, and these are strongly integrated into waste form design. Speciation and leach resistance of Cs and alkalis in cementitious products and geo-polymers are being studied. Recently we have embarked on studies of candidate inert matrix fuels for Pu burning. We also have a considerable program directed at basic understanding of the waste forms in regard to crystal chemistry, dissolution behaviour in aqueous media, radiation damage effects and optimum processing techniques. (authors)

Process description and plant design for preparing ceramic high-level waste forms

International Nuclear Information System (INIS)

Grantham, L.F.; McKisson, R.L.; Guon, J.; Flintoff, J.F.; McKenzie, D.E.

1983-01-01

The ceramics process flow diagram has been simplified and upgraded to utilize only two major processing steps - fluid-bed calcination and hot isostatic press consolidating. Full-scale fluid-bed calcination has been used at INEL to calcine high-level waste for 18 y; and a second-generation calciner, a fully remotely operated and maintained calciner that meets ALARA guidelines, started calcining high-level waste in 1982. Full-scale hot isostatic consolidation has been used by DOE and commercial enterprises to consolidate radioactive components and to encapsulate spent fuel elements for several years. With further development aimed at process integration and parametric optimization, the operating knowledge of full-scale demonstration of the key process steps should be rapidly adaptable to scale-up of the ceramic process to full plant size. Process flowsheets used to prepare ceramic and glass waste forms from defense and commercial high-level liquid waste are described. Preliminary layouts of process flow diagrams in a high-level processing canyon were prepared and used to estimate the preliminary cost of the plant to fabricate both waste forms. The estimated costs for using both options were compared for total waste management costs of SRP high-level liquid waste. Using our design, for both the ceramic and glass plant, capital and operating costs are essentially the same for both defense and commercial wastes, but total waste management costs are calculated to be significantly less for defense wastes using the ceramic option. It is concluded from this and other studies that the ceramic form may offer important advantages over glass in leach resistance, waste loading, density, and process flexibility. Preliminary economic calculations indicate that ceramics must be considered a leading candidate for the form to immobilize high-level wastes

Identification, classification and management of industrial waste in Kavir steel complex according to the Bazel convention and RCRA

OpenAIRE

Mohammad Hasan Ehrampoush; Mohsen Hesami Arani; Mohammad Taghi Ghaneian; Asghar Ebrahimi; Masoud Shafiee

2016-01-01

Introduction: Requiring industries for implementing industrial waste management programs and planning for proper waste disposal is essential in order to achieve sustainable development. Therefore, industrial waste management program was done in Kavir Steel Complex, in Aran va Bidgol region to identify and classify industrial waste and also to present solutions for improving waste management. In this complex, production process is hot rolling steel and the product is rebar. Material and Me...

Economical utilization of hot water - an important precondition for an efficient utilization of waste heat in milk cooling

Energy Technology Data Exchange (ETDEWEB)

Kaiser, E; Pflug, C

1985-01-01

Indispensable both in the field of hydroecological and energy policies is the economical utilization of hot water. Hydroecological process analyses in specialized dairy cattle plants have shown that the specific mean annual abstraction of hot water (50/sup 0/C) may be reduced to 14 l per cow and per day. The proportionate contribution of different operational sectors and methods to arrive at the standards are pointed out. Economizing dairy cattly plants reducing hot water consumption as indicated and reaching average milking outputs of >= 1 l per cow and per day may thus bridge the summer season by heat recovery processes producing a sufficient quantity of hot water and allowing a shutdown of all heating units. At present the majority of dairy cattle plants cannot yet dispense with supplementary water during the remaining months. The hot water consumption rate is highest at the end of shifts. In double-shifted dairy cattle plants the estimated maximum hourly consumption amounts to 12 per cent of the average daily consumption. (orig.).

''FIXBOX'' - a new technique for the reliable conditioning of plutonium waste solutions

International Nuclear Information System (INIS)

Bruchertseifer, H.; Sommer, E.; Steinemann, M.; Bart, G.

1994-01-01

''FIXBOX'' - A new technique and facility for the conditioning of plutonium waste solutions has been developed and brought into operation in the Hot-laboratory at PSI, for the solidification of the waste from the research programmes. The facility is situated in glove-boxes for handling alpha activity and gamma-shielded for conditioning of fission product-containing waste. This report gives a brief description of the FIXBOX facility, the procedure and the first results of the cementation of plutonium waste solutions. As a result of this solidification, the actinide waste is homogeneous and strongly bound in the cement. The presence of gluconic acid and other complexing agents in the waste solution will not disturb this process. (author) figs., tabs., refs

Overview of mineral waste form development for the electrometallurgical treatment of spent nuclear fuel

International Nuclear Information System (INIS)

Pereira, C.; Lewis, M.A.; Ackerman, J.P.

1996-01-01

Argonne is developing a method to treat spent nuclear fuel in a molten salt electrorefiner. Wastes from this treatment will be converted into metal and mineral forms for geologic disposal. A glass-bonded zeolite is being developed to serve as the mineral waste form that will contain the fission products that accumulate in the electrorefiner salt. Fission products are ion exchanged from the salt into the zeolite A structure. The crystal structure of the zeolite after ion exchange is filled with salt ions. The salt-loaded zeolite A is mixed with glass frit and hot pressed. During hot pressing, the zeolite A may be converted to sodalite which also retains the waste salt. The glass-bonded zeolite is leach resistant. MCC-1 testing has shown that it has a release rate below 1 g/(m 2 day) for all elements

Survey of concrete waste forms

International Nuclear Information System (INIS)

Moore, J.G.

1981-01-01

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

Low-risk alternative waste forms for problematic high-level and long-lived nuclear wastes

International Nuclear Information System (INIS)

Stewart, M.W.A.; Begg, B.D.; Moricca, S.; Day, R.A.

2006-01-01

Full text: The highest cost component the nuclear waste clean up challenge centres on high-level waste (HLW) and consequently the greatest opportunity for cost and schedule savings lies with optimising the approach to HLW cleanup. The waste form is the key component of the immobilisation process. To achieve maximum cost savings and optimum performance the selection of the waste form should be driven by the characteristics of the specific nuclear waste to be immobilised, rather than adopting a single baseline approach. This is particularly true for problematic nuclear wastes that are often not amenable to a single baseline approach. The use of tailored, high-performance, alternative waste forms that include ceramics and glass-ceramics, coupled with mature process technologies offer significant performance improvements and efficiency savings for a nuclear waste cleanup program. It is the waste form that determines how well the waste is locked up (chemical durability), and the number of repository disposal canisters required (waste loading efficiency). The use of alternative waste forms for problematic wastes also lowers the overall risk by providing high performance HLW treatment alternatives. The benefits tailored alternative waste forms bring to the HLW cleanup program will be briefly reviewed with reference to work carried out on the following: The HLW calcines at the Idaho National Laboratory; SYNROC ANSTO has developed a process utilising a glass-ceramic combined with mature hot-isostatic pressing (HIP) technology and has demonstrated this at a waste loading of 80 % and at a 30 kg HIP scale. The use of this technology has recently been estimated to result in a 70 % reduction in waste canisters, compared to the baseline borosilicate glass technology; Actinide-rich waste streams, particularly the work being done by SYNROC ANSTO with Nexia Solutions on the Plutonium-residues wastes at Sellafield in the UK, which if implemented is forecast to result in substantial

HOT 2011

DEFF Research Database (Denmark)

Lund, Henriette Romme

En undersøgelse af, hvad der er hot - og burde være hot på læseområdet. I undersøgelsen deltager 21 læsekyndige fra praksisfeltet, professionshøjskolerne og forskningsområdet.......En undersøgelse af, hvad der er hot - og burde være hot på læseområdet. I undersøgelsen deltager 21 læsekyndige fra praksisfeltet, professionshøjskolerne og forskningsområdet....

Nuclear wastes: status and perspectives; Les Dechets nucleaires: etat des lieux et perspectives

Energy Technology Data Exchange (ETDEWEB)

Barre, B.; Bazile, F.; Bonin, B.; Marvy, A.; Ouzounian, G.; Wizsenfeld, B.

2011-07-01

Radioactive wastes are considered by the large public as a problem which mortgages the future of nuclear industry. Today, the main principles of radioactive wastes management are defined: separation, conditioning, storage and disposal are the main elements of a strategy allowing to protect the man and the environment from the impacts of the inevitable ultimate wastes. In a period of debate concerning the future of these wastes, this book aims at supplying to the public the elements which will allow him to build his own opinion about this hot topic. (J.S.)

Process and device for decontamination of the waste gas of the fuel circuit of a fusion reactor from tritium and/or deuterium in waste gas containing them in chemically bound form

International Nuclear Information System (INIS)

Penzhorn, R.D.; Glugla, M.

1987-01-01

The invention concerns a process and a device for the decontamination of the wate gases of the fuel circuit of a fusion reactor from tritum and/or deuterium in waste gas containing them in chemically bound form, in which the waste gas is taken over an oxidation catalyst and then over a hot metal bed, tritium and/or deuterium is released from its compounds, separated from the waste gas and is returned to the fuel circuit. The process is intended to prevent losses of tritum and/or deuterium by permeation and the high loading of the hot metal getter materials, as occurs in the previously known corresponding process, and to avoid the formation of nitrogen oxides. This is achieved by: a) The catalytic oxidation reaction being carried out at a temperature of 200 0 C to 300 0 C. b) The gas mixture then being brought into contact with a hot metal bed at 200 0 C to 300 0 C to remove the remaining O 2 and for the selective conversion of the proportion of water into the hydrogen isotope. c) The gas mixture being brought into contact with a diaphragm made of palladium or a palladium-silver alloy at 400 0 C to 450 0 C to decompose the ammonia, all the released hydrogen isotope being passed through the diaphragm, separated from the remaining waste gas flow and removed. (orig.) [de

Mechanistic Evaluation of the Effect of Calcium Carbide Waste on ...

African Journals Online (AJOL)

OLUWASOGO

amounts of wastes from ceramic, steel industry and coal-fired power plants every year. ... slag dust, hydrated lime, hydraulic cement, fly ash, loess or other suitable ..... (2011). The. Effect of using Glass Powder Filler on Hot Asphalt Concrete.

Vaporizing Flow in Hot Fractures: Observations from Laboratory Experiments

International Nuclear Information System (INIS)

Kneafsey, T.; Pruess, K.

1998-01-01

Understanding water seepage in hot fractured rock is important in a number of fields including geothermal energy recovery and nuclear waste disposal. Heat-generating high-level nuclear waste packages which will be emplaced in the partially saturated fractured tuffs at the potential high-level nuclear waste repository at Yucca Mountain, Nevada, if it becomes a high-level nuclear waste repository, will cause significant impacts on moisture distribution and migration. Liquid water, which occupies anywhere from 30 to 100% of the porespace, will be vaporized as the temperature reaches the boiling temperature. Flowing primarily in fractures, the vapor will condense where it encounters cooler rock, generating mobile water. This water will flow under gravitational and capillary forces and may flow back to the vicinity of the emplaced waste where it may partially escape vaporization. Water flowing down (sub-) vertical fractures may migrate considerable distances through fractured rock that is at above-boiling temperatures; thus, flowing condensate may contact waste packages, and provide a pathway for the transport of water-soluble radionuclides downward to the saturated zone. Thermally-driven flow processes induced by repository heat may be as important or even more important for repository performance than natural infiltration. For a nominal thermal loading of 57 kW/acre, vaporization may generate an average equivalent percolation flux from condensate of 23.1 mm/yr over 1,000 years, and 5.2 mm/yr over 10,000 years. These numbers are comparable to or larger than current estimates of net infiltration at Yucca Mountain. This condensate, which is generated in the immediate vicinity (meters) of the waste packages, will likely have a larger impact on waste package and repository performance than a similar amount of water introduced at the land surface

Physical and mechanical properties of parallel strand lumber made from hot pre-pressed long strand oil palm trunk waste

Science.gov (United States)

Fridiyanti, Inayah; Massijaya, M. Y.

2018-03-01

This research was focused on the utilization of oil palm trunk waste as a Parallel Strand Lumber (PSL) raw material. This research aimed to analyze the effect of adhesive types and glue spreads to the physical and mechanical properties of PSL. The adhesive types used were isocyanate and urea formaldehyde adhesives. The glue spreads used were 150 g/m2 and 300 g/m2. The research results showed that the moisture content of PSL ranged from 9.30% to 11.80%. The PSL density ranged from 0.64 to 0.78 g/cm3. The volume shrinkage ranged from 5.69 to 7.17%. Modulus of Elasticity (MOE) parallel to the grain and edge side ranged from 51.6 × 103 to 98.3 × 103 kg/cm2, and 62.1 × 103 to 99.9 × 103 kg/cm2, respectively. The Modulus of Rupture (MOR) parallel to the grain and edge side ranged from 269 to 724 kg/cm2 and 342 to 728 kg/cm2, respectively. The PSL hardness perpendicular to the grain, parallel to the grain and the edge side ranged from 135 to 300 kg/cm2, 87 to 321 kg/cm2, and 128 to 251 kg/cm2, respectively. The compressive strength ranged from 181 to 231 kg/cm2. The best adhesive and glue spreads of PSL was isocyanate with glue spread 300 g/m2. PSL made from hot pre-pressed long strand of oil palm trunk waste bonded by isocyanate fulfill JAS 1152: 2007. However, those of bonded by urea formaldehyde failed to fulfill the standard. The physical and mechanical properties of PSL made from oil palm trunk were better compared to those of solid oil palm trunk.

Treatment of alpha-bearing combustible wastes using acid digestion

International Nuclear Information System (INIS)

Lerch, R.E.; Allen, C.R.; Blasewitz, A.G.

1978-01-01

Acid digestion has been developed at the Hanford Engineering Development Laboratory (HEDL) in Richland, Washington to reduce the volume of combustible nuclear waste materials, while converting them to an inert, noncombustible residue. A 100 kg/day test unit has recently been constructed to demonstrate the process using radioactively contaminated combustible wastes. The unit, called the Radioactive Acid Digestion Test Unit (RADTU) was completed in September 1977 and is currently undergoing cold shakedown tests. Hot operation is expected in May 1978. Features of RADTU include: storage and transfer station for incoming wastes, a feed preparation station, an extrusion feed mechanism for transfer of the waste to the acid digester, the acid digester a residue recovery system, and an off-gas treatment system

Treatment of alpha-bearing combustible wastes using acid digestion

International Nuclear Information System (INIS)

Lerch, R.E.; Allen, C.R.; Blasewitz, A.G.

1977-11-01

Acid digestion has been developed at the Hanford Engineering Development Laboratory (HEDL) in Richland, Washington to reduce the volume of combustible nuclear waste materials, while converting them to an inert, noncombustible residue. A 100 kg/day test unit has recently been constructed to demonstrate the process using radioactively contaminated combustible wastes. The unit, called the Radioactive Acid Digestion Test Unit (RADTU) was completed in September 1977 and is currently undergoing cold shakedown tests. Hot operation is expected in May 1978. Features of RADTU include: storage and transfer station for incoming wastes, a feed preparation station, an extrusion feed mechanism for transfer of the waste to the acid digester, the acid digester, a residue recovery system, and an off-gas treatment system

Decommissioning of the Risoe Hot Cell facility

International Nuclear Information System (INIS)

Carlsen, H.

1991-02-01

The Hot Cell facility at Risoe has been in active use since 1964. During the years several types of nuclear fuels have been handled and examined: test reactor fuel pins from the Danish reactor DR3, the Norwegian Halden reactor, etc; power reactor fuel pins from several foreign reactors, including plutonium enriched pins; HTGR fuel from the Dragon reactor. All kinds of physical and chemical non-destructive and destructive post irradiation examinations have been performed. Besides, different radiotherapy sources have been produced, mainly cobalt sources. The general object of the decommissioning programme for the Hot Cell facility was to obtain a safe condition for the total building that does not require the special safety provisions. The hot cell building will be usable for other purposes after decommissioning. The facilicy comprised six concrete cells, lead cells, glove boxes, a shielded unit for temporary storage of waste, frogman area, decontamination areas, workshops, various installations of importance for safe operation of the plant, offices, etc. The tasks comprised e.g. removal of all irradiated fuel items, removal of other radioactive items, removal of contaminated equipment, and decontamination of all the cells and rooms. The goal was to decontaminate all the concrete cells to a degree where no loose contamination exists in the cells, and where the radiation level is so low, that total removal of the cell structures can be done at any time in the future without significant dose commitments. (AB)

Partitioning-transmutation technology: a potential future nuclear waste management option

International Nuclear Information System (INIS)

Nakayama, S.; Morita, Y.; Nishihara, K.; Oigawa, H.

2005-01-01

Partitioning-transmutation technology (PT) will produce radioactive wastes of different physical and chemical properties and in different amounts from those generated in the current nuclear fuel cycle. To assess quantitatively the effects of PT on waste disposal, we first analyzed the amounts of the PT wastes, assumed conditioning for each type of the waste, and then made an attempt to estimate the repository area for disposal of the PT wastes. The properties of the hot Sr-Cs waste form are controlling factors in determining the size of the geologic repository. The disposal area could be reduced if the Sr-Cs fraction is disposed in a different subsurface repository or by long-term storage of the waste under institutional control. Disposal in a subsurface repository was found to comply with the Japanese law in terms of radioactivity constraint, through a performance assessment for disposal of the Sr-Cs fraction. (authors)

Development of the destruction technology for radioactive organic solid wastes

Energy Technology Data Exchange (ETDEWEB)

Oh, Won Zin; Park, H.S.; Lee, K.W. [and others

1999-04-01

The followings were studied through the project entitled 'Technology development for nuclear fuel cycle waste treatment'. 1. Organic waste decomposition technology development A. Destruction technology for organic wastes using Ag(2)-mediated electrochemical oxidation B. Recovery and regeneration technology for the spent chemicals used in the MEO process 2. Radioactive metal waste recycling technology A. Surface decontamination processes B. Decontamination waste treatment technology 3. Volume reduction technology nuclear fuel cycle (NFC) technology A. Estimation of the amount of radwastes and the optimum volume reduction methodology of domestic NFC B. Pretreatment of spent fuel cladding by electrochemical decontamination C. Hot cell process technology for the treatment of NFC wastes 4. Design and fabrication of the test equipment of volume reduction and reuse of alpha contaminated wastes 5. Evaluation on environmental compatibility of NFC A. Development of evaluation methodology on environmental friendliness of NFC B. Residual activity assessment of recycling wastes. (author). 321 refs., 54 tabs., 183 figs.

Development of the destruction technology for radioactive organic solid wastes

International Nuclear Information System (INIS)

Oh, Won Zin; Park, H.S.; Lee, K.W.

1999-04-01

The followings were studied through the project entitled 'Technology development for nuclear fuel cycle waste treatment'. 1. Organic waste decomposition technology development A. Destruction technology for organic wastes using Ag(2)-mediated electrochemical oxidation B. Recovery and regeneration technology for the spent chemicals used in the MEO process 2. Radioactive metal waste recycling technology A. Surface decontamination processes B. Decontamination waste treatment technology 3. Volume reduction technology nuclear fuel cycle (NFC) technology A. Estimation of the amount of radwastes and the optimum volume reduction methodology of domestic NFC B. Pretreatment of spent fuel cladding by electrochemical decontamination C. Hot cell process technology for the treatment of NFC wastes 4. Design and fabrication of the test equipment of volume reduction and reuse of alpha contaminated wastes 5. Evaluation on environmental compatibility of NFC A. Development of evaluation methodology on environmental friendliness of NFC B. Residual activity assessment of recycling wastes. (author). 321 refs., 54 tabs., 183 figs

Improved polyphase ceramic form for high-level defense nuclear waste

International Nuclear Information System (INIS)

Harker, A.B.; Morgan, P.E.D.; Clarke, D.R.; Flintoff, J.J.; Shaw, T.M.

1983-01-01

An improved ceramic nuclear waste form and fabrication process have been developed using simulated Savannah River Plant defense high-level waste compositions. The waste form provides flexibility with respect to processing conditions while exhibiting superior resistance to ground water leaching than other currently proposed forms. The ceramic, consolidated by hot-isostatic pressing at 1040 0 C and 10,000 psi, is composed of six major phases, nepheline, zirconolite, a murataite-type cubic phase, magnetite-type spinel, a magnetoplumbite solid solution, and perovskite. The waste form provides multiple crystal lattice sites for the waste elements, minimizes amorphous intergranular material, and can accommodate waste loadings in excess of 60 wt %. The fabrication of the ceramic can be accomplished with existing manufacturing technology and eliminates the effects of radionuclide volatilization and off-gas induced corrosion experienced with the molten processes for vitreous form production

Dismantling of a hot cell-block and the treatment of the produced concrete bars

International Nuclear Information System (INIS)

Rompf, U.; Brielmayer, M.; Graf, A.; Stutz, U.; Ambos, F.

2003-01-01

A building with hot cells had been operated in Karlstein/Main from 1968 to 1989 in order to perform check-ups at radiated fuel rods and nuclear components. The operation of the system was stopped after an operation period of approximately 20 years. The core part of the building to be disassembled is a U-shaped hot cell-block with nine individual cells, partly consisting of heavy reinforced concrete, located in the ground floor (fig. 1 and fig. 2). The major part of the cells was covered with 10 mm steel plate and provided with approx. 1,400 openings of all different kinds. The wall thickness of the cells was between 0.90 m and 1.10 m. Under these conditions a successful decontamination at the ''existing building structure'' was not possible. Therefore, the non-supporting structures of the hot cell-block were removed in individual blocks by means of sawing and the remaining walls and floors were peeled by using the diamond rope sawing technique. The dismantling took 17 months. A re-treatment of the produced concrete blocks (235 blocks, approx. 970 Mg) to reduce the radioactive waste to a minimum was performed at the Research Centre Karlsruhe, Central Decontamination Department (HDB). The Target of the concrete bar treatment at HDB is to reduce the volume of radioactive waste to a minimum and to add the major part of the concrete bars to harmless utilisation. To achieve the same, initially the more contaminated parts of the bars without openings, such as tubes, cable or ventilating shafts, are removed by means of wire cutting and packed into a KONRAD-Container as radioactive waste. The remaining bar is decontaminated by means of sandblasting and afterwards, following successful release measurement, released from the scope of the regulations under the Atomic Energy. Bars with openings are crushed into small pieces by means of the remote-controlled chisel excavator, in order to separate the individual kinds of material. The rubble is packed into drums and measured by

Can nuclear waste be stored safely at Yucca mountain?

International Nuclear Information System (INIS)

Whipple, C.G.

1996-01-01

In 1987 the federal government narrowed to one its long-term options for disposing of nuclear waste: storing it permanently in a series of caverns excavated out of the rock deep below Yucca mountain in southern Nevada. Whether it makes sense at this time to dispose permanently of spent fuel and radioactive waste in a deep geologic repository is hotly disputed. But the Nuclear Waste Policy Act amendements of 1987 decree that waste be consolidated in Yucca Mountain if the mountain is found suitable. Meanwhile the spent fuel continues to pile up across the country, and 1998 looms, adding urgency to the question: What can science tell us about the ability of the mountain to store nuclear waste safely? This paper discusses this issue and describes how studies of the mountain's history and geology can contribute useful insights but not unequivocal conclusions

Simultaneous heat integration and techno-economic optimization of Organic Rankine Cycle (ORC) for multiple waste heat stream recovery

International Nuclear Information System (INIS)

Yu, Haoshui; Eason, John; Biegler, Lorenz T.; Feng, Xiao

2017-01-01

In the past decades, the Organic Rankine Cycle (ORC) has become a promising technology for low and medium temperature energy utilization. In refineries, there are usually multiple waste heat streams to be recovered. From a safety and controllability perspective, using an intermedium (hot water) to recover waste heat before releasing heat to the ORC system is more favorable than direct integration. The mass flowrate of the intermediate hot water stream determines the amount of waste heat recovered and the final hot water temperature affects the thermal efficiency of ORC. Both, in turn, exert great influence on the power output. Therefore, the hot water mass flowrate is a critical decision variable for the optimal design of the system. This study develops a model for techno-economic optimization of an ORC with simultaneous heat recovery and capital cost optimization. The ORC is modeled using rigorous thermodynamics with the concept of state points. The task of waste heat recovery using the hot water intermedium is modeled using the Duran-Grossmann model for simultaneous heat integration and process optimization. The combined model determines the optimal design of an ORC that recovers multiple waste heat streams in a large scale background process using an intermediate heat transfer stream. In particular, the model determines the optimal heat recovery approach temperature (HRAT), the utility load of the background process, and the optimal operating conditions of the ORC simultaneously. The effectiveness of this method is demonstrated with a case study that uses a refinery as the background process. Sensitivity of the optimal solution to the parameters (electricity price, utility cost) is quantified in this paper. - Highlights: • A new model for Organic Rankine cycle design optimization is presented. • Process heat integration and ORC are considered simultaneously. • Rigorous equation oriented models of the ORC are used for accurate results. • Impact of working

Management of historical waste from research reactors: the Dutch experience

International Nuclear Information System (INIS)

Van Heek, Aliki; Metz, Bert; Janssen, Bas; Groothuis, Ron

2013-01-01

Most radioactive waste emerges as well-defined waste streams from operating power reactors. The management of this is an on-going practice, based on comprehensive (IAEA) guidelines. A special waste category however consists of the historical waste from research reactors, mostly originating from various experiments in the early years of the nuclear era. Removal of the waste from the research site, often required by law, raises challenges: the waste packages must fulfill the acceptance criteria from the receiving storage site as well as the criteria for nuclear transports. Often the aged waste containers do not fulfill today's requirements anymore, and their contents are not well documented. Therefore removal of historical waste requires advanced characterization, sorting, sustainable repackaging and sometimes conditioning of the waste. This paper describes the Dutch experience of a historical waste removal campaign from the Petten High Flux research reactor. The reactor is still in operation, but Dutch legislation asks for central storage of all radioactive waste at the COVRA site in Vlissingen since the availability of the high- and intermediate-level waste storage facility HABOG in 2004. In order to comply with COVRA's acceptance criteria, the complex and mixed inventory of intermediate and low level waste must be characterized and conditioned, identifying the relevant nuclides and their activities. Sorting and segregation of the waste in a Hot Cell offers the possibility to reduce the environmental footprint of the historical waste, by repackaging it into different classes of intermediate and low level waste. In this way, most of the waste volume can be separated into lower level categories not needing to be stored in the HABOG, but in the less demanding LOG facility for low-level waste instead. The characterization and sorting is done on the basis of a combination of gamma scanning with high energy resolution of the closed waste canister and low

HOT 2014

DEFF Research Database (Denmark)

Lund, Henriette

Undersøgelse af, hvad der er hot - og hvad der burde være hot på læseområdet med 21 læsekyndige. Undersøgelsen er gennemført siden 2010. HOT-undersøgelsen er foretaget af Nationalt Videncenter for Læsning - Professionshøjskolerne i samarb. med Dansklærerforeningen...

Analysis on energy-saving path of rural buildings in hot summer and cold winter zone

Science.gov (United States)

Huang, Mingqiang; Li, Jinheng

2018-02-01

Since the reform and opening policy, the construction of rural area in China has become more and more important. The idea of establishing green villages needs to be accepted and recognized by the public. The hot summer and cold winter zone combines two contradictory weather conditions that is cold winter and hot summer. So the living conditions are limited. In response to this climate, residents extensively use electric heaters or air conditioning to adjust the indoor temperature, resulting in energy waste and environmental pollution. In order to improve the living conditions of residents, rural area energy conservation has been put on the agenda. Based on the present situation and energy consumption analysis of the rural buildings in the hot summer and cold winter zone, this article puts forward several energy saving paths from government, construction technology and so on

The OTD Robotics Waste Minimization Program

International Nuclear Information System (INIS)

Couture, S.A.

1992-04-01

The danger to human health and safety posed by exposure to transuranic (TRU) and Pu contaminated materials necessitates remote processing in confined environments. Currently these operations are carried out in gloveboxes and hot-cells by human operators using lead- lined gloves or teleoperated manipulators. Protective clothing worn by operators during gloved operations has contributed significantly to the waste problems currently facing site remediators. The DOE Environmental Restoration and Waste Management (ER/WM) Program is in the process of developing and demonstrating technologies to assist in the remediation of sites that have accumulated wastes generated using these processes over the past five decades. Recognizing that continued use of existing production, recovery and waste treatment systems will compound the remediation problem, DOE has made a commitment to waste minimization. To reduce waste generation during weapons production and waste processing operations, automated processes are being developed and demonstrated for use in future DOE processing facilities as part of OTD's Robotics Technology Development Program. These technologies are currently being applied to pyrochemical processing systems to demonstrate conversion of plutonium oxide to metal. However, these technologies are expected to have applications in a variety of waste processing systems including those used to treat high-level tank wastes, buried wastes requiring remote processing, mixed wastes, and unknown hazardous materials. In addition to reducing the future waste burden of DOE, automated processes are an effective way to comply with existing and anticipated federal, state, and local regulations related to personal health and safety and the health of the environment

Optimization of Waste Collection System Using Underground Containers with Source Separation Plan (Case Study: Zone 3 of Yazd Municipality, Iran

Directory of Open Access Journals (Sweden)

Maryam Morakabatchian

2017-12-01

Conclusion: Optimization of urban waste collection system using underground containers for wet waste and the use of temporary stations of dry wastes, considering the significant economic, environmental and aesthetic advantages can be considered as an appropriate option in Iranian cities especially in areas with hot and humid weather such as Yazd.

Development of remote electrochemical decontamination for hot cell applications

International Nuclear Information System (INIS)

Turner, A.D.; Lain, M.J.; Fletcher, P.A.; Dawson, R.K.; Pottinger, J.S.

1989-01-01

The primary aim of the programme is to develop and evaluate remote electrochemical decontamination systems for metal surfaces. The bulk of the waste volume should be reduced to a reuse or low-level waste disposal category, while concentrating most of the activity in a small volume suitable for immobilisation. The goal of the development programme is to test these techniques in both alpha-active and alpha-beta-gamma hot cells in order to ascertain their usefulness as a component of an overall decommissioning strategy. As a result of the radiological environment, particular emphasis will be placed on remote operation in order to reduce occupational radiation exposure. Two types of technique based on the electrochemical dissolution of thin surface layers of the substrate will be investigated: immersion of small items in tanks for electroetching and in situ electropolishing. In both cases, reagents will be chosen with their subsequent disposal in mind. (Author)

Immobilization of high-level wastes into sintered glass: 2

International Nuclear Information System (INIS)

Bevilacqua, A.M.; Russo, D.O.; Messi de Bernasconi, N.; Audero, M.A.

1987-01-01

High level radioactive wastes are immobilized into borosilicate glasses. Experiences with the variety VG 98/12 (SiO 2 , TiO 2 , Al 2 O 3 , B 2 O 3 , MgO, CaO, Na 2 O) are described. The pressing was performed in a matrix of 12.7 mm diameter, the walls of which were lubricated with sterotex dissolved in Cl 4 C. The sintering was made in an horizontal electric furnace in air atmosphere at temperatures between 500 and 600 deg C. It was observed that the maximum density occurs at 605 deg C. Comparing both the hot and the cold pressing process, it is concluded that: 1) In spite of requiring more complex equipment the hot pressing process has the advantage that lower pressures are applied, with the consequent obtainment of waste blocks with larger diameters, and 2) it is advisable that pressing process should be performed in the definitive can. (M.E.L.) [es

Bioprocessing of a stored mixed liquid waste

Energy Technology Data Exchange (ETDEWEB)

Wolfram, J.H.; Rogers, R.D. [Idaho National Engineering Lab., Idaho Falls, ID (United States); Finney, R. [Mound Applied Technologies, Miamisburg, OH (United States)] [and others

1995-12-31

This paper describes the development and results of a demonstration for a continuous bioprocess for mixed waste treatment. A key element of the process is an unique microbial strain which tolerates high levels of aromatic solvents and surfactants. This microorganism is the biocatalysis of the continuous flow system designed for the processing of stored liquid scintillation wastes. During the past year a process demonstration has been conducted on commercial formulation of liquid scintillation cocktails (LSC). Based on data obtained from this demonstration, the Ohio EPA granted the Mound Applied Technologies Lab a treatability permit allowing the limited processing of actual mixed waste. Since August 1994, the system has been successfully processing stored, {open_quotes}hot{close_quotes} LSC waste. The initial LSC waste fed into the system contained 11% pseudocumene and detectable quantities of plutonium. Another treated waste stream contained pseudocumene and tritium. Data from this initial work shows that the hazardous organic solvent, and pseudocumene have been removed due to processing, leaving the aqueous low level radioactive waste. Results to date have shown that living cells are not affected by the dissolved plutonium and that 95% of the plutonium was sorbed to the biomass. This paper discusses the bioprocess, rates of processing, effluent, and the implications of bioprocessing for mixed waste management.

Establishment of a facility for intrusive characterization of transuranic waste at the Nevada Test Site

International Nuclear Information System (INIS)

Foster, B.D.; Musick, R.G.; Pedalino, J.P.; Cowley, J.L.; Karney, C.C.; Kremer, J.L.

1998-01-01

This paper describes design and construction, project management, and testing results associated with the Waste Examination Facility (WEF) recently constructed at the Nevada Test Site (NTS). The WEF and associated systems were designed, procured, and constructed on an extremely tight budget and within a fast track schedule. Part 1 of this paper focuses on design and construction activities, Part 2 discusses project management of WEF design and construction activities, and Part 3 describes the results of the transuranic (TRU) waste examination pilot project conducted at the WEF. In Part 1, the waste examination process is described within the context of Waste Isolation Pilot Plant (WIPP) characterization requirements. Design criteria are described from operational and radiological protection considerations. The WEF engineered systems are described. These systems include isolation barriers using a glove box and secondary containment structure, high efficiency particulate air (HEPA) filtration and ventilation systems, differential pressure monitoring systems, and fire protection systems. In Part 2, the project management techniques used for ensuring that stringent cost/schedule requirements were met are described. The critical attributes of these management systems are described with an emphasis on team work. In Part 3, the results of a pilot project directed at performing intrusive characterization (i.e., examination) of TRU waste at the WEF are described. Project activities included cold and hot operations. Cold operations included operator training, facility systems walk down, and operational procedures validation. Hot operations included working with plutonium contaminated TRU waste and consisted of waste container breaching, waste examination, waste segregation, data collection, and waste repackaging

Gas flow in and out of a nuclear waste container

International Nuclear Information System (INIS)

Zwahlen, E.D.; Pigford, T.H.; Chambre, P.L.; Lee, W.W.L.

1989-05-01

We analyze the flow of gases out of and into a high-level-waste container in the unsaturated tuff of Yucca Mountain. Containers are expected to fail eventually by localized cracks and penetrations. Even though the penetrations may be small, argon gas initially in the hot container can leak out. As the waste package cools, the pressure inside the container can become less than atmospheric, and air can leak in. 14 C released from the hot fuel-cladding surface can leak out of penetrations, and air inleakage can mobilize additional 14 C and other volatile radioactive species as it oxidizes the fuel cladding and the spent fuel. In an earlier paper we studied the gas flow through container penetrations occurring at the time of emplacement. Here we analyze the flow of gas for various penetration sizes occurring at 300 years. 3 refs., 2 figs

Study of waste-heat recovery and utilization at the Farmington Municipal Power Plant. Final report, December 1, 1980-June 30, 1981

Energy Technology Data Exchange (ETDEWEB)

Leigh, G.G.; Edgel, W.R.; Feldman, K.T. Jr.; Moss, E.J.

1982-03-01

An examination was made of the technical and economc feasibility of utilizing waste heat from the Farmington Municipal Power Plant. First, the production cycles of the natural-gas-fired plant were assessed to determine the quantity and quality of recoverable waste heat created by the plant during its operation. Possibilities for utilizing waste heat from the exhaust gases and the cooling water were then reviewed. Hot water systems that can be used to retrieve heat from hot flue gases were investigated; the heated water can then be used for space heating of nearby buildings. The potential use of waste heat to operate a refrigeration plant was also analyzed. The use of discharged cooling water for hydroelectric generation was studied, as well as its application for commercial agricultural and aquaculture enterprises.

HOT 2010

DEFF Research Database (Denmark)

Lund, Henriette Romme

En undersøgelse af, hvad der er hot - og burde være hot på læseområdet. I undersøgelsen deltager en række læsekyndige fra praksisfeltet, professionshøjskolerne og forskningsområdet. Undersøgelsen er gentaget hvert år siden 2010.......En undersøgelse af, hvad der er hot - og burde være hot på læseområdet. I undersøgelsen deltager en række læsekyndige fra praksisfeltet, professionshøjskolerne og forskningsområdet. Undersøgelsen er gentaget hvert år siden 2010....

HOT 2013

DEFF Research Database (Denmark)

Lund, Henriette Romme

En undersøgelse af, hvad der er hot - og burde være hot på læseområdet. I undersøgelsen deltager en række læsekyndige fra praksisfeltet, professionshøjskolerne og forskningsområdet. Undersøgelsen er gentaget hvert år siden 2010.......En undersøgelse af, hvad der er hot - og burde være hot på læseområdet. I undersøgelsen deltager en række læsekyndige fra praksisfeltet, professionshøjskolerne og forskningsområdet. Undersøgelsen er gentaget hvert år siden 2010....

Suppression of sawtooth oscillations due to hot electrons and hot ions

International Nuclear Information System (INIS)

Zhang, Y.Z.; Berk, H.L.

1989-01-01

The theory of m = 1 kink mode stabilization is discussed in the presence of either magnetically trapped hot electrons or hot ions. For instability hot ion requires particles peaked inside the q = 1 surface, while hot electrons require that its pressure profile be increasing at the q = 1 surface. Experimentally observed sawtooth stabilization usually occurs with off-axis heating with ECRH and near axis heating with ICRH. Such heating may produce the magnetically trapped hot particle pressure profiles that are consistent with theory. 17 refs., 2 figs

Radioactive wastes

International Nuclear Information System (INIS)

Straub, C.P.

1975-01-01

A review is presented on the environmental behavior of radioactive wastes. The management of high-level wastes and waste disposal methods were discussed. Some topics included were ore processing, coagulation, absorption and ion exchange, fixation, ground disposal, flotation, evaporation, transmutation and extraterrestrial disposal. Reports were given of the 226 Ra, 224 Ra and tritium activity in hot springs, 90 Sr concentrations in the groundwater and in White Oak Creek, radionuclide content of algae, grasses and plankton, radionuclides in the Danube River, Hudson River, Pacific Ocean, Atlantic Ocean, Lake Michigan, Columbia River and other surface waters. Analysis showed that 239 Pu was scavenged from Lake Michigan water by phytoplankton and algae by a concentration factor of up to 10,000. Benthic invertebrates and fish showed higher 239 Pu concentrations than did their pelagic counterparts. Concentration factors are also given for 234 Th, 60 Co, Fe and Mr in marine organisms. Two models for predicting the impact of radioactivity in the food chain on man were mentioned. In an accidental release from a light-water power reactor to the ocean, the most important radionuclides discharged were found to be 90 Sr, 137 Cs, 239 Pu and activation products 65 Zr, 59 Fe, and 95 Zr

LASER ABLATION-INDUCTIVELY COUPLED PLASMA-ATOMIC EMISSION SPECTROSCOPY STUDY AT THE 222-S LABORATORY USING HOT-CELL GLOVE BOX PROTOTYPE SYSTEM

International Nuclear Information System (INIS)

Lockrem, L.L.; Owens, J.W.; Seidel, C.M.

2009-01-01

This report describes the installation, testing and acceptance of the Waste Treatment and Immobilization Plant procured laser ablation-inductively coupled plasma-atomic emission spectroscopy (LA-ICP-AES) system for remotely analyzing high-level waste samples in a hot cell environment. The 2005-003; ATS MP 1027, Management Plan for Waste Treatment Plant Project Work Performed by Analytical Technical Services. The APD group at the 222-S laboratory demonstrated acceptable turnaround time (TAT) and provide sufficient data to assess sensitivity, accuracy, and precision of the LA-ICP-AES method

A novel method of utilization of hot dip galvanizing slag using the heat waste from itself for protection from radiation.

Science.gov (United States)

Dong, Mengge; Xue, Xiangxin; Kumar, Ashok; Yang, He; Sayyed, M I; Liu, Shan; Bu, Erjun

2018-02-15

A novel, unconventional, low cost, eco-friendly and effective shielding materials have been made utilizing the hot dip galvanizing slag using the heat waste from itself, thereby saving the natural resources and preventing the environmental pollution. SEM-EDS of shielding materials indicates that the other elements are distributed in Zn element. The mass attenuation properties of shielding materials were measured using a narrow beam geometrical setup at 0.662MeV, 1.17MeV and 1.33MeV. The half value thickness layer, effective atomic number, and electron density were used to analyze the shielding performance of the materials. The EBFs and EABFs for the prepared shielding materials were also studied with incident photon energy for penetration depths upto 40mfp. The shielding effectiveness has been compared with lead, iron, zinc, some standard shielding concretes, different glasses and some alloys. The shielding effectiveness of the prepared samples is almost found comparable to iron, zinc, selected alloys and glasses while better than some standard shielding concretes. In addition, it is also found that the bending strength of all shielding materials is more than 110MPa. Copyright © 2017 Elsevier B.V. All rights reserved.

Management of hospital radioactive wastes

International Nuclear Information System (INIS)

Houy, J.C.; Rimbert, J.C.; Bouvet, C.; Laugle, S.

1997-01-01

The hospital radioactive wastes are of three types: solid, liquid and gaseous. Prior to final evacuation all these wastes are checked by a detector the threshold of which is lower than the standard. This system allows detecting activities very low under the daily recommended threshold of 37 kBq (1μ Ci), for the group II. In metabolic radiotherapy the unsealed sources of iodine 131 will form mainly the wastes arising from the rooms contaminated by the patient himself. In this service anything touching the patient's room most by systematically checked. All the rooms are provided with toilette with two compartments, one connected traditionally to the sewerage system for faeces and the other coupled to tanks for urine storing. The filled reservoirs waits around 10 month span prior to being emptied, after checking, into the sewerage system. The volume activity most be lower than 7 Bq per liter (standard). For the hot labs, injection room and in-vitro lab, the liquid waste retrieved from dedicated stainless sinks are stored in storage tanks and will waits for 2 years before evacuation. The undies coming from the metabolic radiotherapy service are possible contaminated by the patient sheets, pillow cases, etc. These undies freshly contaminated may be contaminating if the contamination is non fixated. All the undies coming from this service are checked like all the wastes by means of the fixed detector. For the solid wastes two evacuation channels are possible: the urban garbage repository for household wastes and the Brest waste repository for hospital wastes. For the liquid waste arising for urines, used washing water, etc, the evacuation will be done towards city sewerage system after storing or dilution. Concerning the liquid wastes presenting chemical risks, they will be evacuated in cans by NETRA. Concerning the gaseous wastes, trapped on active carbon filters, they will be handled like solid wastes and will be directed to the waste repository of Brest. The other

Performance Characteristics of Hero's Turbine Using Hot Water as a Working Fluid

OpenAIRE

FUJII, Terushige; OHTA, Jun-ichi; AKAGAWA, Koji; NAKAMURA, Toshi; ASANO, Hitoshi

1992-01-01

From the viewpoint of energy conservation and the development of new energy resources,it is important to utilize geothermal resources and waste heat from factories. Among energy conversion device,there is a radial outflow reaction turbine,i.e.,Hero's turbine. Performance characteristics of Hero's turbine are analytically and experimentally clarified for flashing expansion of initially subcooled hot water. It is found that: (a)there is an optimum number of revolutions at which maximum tubine e...

Planning for the Management and Disposition of Newly Generated TRU Waste from REDC

International Nuclear Information System (INIS)

Coffey, D. E.; Forrester, T. W.; Krause, T.

2002-01-01

This paper describes the waste characteristics of newly generated transuranic waste from the Radiochemical Engineering and Development Center at the Oak Ridge National Laboratory and the basic certification structure that will be proposed by the University of Tennessee-Battelle and Bechtel Jacobs Company LLC to the Waste Isolation Pilot Plant for this waste stream. The characterization approach uses information derived from the active production operations as acceptable knowledge for the Radiochemical Engineering and Development Center transuranic waste. The characterization approach includes smear data taken from processing and waste staging hot cells, as well as analytical data on product and liquid waste streams going to liquid waste disposal. Bechtel Jacobs Company and University of Tennessee-Battelle are currently developing the elements of a Waste Isolation Pilot Plant-compliant program with a plan to be certified by the Waste Isolation Pilot Plant for shipment of newly generated transuranic waste in the next few years. The current activities include developing interface plans, program documents, and waste stream specific procedures

Ferrocyanide safety program: Credibility of drying out ferrocyanide tank waste by hot spots

International Nuclear Information System (INIS)

Dickinson, D.R.; McLaren, J.M.; Borsheim, G.L.; Crippen, M.D.

1993-04-01

The single-shell waste tanks at the Hanford Site that contain significant quantities of ferrocyanide have been considered a possible hazard, since under certain conditions the ferrocyanide in the waste tanks could undergo an exothermic chemical reaction with the nitrates and nitrites that are also present in the tanks. The purpose of this report is to assess the credibility of local dryout of ferrocyanide due to a hotspot. This report considers the following: What amount of decay heat generation within what volume would be necessary to raise the temperature of the liquid in the sludge to its boiling point? What mechanisms could produce a significant local concentration of heat sources? Is it credible that a waste tank heat concentration could be as large as that required to reach the dryout temperatures? This report also provides a recommendation as to whether infrared scanning of the ferrocyanide tanks is needed. From the analyses presented in this report it is evident that formation of dry, and thus chemically reactive, regions in the ferrocyanide sludge by local hotspots is not credible. This conclusion is subject to reevaluation if future analyses of tank core samples show much higher 137 Cs or 90 Sr concentrations than expected. Since hotspots of concern are not credible, infrared scanning to detect such hotspots is not required for safe storage of tank waste

Waste heat and water recovery opportunities in California tomato paste processing

International Nuclear Information System (INIS)

Amón, Ricardo; Maulhardt, Mike; Wong, Tony; Kazama, Don; Simmons, Christopher W.

2015-01-01

Water and energy efficiency are important for the vitality of the food processing industry as demand for these limited resources continues to increase. Tomato processing, which is dominated by paste production, is a major industry in California – where the majority of tomatoes are processed in the United States. Paste processing generates large amounts of condensate as moisture is removed from the fruit. Recovery of the waste heat in this condensate and reuse of the water may provide avenues to decrease net energy and water use at processing facilities. However, new processing methods are needed to create demand for the condensate waste heat. In this study, the potential to recover condensate waste heat and apply it to the tomato enzyme thermal inactivation processing step (the hot break) is assessed as a novel application. A modeling framework is established to predict heat transfer to tomatoes during the hot break. Heat recovery and reuse of the condensate water are related to energy and monetary savings gained through decreased use of steam, groundwater pumping, cooling towers, and wastewater processing. This analysis is informed by water and energy usage data from relevant unit operations at a commercial paste production facility. The case study indicates potential facility seasonal energy and monetary savings of 7.3 GWh and $166,000, respectively, with most savings gained through reduced natural gas use. The sensitivity of heat recovery to various process variables associated with heat exchanger design and processing conditions is presented to identify factors that affect waste heat recovery. - Highlights: • The potential to recovery waste heat in tomato paste processing is examined. • Heat transfer from evaporator condensate to tomatoes in the hot break is modeled. • Processing facility data is used in model to predict heat recovery energy savings. • The primary benefit of heat recovery is reduced use of natural gas in boilers. • Reusing

Handling of Highly Radioactive Radiation Sources in a Hot Cell Using a Mechanically Driven Cell Crane - 13452

Energy Technology Data Exchange (ETDEWEB)

Klute, Stefan; Huber, Wolfgang-Bruno [Siempelkamp Nukleartechnik GmbH, Am Taubenfeld 25/1, 69123 Heidelberg (Germany); Meyer, Franz [Nuclear Engineering Seibersdorf GmbH, 2444 Seibersdorf (Austria)

2013-07-01

In 2010, Siempelkamp Nukleartechnik GmbH was awarded the contract for design and erection of a Hot Cell for handling and storage of highly radioactive radiation sources. This Hot Cell is part of a new hot cell laboratory, constructed for the NHZ (Neues Handhabungszentrum = New Handling Center) of the Nuclear Engineering Seibersdorf GmbH (NES). All incurring radioactive materials from Austria are collected in the NHZ, where they are safely conditioned and stored temporarily until their final storage. The main tasks of the NES include, apart from the collection, conditioning and storage of radioactive waste, also the reprocessing and the decontamination of facilities and laboratories originating from 45 years of research and development at the Seibersdorf site as well as the operation of the Hot Cell Laboratory [1]. The new Hot Cell Laboratory inside the NHZ consists of the following room areas: - One hot cell, placed in the center, for remote controlled, radiation protected handling of radioactive materials, including an integrated floor storage for the long-term temporary storage of highly radioactive radiation sources; - An anteroom for the loading and unloading of the hot cell; - One control room for the remote controlling of the hot cell equipment; - One floor storage, placed laterally to the hot cell, for burial, interim storage and removal of fissionable radioactive material in leak-proof packed units in 100 l drums. The specific design activity of the hot cell of 1.85 Pbq relating to 1-Me-Radiator including the integrated floor storage influences realization and design of the components used in the cell significantly. (authors)

Development of bitumization plant of radioactive waste

International Nuclear Information System (INIS)

Lee, S.H.; Chun, K.S.; Song, H.Y.; Park, S.H.; Park, W.J.; Ryoo, R.; Cho, W.J.; Wee, K.S.; Lee, J.D.

1981-01-01

This study is to develop the bitumization techniques of radioactive waste concentrates from nuclear power plants (PWR), Cold-run and hot-run bitumization pilot plant were designed and constructed, and testing was carried out. The results show that the range of pH 7 to 8, in which a soluble product could form, is suitable for feeding in the waste and for its incorporation into asphalt, domestic blown asphalt is suitable up to an exposure of 10 8 rad. The estimated viscosity of the asphalt/waste product is about 14 poises at 175 0 C and the asphalt/wastes products in simulated sea-water for three weeks showed no evidence of swelling. The leaching rate from the spent resin/asphalt products in distilled water less than 1/1000 of that from the cement-products. It is considered that the resin content which is suitable for the incorporation of spent resin into asphalt is less than 50 wt%

High-level radioactive waste fixation in sintered vitreous matrix

International Nuclear Information System (INIS)

Russo, D.O.; Messi de Bernasconi, N.; Audero, M.A.

1987-01-01

The safe storage of high-level wastes from fuel elements reprocessing includes, as a first step, the fixation of the same in materials having a good resistance to the leaching in aqueous medium, such as borosilicate glass. As an alternative to the usual method of the molten glasses, a procedure for the sintering of a powdered glass and waste mixture at lower temperatures (600-700 deg C) has been developed, which minimizes the volatilization of active compounds during the process. Two glasses matrices of different composition and characteristics were used, to which the simulated wastes were added in the ratio of a 10% in weight of oxides. Two sintering techniques were employed 1: cold pressing and further sintering; 2: hot pressing and sintering under pressure. The densities were measured, the microstructure of the samples was analyzed and leaching essays were made in distilled water. The pellet's microstructure was observed by means of optical microscopy, by reflection in polished samples and by transparency in thin slices. The presence of crystalline compounds was analyzed by means of x rays and electron microprobe. The results have shown the convenience to continue with hot pressing essays, because a denser product with a higher resistance to the leaching is thus obtained. (M.E.L.) [es

Treatment of solid radioactive waste: Volume reduction of non-combustible waste

International Nuclear Information System (INIS)

Boehme, G.

1982-01-01

Press compaction is very common as for volume reduction of low level radioactive solid waste. In most cases a sorting step and if necessary a fragmenting step are desirable prior to the compaction process. Besides contamination-free loading and unloading techniques are important. Typical technical solutions for mixed solid waste handling and compacting equipment are shown and discussed by means of the lay-out drawings for a medium size radwaste compaction facility. A special technique can be applied if one has to compact active exhaust air filters in a hot cell. KfK has developed a remotely operated mobile equipment for this purpose. As for the nuclear fuel cycle considerable interest is existing in compacting spent fuel halls after fuel dissolution. In various European countries mechanical compaction and high temperature processes are therefore under development. These processes are described and the related equipment is discussed. (orig./RW)

Utilization of Waste Clay from Boron Production in Bituminous Geosynthetic Barrier (GBR-B Production as Landfill Liner

Directory of Open Access Journals (Sweden)

Müfide Banar

2016-01-01

Full Text Available Bituminous geomembranes, one type of geosynthetics, include a hot bituminous mixture with mineral filler and reinforcement. In this study, boron production waste clay (CW was used as filler to produce a geosynthetic barrier with bentonite, waste tire, and bitumen. Bentonite and waste tires were used as auxiliary fillers and bitumen as the binder. CW/bitumen, CW/bentonite/bitumen, and CW/waste tire/bitumen mixtures were prepared by using a laboratory mixer at 100°C. Hot mixtures were extruded into strips by using a lab-scale corotating twin screw extruder (L/D: 40 followed by die casting (2 mm × 100 mm. Glass fleece or nonwoven polyester was used as reinforcement material and while die casting, both sides of the reinforcement materials were covered with bituminous mixture. Thickness, mass per unit area, tensile strength, elongation at yield, and hydraulic conductivity were used to characterize the geomembranes. Among all geomembranes, nonwoven polyester covered with 30% bitumen-70% boron waste clay mixture (PK-BTM30CW70 was found to be the most promising in terms of structure and mechanical behaviour. After that, consequences of its exposure to distilled water (DW, municipal solid waste landfill leachate (L-MSW, and hazardous waste landfill leachate (L-HW were examined to use for an innovative impermeable liner on solid waste landfills.

Status of the high-level nuclear waste disposal program in Japan

International Nuclear Information System (INIS)

Uematsu, K.

1985-01-01

The Japan Atomic Energy Commission (JAEC) initiated a high-level radioactive waste disposal program in 1976. Since then, the Advisory Committee on Radioactive Waste Management of JAEC has revised the program twice. The latest revision was issued in 1984. The committee recommended a four-phase program and the last phase calls for the beginning of emplacement of the high-level nuclear waste into a selected repository in the Year 2000. The first phase is already completed, and the second phase of this decade calls for the selection of a candidate disposal site and the conducting of the RandD of waste disposal in an underground research laboratory and in a hot test facility. This paper covers the current status of the high-level nuclear waste disposal program in Japan

Employing Hot Wire Anemometry to Directly Measure the Water Balance of a Proton Exchange Membrane Fuel Cell

DEFF Research Database (Denmark)

Shakhshir, Saher Al; Berning, Torsten

Proton exchange membrane fuel cells (PEMFC’s) are currently being commercialized for various applications ranging from automotive to stationary such as powering telecom back-up units. In PEMFC’s, oxygen from air is internally combined with hydrogen to form water and produce electricity and waste......-hoc and real time electrical signal of the fuel cell water balance by employing hot wire anemometry. The hot wire sensor is placed into a binary mixture of hydrogen and water vapour, and the voltage signal received gives valuable insight into heat and mass transfer phenomena in a PEMFC. A central question...

Civil engineering challenge with nuclear waste

International Nuclear Information System (INIS)

Day, D.

1985-01-01

The civil engineer can help to solve the problems in disposing of nuclear waste in a deep geologic formation. The site for a nuclear waste repository must be carefully selected so that the geology provides the natural barrier between the waste and the accessible environment specified by the NRC and the EPA. This engineer is familiar with the needed structure and conditions of the host and surrounding rocks, and also the hydraulic mechanisms for limiting the migration of water in the rocks. To dispose of the nuclear waste underground requires stable and long-lasting shafts and tunnels such as civil engineers have designed and constructed for many other uses. The planning, design and construction of the ground surface facilities for a nuclear waste repository involves civil engineering in many ways. The transporation of heavy, metal shielded casks requires special attention to the system of highways and railroads accessing the repository. Structures for handling the shipping casks and transferring the waste onsite and into the deep geologic formation need special considerations. The structures must provide the NRC required containment, including hot cells for remote handling. Therefore, structural design strives for buildings, ventilation structures, shaft headframes, etc., to be earthquake and tornado-proof. These important design bases and considerations for the civil engineer working on a nuclear waste repository are discussed in this paper

Management of Tritium in ITER Waste

International Nuclear Information System (INIS)

Rosanvallon, S.; Benchikhoune, M.; Ciattaglia, S.; Uzan, J. Elbez; Na, B. C.; Taylor, N.; Gastaldi, O.

2011-01-01

ITER will use tritium as fuel. Procedures and processes are thus put in place in order to recover the tritium that is not used in the fusion reaction, including from waste and effluents. The tritium thus recovered can be re-injected into the fuel cycle. Moreover, tritium content and thus outgassing may be a safety concern, because of the potential for releases to the environment, both from the facility and from the final disposal (subjected to stringent acceptance criteria in the current waste final disposal). The aim of this paper is to present the measures considered to deal with the specific case of tritium in the liquid and solid waste that will arise from ITER operation and decommissioning. It concerns the processes that are considered from the waste production to its final disposal and in particular: the tritium removal stages (in-situ divertor baking at 350 C and tritium removal from solid waste and liquid and gaseous effluents), the removal of dust contamination (dust containing tritium produced by plasma-wall interaction and by the maintenance/ refurbishment processes) and the measures to enable safe processing and storage of the waste (wall-liner in the hot cell facility to limit concrete contamination and interim storage enabling tritium decay for waste that could not be directly accepted in the host-country final disposal facilities). (authors)

HOT STARS WITH HOT JUPITERS HAVE HIGH OBLIQUITIES

International Nuclear Information System (INIS)

Winn, Joshua N.; Albrecht, Simon; Fabrycky, Daniel; Johnson, John Asher

2010-01-01

We show that stars with transiting planets for which the stellar obliquity is large are preferentially hot (T eff > 6250 K). This could explain why small obliquities were observed in the earliest measurements, which focused on relatively cool stars drawn from Doppler surveys, as opposed to hotter stars that emerged more recently from transit surveys. The observed trend could be due to differences in planet formation and migration around stars of varying mass. Alternatively, we speculate that hot-Jupiter systems begin with a wide range of obliquities, but the photospheres of cool stars realign with the orbits due to tidal dissipation in their convective zones, while hot stars cannot realign because of their thinner convective zones. This in turn would suggest that hot Jupiters originate from few-body gravitational dynamics and that disk migration plays at most a supporting role.

Development and characterization of new high-level waste form containing LiCl KCl eutectic salts for achieving waste minimization from pyroprocessing

International Nuclear Information System (INIS)

Cho, Yong Zun; Kim, In Tae; Park, Hwan Seo; Ahn, Byeung Gil; Eun, Hee Chul; Son, Seock Mo; Ah, Su Na

2011-12-01

The purpose of this project is to develop new high level waste (HLW) forms and fabrication processes to dispose of active metal fission products that are removed from electrorefiner salts in the pyroprocessing based fuel cycle. The current technology for disposing of active metal fission products in pyroprocessing involves non selectively discarding of fission product loaded salt in a glass-bonded sodalite ceramic waste form. Selective removal of fission products from the molten salt would greatly minimize the amount of HLW generated and methods were developed to achieve selective separation of fission products during a previous I NERI research project (I NERI 2006 002 K). This I NERI project proceeds from the previous project with the development of suitable waste forms to immobilize the separated fission products. The Korea Atomic Energy Research Institute (KAERI) has focused primarily on developing these waste forms using surrogate waste materials, while the Idaho National Laboratory (INL) has demonstrated fabrication of these waste forms using radioactive electrorefiner salts in hot cell facilities available at INL. Testing and characterization of these radioactive materials was also performed to determine the physical, chemical, and durability properties of the waste forms

Tank selection for Light Duty Utility Arm (LDUA) system hot testing in a single shell tank

Energy Technology Data Exchange (ETDEWEB)

Bhatia, P.K.

1995-01-31

The purpose of this report is to recommend a single shell tank in which to hot test the Light Duty Utility Arm (LDUA) for the Tank Waste Remediation System (TWRS) in Fiscal Year 1996. The LDUA is designed to utilize a 12 inch riser. During hot testing, the LDUA will deploy two end effectors (a High Resolution Stereoscopic Video Camera System and a Still/Stereo Photography System mounted on the end of the arm`s tool interface plate). In addition, three other systems (an Overview Video System, an Overview Stereo Video System, and a Topographic Mapping System) will be independently deployed and tested through 4 inch risers.

Tank selection for Light Duty Utility Arm (LDUA) system hot testing in a single shell tank

International Nuclear Information System (INIS)

Bhatia, P.K.

1995-01-01

The purpose of this report is to recommend a single shell tank in which to hot test the Light Duty Utility Arm (LDUA) for the Tank Waste Remediation System (TWRS) in Fiscal Year 1996. The LDUA is designed to utilize a 12 inch riser. During hot testing, the LDUA will deploy two end effectors (a High Resolution Stereoscopic Video Camera System and a Still/Stereo Photography System mounted on the end of the arm's tool interface plate). In addition, three other systems (an Overview Video System, an Overview Stereo Video System, and a Topographic Mapping System) will be independently deployed and tested through 4 inch risers

Selection of efficient options for processing and storage of radioactive waste in countries with small amounts of waste generation

International Nuclear Information System (INIS)

2003-09-01

The report is intended to assist decision makers in countries using nuclear energy for non-power applications to organize their waste management practices. It describes methodologies, criteria and options for the selection of appropriate technologies for processing and storage of low and intermediate level radioactive waste from different nuclear applications. The report reviews both technical and non-technical factors important for decision making and planning, and for implementation of waste management activities at the country and facility levels. It makes practical recommendations for the selection of particular technologies for different scales of waste generation. These wastes may arise from production of radionuclides and their application in industry, agriculture, medicine, education and research. The report also considers waste generated at research reactors, research centers and research laboratories using radioisotopes, as well as waste from decommissioning of research reactors and small nuclear facilities such as hot cells, laboratories and irradiation facilities. Management of uranium mining and milling waste and management of spent fuel from research reactors are not considered in this report. Discussed in detail are: the basic legal, regulatory, administrative and technical requirements set up in a national waste management system and review of the factors and components affecting the selection of an appropriate national waste management system. the origins and characteristics of radioactive waste from different nuclear applications. the technical factors that might affect the selection of waste processing and storage technologies, the main waste management steps, information on available technologies, the basis for planning of waste processing and storage and the selection of a particular option for radioactive waste processing and storage in countries with a different scale of nuclear applications

Selection of efficient options for processing and storage of radioactive waste in countries with small amounts of waste generation

Energy Technology Data Exchange (ETDEWEB)

NONE

2003-09-01

The report is intended to assist decision makers in countries using nuclear energy for non-power applications to organize their waste management practices. It describes methodologies, criteria and options for the selection of appropriate technologies for processing and storage of low and intermediate level radioactive waste from different nuclear applications. The report reviews both technical and non-technical factors important for decision making and planning, and for implementation of waste management activities at the country and facility levels. It makes practical recommendations for the selection of particular technologies for different scales of waste generation. These wastes may arise from production of radionuclides and their application in industry, agriculture, medicine, education and research. The report also considers waste generated at research reactors, research centers and research laboratories using radioisotopes, as well as waste from decommissioning of research reactors and small nuclear facilities such as hot cells, laboratories and irradiation facilities. Management of uranium mining and milling waste and management of spent fuel from research reactors are not considered in this report. Discussed in detail are: the basic legal, regulatory, administrative and technical requirements set up in a national waste management system and review of the factors and components affecting the selection of an appropriate national waste management system. the origins and characteristics of radioactive waste from different nuclear applications. the technical factors that might affect the selection of waste processing and storage technologies, the main waste management steps, information on available technologies, the basis for planning of waste processing and storage and the selection of a particular option for radioactive waste processing and storage in countries with a different scale of nuclear applications.

Hot Laboratories and Remote Handling

International Nuclear Information System (INIS)

2007-01-01

The Opening talk of the workshop 'Hot Laboratories and Remote Handling' was given by Marin Ciocanescu with the communication 'Overview of R and D Program in Romanian Institute for Nuclear Research'. The works of the meeting were structured into three sections addressing the following items: Session 1. Hot cell facilities: Infrastructure, Refurbishment, Decommissioning; Session 2. Waste, transport, safety and remote handling issues; Session 3. Post-Irradiation examination techniques. In the frame of Section 1 the communication 'Overview of hot cell facilities in South Africa' by Wouter Klopper, Willie van Greunen et al, was presented. In the framework of the second session there were given the following four communications: 'The irradiated elements cell at PHENIX' by Laurent Breton et al., 'Development of remote equipment for DUPIC fuel fabrication at KAERI', by Jung Won Lee et al., 'Aspects of working with manipulators and small samples in an αβγ-box, by Robert Zubler et al., and 'The GIOCONDA experience of the Joint Research Centre Ispra: analysis of the experimental assemblies finalized to their safe recovery and dismantling', by Roberto Covini. Finally, in the framework of the third section the following five communications were presented: 'PIE of a CANDU fuel element irradiated for a load following test in the INR TRIGA reactor' by Marcel Parvan et al., 'Adaptation of the pole figure measurement to the irradiated items from zirconium alloys' by Yury Goncharenko et al., 'Fuel rod profilometry with a laser scan micrometer' by Daniel Kuster et al., 'Raman spectroscopy, a new facility at LECI laboratory to investigate neutron damage in irradiated materials' by Lionel Gosmain et al., and 'Analysis of complex nuclear materials with the PSI shielded analytical instruments' by Didier Gavillet. In addition, eleven more presentations were given as posters. Their titles were: 'Presentation of CETAMA activities (CEA analytic group)' by Alain Hanssens et al. 'Analysis of

Performance Characteristics of Hero's Turbine Using Hot Water as a Working Fluid

OpenAIRE

藤井, 照重; 太田, 淳一; 赤川, 浩爾; 中村, 登志; 浅野, 等

1990-01-01

From the view point of energy saving and the development of new energy resources,it is important to utilize geothermal resources and waste heat from factories. As one of the energy conversion expanders,there is a radial outflow reaction turbine(that is,Hero's turbine). Performance characteristics of Hero's turbine using subcooled hot water as a working fluid are clarified analytically and experimentally. It is found that:(a)there is an optimum rotational speed at which maximum turbine efficie...

Tea waste: an effective and economic substrate for oyster mushroom cultivation.

Science.gov (United States)

Yang, Doudou; Liang, Jin; Wang, Yunsheng; Sun, Feng; Tao, Hong; Xu, Qiang; Zhang, Liang; Zhang, Zhengzhu; Ho, Chi-Tang; Wan, Xiaochun

2016-01-30

Tea waste is the residue that remains after tea leaves have been extracted by hot water to obtain water-soluble components. The waste contains a re-usable energy substrate and nutrients which may pollute the environment if they are not dealt with appropriately. Other agricultural wastes have been widely studied as substrates for cultivating mushrooms. In the present study, we cultivated oyster mushroom using tea waste as substrate. To study the feasibility of re-using it, tea waste was added to the substrate at different ratios in different experimental groups. Three mushroom strains (39, 71 and YOU) were compared and evaluated. Mycelia growth rate, yield, biological efficiency and growth duration were measured. Substrates with different tea waste ratios showed different growth and yield performance. The substrate containing 40-60% of tea waste resulted in the highest yield. Tea waste could be used as an effective and economic substrate for oyster mushroom cultivation. This study also provided a useful way of dealing with massive amounts of tea waste. © 2015 Society of Chemical Industry.

Heat Pipe-Assisted Thermoelectric Power Generation Technology for Waste Heat Recovery

Science.gov (United States)

Jang, Ju-Chan; Chi, Ri-Guang; Rhi, Seok-Ho; Lee, Kye-Bock; Hwang, Hyun-Chang; Lee, Ji-Su; Lee, Wook-Hyun

2015-06-01

Currently, large amounts of thermal energy dissipated from automobiles are emitted through hot exhaust pipes. This has resulted in the need for a new efficient recycling method to recover energy from waste hot exhaust gas. The present experimental study investigated how to improve the power output of a thermoelectric generator (TEG) system assisted by a wickless loop heat pipe (loop thermosyphon) under the limited space of the exhaust gas pipeline. The present study shows a novel loop-type heat pipe-assisted TEG concept to be applied to hybrid vehicles. The operating temperature of a TEG's hot side surface should be as high as possible to maximize the Seebeck effect. The present study shows a novel TEG concept of transferring heat from the source to the sink. This technology can transfer waste heat to any local place with a loop-type heat pipe. The present TEG system with a heat pipe can transfer heat and generate an electromotive force power of around 1.3 V in the case of 170°C hot exhaust gas. Two thermoelectric modules (TEMs) for a conductive block model and four Bi2Te3 TEMs with a heat pipe-assisted model were installed in the condenser section. Heat flows to the condenser section from the evaporator section connected to the exhaust pipe. This novel TEG system with a heat pipe can be placed in any location on an automobile.

Melting behaviour of raw materials and recycled stone wool waste

DEFF Research Database (Denmark)

Schultz-Falk, Vickie; Agersted, Karsten; Jensen, Peter Arendt

2018-01-01

Stone wool is a widely used material for building insulation, to provide thermal comfort along with fire stability and acoustic comfort for all types of buildings. Stone wool waste generated either during production or during renovation or demolition of buildings can be recycled back into the sto...... wool melt production. This study investigates and compares the thermal response and melting behaviour of a conventional stone wool charge and stone wool waste. The study combines differential scanning calorimetry (DSC), hot stage microscopy (HSM) and X-ray diffraction (XRD). DSC reveals...... that the conventional charge and stone wool waste have fundamentally different thermal responses, where the charge experiences gas release, phase transition and melting of the individual raw materials. The stone wool waste experiences glass transition, crystallization and finally melting. Both DSC and HSM measurements...

Designing Advanced Ceramic Waste Forms for Electrochemical Processing Salt Waste

International Nuclear Information System (INIS)

Ebert, W. L.; Snyder, C. T.; Frank, Steven; Riley, Brian

2016-01-01

This report describes the scientific basis underlying the approach being followed to design and develop ''advanced'' glass-bonded sodalite ceramic waste form (ACWF) materials that can (1) accommodate higher salt waste loadings than the waste form developed in the 1990s for EBR-II waste salt and (2) provide greater flexibility for immobilizing extreme waste salt compositions. This is accomplished by using a binder glass having a much higher Na_2O content than glass compositions used previously to provide enough Na+ to react with all of the Cl- in the waste salt and generate the maximum amount of sodalite. The phase compositions and degradation behaviors of prototype ACWF products that were made using five new binder glass formulations and with 11-14 mass% representative LiCl/KCl-based salt waste were evaluated and compared with results of similar tests run with CWF products made using the original binder glass with 8 mass% of the same salt to demonstrate the approach and select a composition for further studies. About twice the amount of sodalite was generated in all ACWF materials and the microstructures and degradation behaviors confirmed our understanding of the reactions occurring during waste form production and the efficacy of the approach. However, the porosities of the resulting ACWF materials were higher than is desired. These results indicate the capacity of these ACWF waste forms to accommodate LiCl/KCl-based salt wastes becomes limited by porosity due to the low glass-to-sodalite volume ratio. Three of the new binder glass compositions were acceptable and there is no benefit to further increasing the Na content as initially planned. Instead, further studies are needed to develop and evaluate alternative production methods to decrease the porosity, such as by increasing the amount of binder glass in the formulation or by processing waste forms in a hot isostatic press. Increasing the amount of binder glass to eliminate porosity will decrease the waste

Designing Advanced Ceramic Waste Forms for Electrochemical Processing Salt Waste

Energy Technology Data Exchange (ETDEWEB)

Ebert, W. L. [Argonne National Lab. (ANL), Argonne, IL (United States); Snyder, C. T. [Argonne National Lab. (ANL), Argonne, IL (United States); Frank, Steven [Argonne National Lab. (ANL), Argonne, IL (United States); Riley, Brian [Argonne National Lab. (ANL), Argonne, IL (United States)

2016-03-01

This report describes the scientific basis underlying the approach being followed to design and develop “advanced” glass-bonded sodalite ceramic waste form (ACWF) materials that can (1) accommodate higher salt waste loadings than the waste form developed in the 1990s for EBR-II waste salt and (2) provide greater flexibility for immobilizing extreme waste salt compositions. This is accomplished by using a binder glass having a much higher Na₂O content than glass compositions used previously to provide enough Na+ to react with all of the Cl– in the waste salt and generate the maximum amount of sodalite. The phase compositions and degradation behaviors of prototype ACWF products that were made using five new binder glass formulations and with 11-14 mass% representative LiCl/KCl-based salt waste were evaluated and compared with results of similar tests run with CWF products made using the original binder glass with 8 mass% of the same salt to demonstrate the approach and select a composition for further studies. About twice the amount of sodalite was generated in all ACWF materials and the microstructures and degradation behaviors confirmed our understanding of the reactions occurring during waste form production and the efficacy of the approach. However, the porosities of the resulting ACWF materials were higher than is desired. These results indicate the capacity of these ACWF waste forms to accommodate LiCl/KCl-based salt wastes becomes limited by porosity due to the low glass-to-sodalite volume ratio. Three of the new binder glass compositions were acceptable and there is no benefit to further increasing the Na content as initially planned. Instead, further studies are needed to develop and evaluate alternative production methods to decrease the porosity, such as by increasing the amount of binder glass in the formulation or by processing waste forms in a hot isostatic press. Increasing the amount of binder glass to eliminate porosity will decrease

Treatment of concrete bars from the dismantling of hot cells

International Nuclear Information System (INIS)

Graf, A.; Stutz, U.; Valencia, L.

2002-01-01

The Central Decontamination Operations Department (HDB) of the Karlsruhe Research Center operates facilities for the disposal of radioactive waste. In general, their objective is to decontaminate radioactive residues for unrestricted release in order to minimize the volume of waste products suitable for repository storage. In the case of about 120 concrete bars from the dismantling of hot cells, we reduce the volume of radioactive waste by sawing off the most contaminated parts of the bar. If there are no insertions such as cables or ventilation systems, the rest of the bar is sandblasted and its activity manually measured to ensure compliance with the release criteria. Otherwise, the bar is minced into small pieces by a power shovel. Afterwards, the rubble is filled into drums and its activity is measured by the clearance measurement facility. If the rubble and the sandblasted bars do not exceed the activity limit specified by the release criteria, the material is disposed of without further regulations for unrestricted use. Those parts of the bars which can not be released must be stored in special containers suitable for the KONRAD final disposal. Using this method, about 70 % of the total mass can be released. (author)

Stabilization/solidification of hot dip galvanizing ash using different binders.

Science.gov (United States)

Vinter, S; Montanes, M T; Bednarik, V; Hrivnova, P

2016-12-15

This study focuses on solidification of hot dip-galvanizing ash with a high content of zinc and soluble substances. The main purpose of this paper is to immobilize these pollutants into a matrix and allow a safer way for landfill disposal of that waste. Three different binders (Portland cement, fly ash and coal fluidized-bed combustion ash) were used for the waste solidification. Effectiveness of the process was evaluated using leaching test according to EN 12457-4 and by using the variance analysis and the categorical multifactorial test. In the leaching test, four parameters were observed: pH, zinc concentration in leachate, and concentration of chlorides and dissolved substances in leachate. The acquired data was then processed using statistical software to find an optimal solidifying ratio of the addition of binder, water, and waste to the mixture, with the aim to fulfil the requirement for landfill disposal set by the Council Decision 2003/33/EC. The influence on the main observed parameters (relative amount of water and a binder) on the effectiveness of the used method and their influence of measured parameters was also studied. Copyright © 2016 Elsevier B.V. All rights reserved.

Standard guide for characterization of radioactive and/or hazardous wastes for thermal treatment

CERN Document Server

American Society for Testing and Materials. Philadelphia

2003-01-01

1.1 This guide identifies methods to determine the physical and chemical characteristics of radioactive and/or hazardous wastes before a waste is processed at high temperatures, for example, vitrification into a homogeneous glass ,glass-ceramic, or ceramic waste form. This includes waste forms produced by ex-situ vitrification (ESV), in-situ vitrification (ISV), slagging, plasma-arc, hot-isostatic pressing (HIP) and/or cold-pressing and sintering technologies. Note that this guide does not specifically address high temperature waste treatment by incineration but several of the analyses described in this guide may be useful diagnostic methods to determine incinerator off-gas composition and concentrations. The characterization of the waste(s) recommended in this guide can be used to (1) choose and develop the appropriate thermal treatment methodology, (2) determine if waste pretreatment is needed prior to thermal treatment, (3) aid in development of thermal treatment process control, (4) develop surrogate wa...

Tank Waste Remediation System tank waste pretreatment and vitrification process development testing requirements assessment

International Nuclear Information System (INIS)

Howden, G.F.

1994-01-01

A multi-faceted study was initiated in November 1993 to provide assurance that needed testing capabilities, facilities, and support infrastructure (sampling systems, casks, transportation systems, permits, etc.) would be available when needed for process and equipment development to support pretreatment and vitrification facility design and construction schedules. This first major report provides a snapshot of the known testing needs for pretreatment, low-level waste (LLW) and high-level waste (HLW) vitrification, and documents the results of a series of preliminary studies and workshops to define the issues needing resolution by cold or hot testing. Identified in this report are more than 140 Hanford Site tank waste pretreatment and LLW/HLW vitrification technology issues that can only be resolved by testing. The report also broadly characterizes the level of testing needed to resolve each issue. A second report will provide a strategy(ies) for ensuring timely test capability. Later reports will assess the capabilities of existing facilities to support needed testing and will recommend siting of the tests together with needed facility and infrastructure upgrades or additions

Tank Waste Remediation System tank waste pretreatment and vitrification process development testing requirements assessment

Energy Technology Data Exchange (ETDEWEB)

Howden, G.F.

1994-10-24

A multi-faceted study was initiated in November 1993 to provide assurance that needed testing capabilities, facilities, and support infrastructure (sampling systems, casks, transportation systems, permits, etc.) would be available when needed for process and equipment development to support pretreatment and vitrification facility design and construction schedules. This first major report provides a snapshot of the known testing needs for pretreatment, low-level waste (LLW) and high-level waste (HLW) vitrification, and documents the results of a series of preliminary studies and workshops to define the issues needing resolution by cold or hot testing. Identified in this report are more than 140 Hanford Site tank waste pretreatment and LLW/HLW vitrification technology issues that can only be resolved by testing. The report also broadly characterizes the level of testing needed to resolve each issue. A second report will provide a strategy(ies) for ensuring timely test capability. Later reports will assess the capabilities of existing facilities to support needed testing and will recommend siting of the tests together with needed facility and infrastructure upgrades or additions.

Minimizing temperature instability of heat recovery hot water system utilizing optimized thermal energy storage

Science.gov (United States)

Suamir, I. N.; Sukadana, I. B. P.; Arsana, M. E.

2018-01-01

One energy-saving technology that starts gaining attractive for hotel industry application in Indonesia is the utilization of waste heat of a central air conditioning system to heat water for domestic hot water supply system. Implementing the technology for such application at a hotel was found that hot water capacity generated from the heat recovery system could satisfy domestic hot water demand of the hotel. The gas boilers installed in order to back up the system have never been used. The hot water supply, however, was found to be instable with hot water supply temperature fluctuated ranging from 45 °C to 62 °C. The temperature fluctuations reaches 17 °C, which is considered instable and can reduce hot water usage comfort level. This research is aimed to optimize the thermal energy storage in order to minimize the temperature instability of heat recovery hot water supply system. The research is a case study approach based on cooling and hot water demands of a hotel in Jakarta-Indonesia that has applied water cooled chillers with heat recovery systems. The hotel operation with 329 guest rooms and 8 function rooms showed that hot water production in the heat recovery system completed with 5 m3 thermal energy storage (TES) could not hold the hot water supply temperature constantly. The variations of the cooling demand and hot water demands day by day were identified. It was found that there was significant mismatched of available time (hours) between cooling demand which is directly correlated to the hot water production from the heat recovery system and hot water usage. The available TES system could not store heat rejected from the condenser of the chiller during cooling demand peak time between 14.00 and 18.00 hours. The extra heat from the heat recovery system consequently increases the temperature of hot water up to 62 °C. It is about 12 K above 50 °C the requirement hot water temperature of the hotel. In contrast, the TES could not deliver proper

Design and management of hot-laboratories. [Manual]. Hottorabo no sekkei to kanri

Energy Technology Data Exchange (ETDEWEB)

1976-09-01

This document is a manual for the design and management of hot-laboratories. It is composed of three parts. The first part is devoted to the design of hot-laboratories. Items included here are; conceptual design; many regulations which must be considered at design stage; design of cave and its shielding; and the design of building, ventilation, and draining. Many examples of specific designs are presented by figures and photographs. The second part is concerned with the methods of operation management. Organizational structure, scheduling of operation, process management, and regulatory problems are discussed with some examples. Technological problems associated with the operation of a hot laboratory (e.g., manipulator, transfer machine, maintenance, and decontamination) are also discussed based on the authors' experiences. An example of the operation manual is presented for reference. The third part is devoted to the safety management and the training of personnel. The regulations by law are briefly explained. Most of this part is devoted to the problem of monitoring radio-activity. Monitoring of control areas, radio-active wastes, and personal dosage is discussed together with many other specific monitoring problems. As for training, the purpose and the present status are explained.

Chemistry of Hot Spring Pool Waters in Calamba and Los Banos and Potential Effect on the Water Quality of Laguna De Bay

Science.gov (United States)

Balangue, M. I. R. D.; Pena, M. A. Z.; Siringan, F. P.; Jago-on, K. A. B.; Lloren, R. B.; Taniguchi, M.

2014-12-01

Since the Spanish Period (1600s), natural hot spring waters have been harnessed for balneological purposes in the municipalities of Calamba and Los Banos, Laguna, south of Metro Manila. There are at more than a hundred hot spring resorts in Brgy. Pansol, Calamba and Tadlac, Los Banos. These two areas are found at the northern flanks of Mt. Makiling facing Laguna de Bay. This study aims to provide some insights on the physical and chemical characteristics of hot spring resorts and the possible impact on the lake water quality resulting from the disposal of used water. Initial ocular survey of the resorts showed that temperature of the pool water ranges from ambient (>300C) to as high as 500C with an average pool size of 80m3. Water samples were collected from a natural hot spring and pumped well in Los Banos and another pumped well in Pansol to determine the chemistry. The field pH ranges from 6.65 to 6.87 (Pansol springs). Cation analysis revealed that the thermal waters belonged to the Na-K-Cl-HCO3 type with some trace amount of heavy metals. Methods for waste water disposal are either by direct discharge down the drain of the pool or by discharge in the public road canal. Both methods will dump the waste water directly into Laguna de Bay. Taking in consideration the large volume of waste water used especially during the peak season, the effect on the lake water quality would be significant. It is therefore imperative for the environmental authorities in Laguna to regulate and monitor the chemistry of discharges from the pool to protect both the lake water as well as groundwater quality.

Immobilization of high-level wastes into sintered glass: 1

International Nuclear Information System (INIS)

Russo, D.O.; Messi de Bernasconi, N.; Audero, M.A.

1987-01-01

In order to immobilize the high-level radioactive wastes from fuel elements reprocessing, borosilicate glass was adopted. Sintering experiments are described with the variety VG 98/12 (SiO 2 , TiO 2 , Al 2 O 3 , B 2 O 3 , MgO, CaO and Na 2 O) (which does not present devitrification problems) mixed with simulated calcinated wastes. The hot pressing line (sintering under pressure) was explored in two variants 1: In can; 2: In graphite matrix with sintered pellet extraction. With scanning electron microscopy it is observed that the simulated wastes do not disolve in the vitreous matrix, but they remain dispersed in the same. The results obtained point out that the leaching velocities are independent from the density and from the matrix type employed, as well as from the fact that the wastes do no dissolve in the matrix. (M.E.L.) [es

Environment-friendly heat supply with natural refrigerants. Large heat pumps use industrial waste heat and waste water; Umweltschonende Waermeversorgung mit natuerlichen Kaeltemitteln. Grosswaermepumpen nutzen industrielle Abwaerme und Abwaesser

Energy Technology Data Exchange (ETDEWEB)

Anon.

2011-01-15

Everywhere, where industrial processes occur or coldness is produced, simultaneously heat is produced. While many private houses use geothermal energy or ambient air for the production of heat, waste water and waste heat prove to be optimal energy sources for the industrial need due to higher output temperatures. By means of large heat pumps the residual heat is used for heating or the supply of hot water for example in local heat supply grids and makes an important contribution to climate protection.

Project B-589, 300 Area transuranic waste interim storage project engineering study

International Nuclear Information System (INIS)

Greenhalgh, W.O.

1985-08-01

The purpose of the study was to look at various alternatives of taking newly generated, remote-handled transuranic waste (caisson waste) in the 300 Area, performing necessary transloading operations and preparing the waste for storage. The prepared waste would then be retrieved when the Waste Isolation Pilot Plant becomes operational and transshipped to the repository in New Mexico with a minimum of inspection and packaging. The scope of this study consisted of evaluating options for the transloading of the TRU wastes for shipment to a 200 Area storage site. Preconceptual design information furnished as part of the engineering study is listed below: produce a design for a clean, sealed waste canister; hot cell loadout system for the waste; in-cell loading or handling equipment; determine transshipment cask options; determine assay system requirements (optional); design or specify transport equipment required; provide a SARP cost estimate; determine operator training requirements; determine waste compaction equipment needs if desirable; develop a cost estimate and approximate schedule for a workable system option; and update the results presented in WHC Document TC-2025

Process and plant for manipulating radioactive waste

International Nuclear Information System (INIS)

Baatz, H.; Rittscher, D.

1979-01-01

To prepare waste for engineered storage, it is vitrified or calcined (pelleted) into solid pieces in a hot cell. Subsequently the pieces are filled into a radiation protection container made of spherulite cast iron, in which there is a liquid metal. The pieces are embedded in this metal matrix. During the embedding process, the radiation protection container may be heated externally or internally (lost heat device). (DG) [de

Preparation of plutonium waste forms with ICPP calcined high-level waste

Energy Technology Data Exchange (ETDEWEB)

Staples, B.A.; Knecht, D.A. [Lockheed Idaho Technologies Co., Idaho Falls, ID (United States); O`Holleran, T.P. [Argonne National Lab.-West, Idaho Falls, ID (United States)] [and others

1997-05-01

Glass and glass-ceramic forms developed for the immobilization of calcined high-level wastes generated by Idaho Chemical Processing Plant (ICPP) fuel reprocessing activities have been investigated for ability to immobilize plutonium and to simultaneously incorporate calcined waste as an anti-proliferation barrier. Within the forms investigated, crystallization of host phases result in an increased loading of plutonium as well as its incorporation into potentially more durable phases than the glass. The host phases were initially formed and characterized with cerium (Ce{sup +4}) as a surrogate for plutonium (Pu{sup +4}) and samarium as a neutron absorber for criticality control. Verification of the surrogate testing results were then performed replacing cerium with plutonium. All testing was performed with surrogate calcined high-level waste. The results of these tests indicated that a potentially useful host phase, based on zirconia, can be formed either by devitrification or solid state reaction in the glass studied. This phase incorporates plutonium as well as samarium and the calcined waste becomes part of the matrix. Its ease of formation makes it potentially useful in excess plutonium dispositioning. Other durable host phases for plutonium and samarium, including zirconolite and zircon have been formed from zirconia or alumina calcine through cold press-sintering techniques and hot isostatic pressing. Host phase formation experiments conducted through vitrification or by cold press-sintering techniques are described and the results discussed. Recommendations are given for future work that extends the results of this study.

Preparation of plutonium waste forms with ICPP calcined high-level waste

International Nuclear Information System (INIS)

Staples, B.A.; Knecht, D.A.; O'Holleran, T.P.

1997-05-01

Glass and glass-ceramic forms developed for the immobilization of calcined high-level wastes generated by Idaho Chemical Processing Plant (ICPP) fuel reprocessing activities have been investigated for ability to immobilize plutonium and to simultaneously incorporate calcined waste as an anti-proliferation barrier. Within the forms investigated, crystallization of host phases result in an increased loading of plutonium as well as its incorporation into potentially more durable phases than the glass. The host phases were initially formed and characterized with cerium (Ce +4 ) as a surrogate for plutonium (Pu +4 ) and samarium as a neutron absorber for criticality control. Verification of the surrogate testing results were then performed replacing cerium with plutonium. All testing was performed with surrogate calcined high-level waste. The results of these tests indicated that a potentially useful host phase, based on zirconia, can be formed either by devitrification or solid state reaction in the glass studied. This phase incorporates plutonium as well as samarium and the calcined waste becomes part of the matrix. Its ease of formation makes it potentially useful in excess plutonium dispositioning. Other durable host phases for plutonium and samarium, including zirconolite and zircon have been formed from zirconia or alumina calcine through cold press-sintering techniques and hot isostatic pressing. Host phase formation experiments conducted through vitrification or by cold press-sintering techniques are described and the results discussed. Recommendations are given for future work that extends the results of this study

Potential ability of zeolite to generate high-temperature vapor using waste heat

Science.gov (United States)

Fukai, Jun; Wijayanta, Agung Tri

2018-02-01

In various material product industries, a large amount of high temperature steam as heat sources are produced from fossil fuel, then thermal energy retained by condensed water at lower than 100°C are wasted. Thermal energies retained by exhaust gases at lower than 200°C are also wasted. Effective utilization of waste heat is believed to be one of important issues to solve global problems of energy and environment. Zeolite/water adsorption systems are introduced to recover such low-temperature waste heats in this study. Firstly, an adsorption steam recovery system was developed to generate high temperature steam from unused hot waste heat. The system used a new principle that adsorption heat of zeolite/water contact was efficiently extracted. A bench-scaled system was constructed, demonstrating contentious generation of saturated steam nearly 150°C from hot water at 80°C. Energy conservation is expected by returning the generated steam to steam lines in the product processes. Secondly, it was demonstrated that superheated steam/vapor at higher than 200°C could be generated from those at nearly 120°C using a laboratory-scaled setup. The maximum temperature and the time variation of output temperature were successfully estimated using macroscopic heat balances. Lastly, the maximum temperatures were estimated whose saturate air at the relative humidity 20-80% were heated by the present system. Theoretically, air at higher than 200°C was generated from saturate air at higher than 70°C. Consequently, zeolite/water adsorption systems have potential ability to regenerate thermal energy of waste water and exhaust gases.

An assessment of methods for immobilizing reprocessed radioactive waste

International Nuclear Information System (INIS)

Murthy, M.K.; Baranyi, A.D.

1980-05-01

Nuclear waste forms presently used for the disposal of high-level wastes and other potential waste forms under development were studied using information available in the literature and by visits to the laboratories. The following waste forms were considered: Borosilicate glass, high-silica glass, glassceramics, supercalcine ceramics, synroc ceramics, borosilicate glass beads in a metal matrix, supercalcine and synroc ceramics in a metal matrix and coated ceramics. The following conclusions have been reached: To date the best developed wasteform, both in terms of overall product quality and process development, is monolithic borosilicate glass. However, hydrothermal instability is a major concern. Borosilicate glass in metal matrix waste form has better properties than monolithic borosilicate glass waste form. The process has been proven on a pilot scale. Hence, it is considered very close to monolithic glass in terms of overall development. The product qualities of the other waste forms are better than borosilicate glass. However, process development for these alternative waste forms is still in a conceptual stage. The technological basis for processing ceramic waste forms exists in a well developed state. Nevertheless, adaptation of the technology to continuous hot-cell operation, although feasible, has not been demonstrated. In view of the product potential of ceramic waste forms it is felt that their development should be given emphasis at this time. (auth)

Waste Disposition Issues and Resolutions at the TRU Waste Processing Center at Oak Ridge TN

International Nuclear Information System (INIS)

Gentry, R.

2009-01-01

This paper prepared for the Waste Management Conference 2009 provides lessons learned from the Transuranic (TRU) Waste Processing Center (TWPC) associated with development of approaches used to certify and ensure disposition of problematic TRU wastes at the Waste Isolation Pilot Plant (WIPP) site. The TWPC is currently processing the inventory of available waste TRU waste at the Oak Ridge National Lab (ORNL). During the processing effort several waste characteristics were identified/discovered that did not conform to the normal standards and processes for disposal at WIPP. Therefore, the TWPC and ORNL were challenged with determining a path forward for this problematic, special case TRU wastes to ensure that they can be processed, packaged, and shipped to WIPP. Additionally, unexpected specific waste characteristics have challenged the project to identify and develop processing methods to handle problematic waste. The TWPC has several issues that have challenged the projects ability to process RH Waste. High Neutron Dose Rate resulting from both Californium and Curium in the waste stream challenge the RH-TRU 72-B limit for dose rate measured from the side of the package under normal conditions of transport, as specified in Chapter 5.0 of the RH-TRU 72-B SAR (i.e., ≤10 mrem/hour at 2 meters). Difficult to process waste in the hot cell has introduced processing and handling difficulties included problems associated with the disposition of prohibited items that fall out of the waste stream such as liquids, aerosol cans, etc. Lastly, multiple waste streams require characterization and AK challenge the ability to generate dose-to curie models for the waste. Repackaging is one solution to the high neutron dose rate issue. In parallel, an effort is underway to request a change to the TRAMPAC requirements to allow shielding in the drum or canister to reduce the impact of the high neutron dose rates. Due diligence on supporting AK efforts is important in ensuring adequate

Design and installation of a laboratory-scale system for radioactive waste treatment

International Nuclear Information System (INIS)

Berger, D.N.; Knox, C.A.; Siemens, D.H.

1980-05-01

Described are the mechanical design features and remote installation of a laboratory-scale radiochemical immobilization system which is to provide a means at Pacific Northwest Laboratory of studying effluents generated during solidification of high-level liquid radioactive waste. Detailed are the hot cell, instrumentation, two 4-in. and 12-in. service racks, the immobilization system modules - waste feed, spray calciner unit, and effluent - and a gamma emission monitor system for viewing calcine powder buildup in the spray calciner/in-can melter

Structure and short time degradation studies of sodium zirconium phosphate ceramics loaded with simulated fast breeder (FBR) waste

Energy Technology Data Exchange (ETDEWEB)

Ananthanarayanan, A., E-mail: arvinda@barc.gov.in [Process Development Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Ambashta, R.D., E-mail: aritu@barc.gov.in [Process Development Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Sudarsan, V. [Chemistry Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Ajithkumar, T. [Applied Catalysis Unit, National Chemical Laboratory, Pune 411008 (India); Sen, D.; Mazumder, S. [Solid State Physics Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Wattal, P.K. [Process Development Division, Bhabha Atomic Research Centre, Mumbai 400085 (India)

2017-04-15

Sodium zirconium phosphate (NZP) ceramics have been prepared using conventional sintering and hot isostatic pressing (HIP) routes. The structure of NZP ceramics, prepared using the HIP route, has been compared with conventionally sintered NZP using a combination of X-ray diffraction (XRD) and ({sup 31}P and {sup 23}Na) nuclear magnetic resonance (NMR) spectroscopy techniques. It is observed that NZP with no waste loading is aggressive toward the steel HIP-can during hot isostatic compaction and significant fraction of cations from the steel enter the ceramic material. Waste loaded NZP samples (10 wt% simulated FBR waste) show significantly low can-interaction and primary NZP phase is evident in this material. Upon exposure of can-interacted and waste loaded NZP to boiling water and steam, {sup 31}P NMR does not detect any major modifications in the network structure. However, the {sup 23}Na NMR spectra indicate migration of Na{sup +} ions from the surface and possible re-crystallization. This is corroborated by Small-Angle Neutron Scattering (SANS) data and Scanning Electron Microscopy (SEM) measurements carried out on these samples.

A Fire-Retardant Composite Made from Domestic Waste and PVA

Directory of Open Access Journals (Sweden)

Neni Surtiyeni

2016-01-01

Full Text Available We report the synthesis of a composite from domestic waste with the strength of wood building materials. We used original domestic waste with only a simple pretreatment to reduce the processing cost. The wastes were composed of organic components (generally originating from foods, paper, plastics, and clothes; the average fraction of each type of waste mirrored the corresponding fractions of wastes in the city of Bandung, Indonesia. An initial survey of ten landfills scattered through Bandung was conducted to determine the average fraction of each component in the waste. The composite was made using a hot press. A large number of synthesis parameters were tested to determine the optimum ones. The measured mechanical strength of the produced composite approached the mechanical properties of wood building materials. A fire-retardant powder was added to retard fire so that the composite could be useful for the construction of residential homes of lower-income people who often have problems with fire. Fire tests showed that the composites were more resistant to fire than widely used wood building materials.

16S RRNA Gene Analysis of Chlorate Reducing Thermophilic Bacteria From Local Hot Spring

OpenAIRE

Aminin, Agustina L. N; Katulistiwasari, Puri; Mulyani, Nies Suci

2011-01-01

Chlorates waste remediation by biological processes has been the object of current research. Strain CR, the chlorate reducing bacteria was isolated from Gedongsongo hot spring using minimal medium broth containing chlorates and acetate at 55oC. The determination of chlorate reduction from medium was carried out using turbidimetric method. CR isolate showed reducing ability 18% after four days of incubation. The phenotypic character of CR isolate including rod-shaped cells, gram-positive bacte...

LASER ABLATION-INDUCTIVELY COUPLED PLASMA-ATOMIC EMISSION SPECTROSCOPY STUDY AT THE 222-S LABORATORY USING HOT-CELL GLOVE BOX PROTOTYPE SYSTEM

International Nuclear Information System (INIS)

Seidel, C.M.; Jain, J.; Owens, J.W.

2009-01-01

This report describes the installation, testing, and acceptance of the Waste Treatment and Immobilization Plant (WTP) procured laser ablation-inductively coupled plasma-atomic emission spectroscopy (LA-ICP-AES) system for remotely analyzing high-level waste (HLW) samples in a hot cell environment. The work was completed by the Analytical Process Development (APD) group in accordance with Task Order 2005-003; ATS MP 1027, Management Plan for Waste Treatment Plant Project Work Performed by Analytical Technical Services. The APD group at the 222-S Laboratory demonstrated acceptable turnaround time (TAT) and provide sufficient data to assess sensitivity, accuracy, and precision of the LA-ICP-AES method

Economic comparison of crystalline ceramic and glass waste forms for HLW disposal

International Nuclear Information System (INIS)

McKee, R.W.; Daling, P.M.; Wiles, L.E.

1983-05-01

A titanate-based, crystalline ceramic produced by hot isostatic pressing has been proposed as a potentially more stable and improved waste form for high-level nuclear waste disposal compared to the currently favored borosilicate glass waste form. This paper describes the results of a study to evaluate the relative costs for disposal of high-level waste from a 70,000 metric ton equivalent (MTE) system. The entire waste management system, including waste processing and encapsulation, transportation, and final repository disposal, was included in this analysis. The repository concept is based on the current basalt waste isolation project (BWIP) reference design. A range of design basis alternatives is considered to determine if this would influence the relative economics of the two waste forms. A thermal analysis procedure was utilized to define optimum canister sizes to assure that each waste form was compared under favorable conditions. Repository costs are found to favor the borosilicate glass waste form while transportation costs greatly favor the crystalline ceramic waste form. The determining component in the cost comparison is the waste processing cost, which strongly favors the borosilicate glass process because of its relative simplicity. A net cost advantage on the order of 12% to 15% on a waste management system basis is indicated for the glass waste form

Impact Factors Analysis of the Hot Side Temperature of Thermoelectric Module

Science.gov (United States)

Zhang, Xingyu; Tan, Gangfeng; Yang, Bo

2018-03-01

The thermoelectric generator (TEG) plays a crucial role in converting the waste energy of exhaust into electricity, which ensures energy saving and increased fuel utilization efficiency. In the urban driving cycle, frequent vehicle operation, like deceleration or acceleration, results in continuous variation of the exhaust temperature. In order to make the operating performance stable, and to weaken the adverse effects of the frequent variation of the exhaust temperature on the lifetime and work efficiency of the electronic components of TEG systems, the output voltage of the thermoelectric (TE) module should stay more stable. This article provides an improved method for the temperature stability of the TE material hot side based on sandwiching material. From the view of the TEG system's average output power and the hot side temperature stability of the TE material, the analyzing factors, including the fluctuation frequency of the exhaust temperature and the physical properties and thickness of the sandwiching material are evaluated, respectively, in the sine and new European driving cycle (NEDC) fluctuation condition of the exhaust temperature. The results show few effects of sandwiching material thickness with excellent thermal conductivity on the average output power. During the 150-170 s of the NEDC test condition, the minimum hot side temperatures with a BeO ceramic thickness of 2 mm and 6 mm are, respectively, 537.19 K and 685.70 K, which shows the obvious effect on the hot side temperature stability of the BeO ceramic thickness in the process of acceleration and deceleration of vehicle driving.

Addition of electric arc furnace dust in hot metal changing the form of addition

International Nuclear Information System (INIS)

Marques Sobrinho, Vicente de Paulo Ferreira; Oliveira, Jose Roberto de; Vieira, Estefano Aparecido; Telles, Victor Bridi; Grillo, Felipe Fardin; Tenorio, Jorge Alberto Soares; Espinosa, Denise Crocce Romano

2014-01-01

This research aims to study the incorporation of the mass of electric arc furnace dust (EAFD), by addition in hot metal (1.78% Si) at a temperature of 1,400 degrees Celsius. The EAFD is from a steel plant producing long steel. The addition of the EAFD was as received, in the form of briquettes without agitation of the hot metal and in the form of briquettes with agitation of the hot metal. Previously, the EAFD was characterized using the following techniques: chemical analysis, size analysis, X-ray diffraction, scanning electron microscopy (SEM) and energy dispersive spectroscopy (EDS) microanalysis. The achievement of fusion experiments in laboratory scale, took place in a vertical tubular furnace with temperature control. The fusion experiments to assess the incorporation of EAFD mass used graphite crucibles. After cooling, the hot metal and the slag, remaining in the crucible, were weighed to do a mass balance. A flow of inert gas (argon) was maintained inside the furnace during the experiments. Results show that the experiment with addition of EAFD as received presents the best result of incorporating the mass of the final hot metal (1.73%) combined with the lowest percentage of volatilized mass of EAFD (46.52%). The experiment addition of EAFD in the form of briquette with agitation of hot metal presents the lowest percentage of slag mass (4.58%). The zinc content of volatilized EAFD (64.30%) is higher than the zinc content of the imported ore concentrate (52%) and zinc content of the national ore concentrate (12% to 39%). The presence of lead and cadmium in the slag characterizing it as a hazardous solid waste. (author)

EFFECT OF EXTRACTIVES AND CARBONIZATION TEMPERATURE ON ENERGY CHARACTERISTICS OF WOOD WASTE IN AMAZON RAINFOREST

Directory of Open Access Journals (Sweden)

Jordão Cabral Moulin

2017-06-01

Full Text Available The objective of this study was to evaluate the effect of extractives soluble in hot water, besides final carbonization temperatures, on the gravimetric yield and properties of charcoal for waste of three native forest species from the Amazon region. Waste cuttings of Ipé, Grapia and Maçaranduba species, from the machine processing for joinery of a company in the State of Pará, were used. Carbonization was carried out in an adapted electric furnace with a heating rate of 1.67°C min-1 and final temperatures of 500, 600 and 700°C. The waste was carbonized fresh after extraction in hot water to remove extractives. Gravimetric yields were analyzed, as well as chemical features and high heating value. In the evaluation of the experiment, arranged in a factorial scheme with three factors (species x temperature x material with and without extraction, and Principal Component Analysis used too. The presence of extractives (soluble in hot water from wood waste had little influence on the gravimetric yield and immediate chemical composition of charcoal; however, it showed a greater high heating value and lower contents of hydrogen and nitrogen. The increase in the final carbonization temperature reduced the gravimetric yield in charcoal, the content of volatile materials and hydrogen, with a higher content of fixed carbon, carbon and high heating value. The treatments with the best energy characteristics were obtained from Ipé and Maçaranduba charcoals with extractives produced at 600°C, in addition to Ipê and Maçaranduba charcoals with and without extractives obtained at 700°C.

Decontamination of alpha-bearing solid wastes and plutonium recovery

International Nuclear Information System (INIS)

Koehly, G.; Madic, C.; Lecomte, M.; Bourges, J.; Saulze, J.L.; Broudic, J.C.

1993-01-01

Nuclear activities in the Radiochemistry building of Fontenay-aux-Roses Nuclear Research Center concern principally the study of fuel reprocessing and the production of transuranium isotopes. During these activities solid wastes are produced. In order to improve the management of these wastes, it has been decided to build new facilities: a group of three glove-boxes named ELISE for the treatment of α active solid waste and a hot-cell, PROLIXE, for the treatment of solid wastes. Leaching processes were developed in order to: decontaminate these wastes and recover actinide elements, particularly the highly valuable plutonium, from the leachates. The processes developed are sufficiently flexible to be able to accommodate solid wastes produced in other facilities. Laboratory studies were conducted to develop the leaching process based on the use of electrogenerated Ag(II) species which is particularly suitable to provoke the dissolution of PuO 2 . Successful exhaustive Pu decontaminations with DF(Pu) higher than 10 4 were achieved for the first time during the treatment of stainless steel PuO 2 cans (future MELOX plant) by electrogenerated Ag (II) in nitric acid medium

Hot Surface Ignition

OpenAIRE

Tursyn, Yerbatyr; Goyal, Vikrant; Benhidjeb-Carayon, Alicia; Simmons, Richard; Meyer, Scott; Gore, Jay P.

2015-01-01

Undesirable hot surface ignition of flammable liquids is one of the hazards in ground and air transportation vehicles, which primarily occurs in the engine compartment. In order to evaluate the safety and sustainability of candidate replacement fuels with respect to hot surface ignition, a baseline low lead fuel (Avgas 100 LL) and four experimental unleaded aviation fuels recommended for reciprocating aviation engines were considered. In addition, hot surface ignition properties of the gas tu...

The evaporative drying of sludge by immersion in hot oil: Effects of oil type and temperature

Energy Technology Data Exchange (ETDEWEB)

Ohm, Tae-In, E-mail: tiohm1@hanbat.ac.kr [Department of Environmental Engineering, Hanbat National University, San 16-1 Duckmyung-dong, Yusung-gu, Daejeon 305-719 (Korea, Republic of); Chae, Jong-Seong; Lim, Kwang-Soo [Department of Environmental Engineering, Hanbat National University, San 16-1 Duckmyung-dong, Yusung-gu, Daejeon 305-719 (Korea, Republic of); Moon, Seung-Hyun [Waste Energy Research Center, Korea Institute of Energy Research, Jang-dong Yusung-gu, Daejeon 305-343 (Korea, Republic of)

2010-06-15

We investigated the evaporative drying by immersion in hot oil (EDIHO) method for drying sludge. This involved heating oil to a temperature higher than that needed for moisture to be evaporated from the sludge by turbulent heat and mass transfer. We fry-dried sewage and leather plant sludge for 10 min in each of four different oils (waste engine, waste cooking, refined waste, and B-C heavy) and three different temperatures (140 deg. C, 150 deg. C, and 160 deg. C). Drying efficiency was found to be greater for higher temperatures. However, giving consideration to energy efficiency we suggest that the optimal temperature for fry-drying sludge is 150 deg. C. At 150 deg. C, the water content of sewage sludge reduced from 78.9% to between 1.5% (with waste cooking oil) and 3.8% (with waste engine oil). The reduction in water content for leather plant sludge fry-dried at 150 deg. C was from 81.6% to between 1% (with waste cooking oil) and 6.5% (with refined waste oil). The duration of the constant rate-drying period was also influenced by the type of oil used: refined waste oil > waste engine oil > B-C heavy oil > waste cooking oil. The duration at 150 deg. C with waste cooking oil was 3 min for sewage sludge and 2 min for leather plant sludge. It is likely that the drying characteristics of oil are influenced by its thermal properties, including its specific heat, and molecular weight.

The evaporative drying of sludge by immersion in hot oil: Effects of oil type and temperature

International Nuclear Information System (INIS)

Ohm, Tae-In; Chae, Jong-Seong; Lim, Kwang-Soo; Moon, Seung-Hyun

2010-01-01

We investigated the evaporative drying by immersion in hot oil (EDIHO) method for drying sludge. This involved heating oil to a temperature higher than that needed for moisture to be evaporated from the sludge by turbulent heat and mass transfer. We fry-dried sewage and leather plant sludge for 10 min in each of four different oils (waste engine, waste cooking, refined waste, and B-C heavy) and three different temperatures (140 deg. C, 150 deg. C, and 160 deg. C). Drying efficiency was found to be greater for higher temperatures. However, giving consideration to energy efficiency we suggest that the optimal temperature for fry-drying sludge is 150 deg. C. At 150 deg. C, the water content of sewage sludge reduced from 78.9% to between 1.5% (with waste cooking oil) and 3.8% (with waste engine oil). The reduction in water content for leather plant sludge fry-dried at 150 deg. C was from 81.6% to between 1% (with waste cooking oil) and 6.5% (with refined waste oil). The duration of the constant rate-drying period was also influenced by the type of oil used: refined waste oil > waste engine oil > B-C heavy oil > waste cooking oil. The duration at 150 deg. C with waste cooking oil was 3 min for sewage sludge and 2 min for leather plant sludge. It is likely that the drying characteristics of oil are influenced by its thermal properties, including its specific heat, and molecular weight.

Mineral-modeled ceramics for long-term storage of high-level nuclear wastes

International Nuclear Information System (INIS)

Vance, E.R.

1980-01-01

Over the past ten years, Penn State's Materials Research Laboratory has done extensive work on mineral-modeled ceramics for high-level nuclear waste storage. These ceramics are composed of several mineral analogues that form a monolithic polycrystalline aggregate. Mineral-modeling can be made in a similar fashion to nuclear waste glasses, and their naturally occurring analogues are known to last millions, and even billions, of years in hot, wet conditions. It is believed that such ceramics could reduce dispersal of radionuclides by leaching to a minimum

Bituminization of liquid radioactive wastes. Part 1

International Nuclear Information System (INIS)

Gradev, G.D.; Ivanov, V.I.; Stefanova, I.G.; Milusheva, A.G.; Guteva, E.S.; Zhelyazkov, V.T.; Stefanov, G.I.; G'oshev, G.S.

1991-01-01

Salt-bitumen products are produced by the method of 'hot mixing' of some Bulgarian bitumens (road bitumen PB 66/99 and the hydroinsulating bitumen HB 80/25) and salts (chlorides, sulphates, borates, salt mixtures modelling the liquid waste from nuclear power plants) in different ratios to determine the optimum conditions for bituminization of liquid radioactive waste. The penetration, ductility and softening temperature were determined. The sedimentation properties and the thermal resistance of the various bitumen-salt mixtures were studied. The most suitable bitumen for technological research at the Kozloduy NPP was found to be the road bitumen PB 66/90 with softening temperature at 48 o C. The optimum amount of salts incorporated in the bitumen - about 45% - was found. No exothermal effects were observed in the bituminization process in the temperature range of up to 200 o C. The results obtained may be useful in the elaboration of a technology for bituminization of liquid radioactive wastes in the Kozloduy NPP. 4 tabs., 5 figs., 4 refs

Relationship between hot spot residues and ligand binding hot spots in protein-protein interfaces.

Science.gov (United States)

Zerbe, Brandon S; Hall, David R; Vajda, Sandor; Whitty, Adrian; Kozakov, Dima

2012-08-27

In the context of protein-protein interactions, the term "hot spot" refers to a residue or cluster of residues that makes a major contribution to the binding free energy, as determined by alanine scanning mutagenesis. In contrast, in pharmaceutical research, a hot spot is a site on a target protein that has high propensity for ligand binding and hence is potentially important for drug discovery. Here we examine the relationship between these two hot spot concepts by comparing alanine scanning data for a set of 15 proteins with results from mapping the protein surfaces for sites that can bind fragment-sized small molecules. We find the two types of hot spots are largely complementary; the residues protruding into hot spot regions identified by computational mapping or experimental fragment screening are almost always themselves hot spot residues as defined by alanine scanning experiments. Conversely, a residue that is found by alanine scanning to contribute little to binding rarely interacts with hot spot regions on the partner protein identified by fragment mapping. In spite of the strong correlation between the two hot spot concepts, they fundamentally differ, however. In particular, while identification of a hot spot by alanine scanning establishes the potential to generate substantial interaction energy with a binding partner, there are additional topological requirements to be a hot spot for small molecule binding. Hence, only a minority of hot spots identified by alanine scanning represent sites that are potentially useful for small inhibitor binding, and it is this subset that is identified by experimental or computational fragment screening.

High-temperature and high-power-density nanostructured thermoelectric generator for automotive waste heat recovery

International Nuclear Information System (INIS)

Zhang, Yanliang; Cleary, Martin; Wang, Xiaowei; Kempf, Nicholas; Schoensee, Luke; Yang, Jian; Joshi, Giri; Meda, Lakshmikanth

2015-01-01

Highlights: • A thermoelectric generator (TEG) is fabricated using nanostructured half-Heusler materials. • The TE unicouple devices produce superior power density above 5 W/cm"2. • A TEG system with over 1 kW power output is demonstrated by recovering automotive waste heat. - Abstract: Given increasing energy use as well as decreasing fossil fuel sources worldwide, it is no surprise that interest in promoting energy efficiency through waste heat recovery is also increasing. Thermoelectric generators (TEGs) are one of the most promising pathways for waste heat recovery. Despite recent thermoelectric efficiency improvement in nanostructured materials, a variety of challenges have nevertheless resulted in few demonstrations of these materials for large-scale waste heat recovery. Here we demonstrate a high-performance TEG by combining high-efficiency nanostructured bulk materials with a novel direct metal brazing process to increase the device operating temperature. A unicouple device generates a high power density of 5.26 W cm"−"2 with a 500 °C temperature difference between hot and cold sides. A 1 kW TEG system is experimentally demonstrated by recovering the exhaust waste heat from an automotive diesel engine. The TEG system operated with a 2.1% heat-to-electricity efficiency under the average temperature difference of 339 °C between the TEG hot- and cold-side surfaces at a 550 °C exhaust temperature. The high-performance TEG reported here open up opportunities to use TEGs for energy harvesting and power generation applications.

Characterization of mixed CH-TRU waste at Argonne-West

International Nuclear Information System (INIS)

Dwight, C.C.; Guay, K.P.; Courtney, J.C.; Higgins, P.J.

1993-01-01

Argonne National Laboratory is participating in the Department of Energy's Waste Isolation Pilot Plant (WIPP) Experimental Test Program by characterizing and repackaging mixed contact-handled transuranic waste. Argonne's initial activities in the Program were described last year at Waste Management '92. Since then, additional waste has been characterized and repackaged, resulting in six bins ready for shipment to WIPP upon the initiation of the bin tests. Lessons learned from these operations are being factored in the design and installation of a new characterization facility, the Enhanced Waste Characterization Facility (EWCF). The objectives of the WIPP Experimental Test Program have also undergone change since last year leading to an accelerated effort to factor sludge sampling capability into the EWCF. Consequently, the initiation of non-sludge operations in the waste characterization chamber has been delayed to Summer 1993 while the sludge sampling modifications are incorporated into the facility. Benefits in operational flexibility, effectiveness, and efficiency and reductions in potential facility and personnel contamination and exposure are expected from the enhanced waste characterization facility within the Hot Fuel Examination Facility at Argonne-West. This paper summarizes results and lessons learned from recent characterization and repackaging efforts and future plans for characterization. It also describes design features and status of the EWCF

Energy flux of hot atoms

International Nuclear Information System (INIS)

Wotzak, G.P.; Kostin, M.D.

1976-01-01

The process in which hot atoms collide with thermal atoms of a gas, transfer kinetic energy to them, and produce additional hot atoms is investigated. A stochastic method is used to obtain numerical results for the spatial and time dependent energy flux of hot atoms in a gas. The results indicate that in hot atom systems a front followed by an intense energy flux of hot atoms may develop

Preliminary report for the license of a hot cell that will be use in the technology development for the obtention of Mo-99

International Nuclear Information System (INIS)

Fucugauchi, L.A.; Millan S, S.; Lopez M, A.E.; Lopez C, R; Sanchez M, V.; Reynoso V, R.; Vera, A.

1991-05-01

A preliminary report for the license of a hot cell that will be used in the development of the technology for the obtaining of Mo-99 is presented. The following topics are also included: objective of the project, technical description, description of the prototype cell, handling of radioactive wastes, lists of equipment that will be used, risk analysis, curricula, quality assurance plan and an annex with the report on handling of radioactive wastes presented to the PAGD-IAEA. (Author)

Modelling Hot Air Balloons.

Science.gov (United States)

Brimicombe, M. W.

1991-01-01

A macroscopic way of modeling hot air balloons using a Newtonian approach is presented. Misleading examples using a car tire and the concept of hot air rising are discussed. Pressure gradient changes in the atmosphere are used to explain how hot air balloons work. (KR)

Denitration and chemical precipitation of medium level liquid wastes and conditioning of high level wastes from low level liquid wastes by a roll dryer and subsequent vitrification

International Nuclear Information System (INIS)

Halaszovich, S.; Dix, S.; Harms, R.

1987-01-01

Medium level liquid waste (MAW) from the reprocessing need after being fixed in cement an additional shielding to meet required radiation limits for handling and transportation. Normally this shielding consists of concrete and its weight and volume is several times higher than that of the waste product itself. By means of caesium separation using nickel-potassium-hexacyanoferrate and after few years of interim storage waiting for the decay of Ruthenium and Antimony the activities will be reduced below permissible values. (13 MBq/l in waste solution for Cs, 28 MBq/l for Sb and 34 MBq/l for Ru). Below these limits there is no need for additional shielding after cementation in a 400 l drum. Experimental results show, that Caesium can be precipitated and separated effectively not only in laboratory but also in a larger scale under hot cell conditions. The process investigated in this work has been developed from the FIPS process for vitrification of highly radioactive fission product solutions. It consists of: denitration, precipitation, sludge separation, drying and melting

Applications of thermal energy storage to waste heat recovery in the food processing industry

Science.gov (United States)

Trebilcox, G. J.; Lundberg, W. L.

1981-03-01

The canning segment of the food processing industry is a major energy user within that industry. Most of its energy demand is met by hot water and steam and those fluids, in addition to product cooling water, eventually flow from the processes as warm waste water. To minimize the possibility of product contamination, a large percentage of that waste water is sent directly to factory drains and sewer systems without being recycled and in many cases the thermal energy contained by the waste streams also goes unreclaimed and is lost from further use. Waste heat recovery in canning facilities can be performed economically using systems that employ thermal energy storage (TES). A project was proposed in which a demonstration waste heat recovery system, including a TES feature, would be designed, installed and operated.

The results of decontamination and decommissioning of experimental DUPIC equipment at PIEF 9405 hot cell

Energy Technology Data Exchange (ETDEWEB)

Lee, H. H.; Park, J. J.; Shin, J. M.; Cho, K. H.; Yang, M. S.; Lee, E. P. [Korea Atomic Energy Research Institute, Taejeon (Korea)

2002-02-01

The characterization experiment for powder and sintered fuel had been performed using about 1 kg-U spent PWR fuel at No. 9405 hot-cell in PIEF(Post Irradiated Experiment Facility) since early in 1999. Currently, the experiments in PIEF have been completed. Since all DUPIC equipment in hot-cell are contaminated by high radioactive material, the decontamination and dismantlement must be performed remotely by M/S manipulator. During the radioactive waste packing and transportation, the reduction method of radiation exposure has to be considered. This report describes the basic plan for dismantlement/decontamination of the characterization equipment (power and sintered fuel). And methods of measurement/packing/transportation, method of dismantlement/decontamination of the experimental apparatus and the reduction method of radiation dose exposure, etc. are explained in order. 7 refs., 42 figs., 10 tabs. (Author)

Software Simulation of Hot Tearing

DEFF Research Database (Denmark)

Andersen, S.; Hansen, P.N.; Hattel, Jesper Henri

1999-01-01

The brittleness of a solidifying alloy in a temperature range near the solidus temperature has been recognised since the fifties as the mechanism responsible for hot tearing. Due to this brittlenes, the metal will crack under even small amounts of strain in that temperature range. We see these hot...... tears in castings close to hot centres, where the level of strain is often too high.Although the hot tearing mechanism is well understood, until now it has been difficult to do much to reduce the hot tearing tendency in a casting. In the seventies, good hot tearing criteria were developed by considering...... the solidification rate and the strain rate of the hot tear prone areas. But, until recently it was only possible to simulate the solidification rate, so that the criteria could not be used effectively.Today, with new software developments, it is possible to also simulate the strain rate in the hot tear prone areas...

Progress report on research and development work 1991 of the Department of Hot Chemistry, Karlsruhe Nuclear Research Center

International Nuclear Information System (INIS)

1991-03-01

In the year under review, the Institute of Hot Chemistry (IHCH) was in the midst of a thematic reorientation process. The priority of future chemical-technical work will be in the field of the development of supercriticality processes. The objective of such work consists in seeking new ways for getting rid of resistant chemical pollutants (halogenated organic compounds). The following projects are presented in detail: 1) Waste control in the environment (communal waste management; water and soil; emission-reducing processes; highly polluted soils); 2) Solid state and materials research (chemistry of materials research); basic physical research (neutrino and particle physics); 3) Nuclear waste management (concluding work on reprocessing technology), and 4) Other research projects (Institute-related research). The Annex lists the publications made by the IHCH staff. (BBR) [de

A multifunction radioactive waste monitoring system

International Nuclear Information System (INIS)

Edeline, J.C.; Libs, G.

1991-01-01

The monitoring of unknown radioactive transuranic wastes mixed with fission products (FP) needs several measuring technics: passive and active neutron methods, gamma rays spectrometry and, sometimes, emission tomography to localize the hot points in the waste packages. The goal is to achieve a whole system from the most up-dated electronics sub-assemblies to provide these characterization measurement at the lowest cost and in the simplest manner. The control of the different measurements is made by only one micro-processor and an unusual way of using the gamma spectrometry A.D.C. and multichannel analyser makes possible to control the neutron analogic electronics: neutron counter high-voltage supplies, amplifiers and discriminators; many of the gamma spectrometry sub-assemblies are also used for the gamma emission tomography. The different measurements are automated and different programs offer the possibility to choice the proper measurement methods for each item. The waste package handling apparatus is not included in the system but the control of such handling might be performed by the micro-computer. We describe the main parts and features of the system [fr

Effect of aluminum and silicon reactants and process parameters on glass-ceramic waste form characteristics for immobilization of high-level fluorinel-sodium calcined waste

International Nuclear Information System (INIS)

Vinjamuri, K.

1993-06-01

In this report, the effects of aluminum and silicon reactants, process soak time and the initial calcine particle size on glass-ceramic waste form characteristics for immobilization of the high-level fluorinel-sodium calcined waste stored at the Idaho Chemical Processing Plant (ICPP) are investigated. The waste form characteristics include density, total and normalized elemental leach rates, and microstructure. Glass-ceramic waste forms were prepared by hot isostatically pressing (HIPing) a pre-compacted mixture of pilot plant fluorinel-sodium calcine, Al, and Si metal powders at 1050 degrees C, 20,000 psi for 4 hours. One of the formulations with 2 wt % Al was HIPed for 4, 8, 16 and 24 hours at the same temperature and pressure. The calcine particle size range include as calcined particle size smaller than 600 μm (finer than -30 mesh, or 215 μm Mass Median Diameter, MMD) and 180 μm (finer than 80 mesh, or 49 μm MMD)

Hot tub folliculitis

Science.gov (United States)

... survives in hot tubs, especially tubs made of wood. Symptoms The first symptom of hot tub folliculitis ... may help prevent the problem. Images Hair follicle anatomy References D'Agata E. Pseudomonas aeruginosa and other ...

Operational Strategies for Low-Level Radioactive Waste Disposal Site in Egypt - 13513

International Nuclear Information System (INIS)

Mohamed, Yasser T.

2013-01-01

The ultimate aims of treatment and conditioning is to prepare waste for disposal by ensuring that the waste will meet the waste acceptance criteria of a disposal facility. Hence the purpose of low-level waste disposal is to isolate the waste from both people and the environment. The radioactive particles in low-level waste emit the same types of radiation that everyone receives from nature. Most low-level waste fades away to natural background levels of radioactivity in months or years. Virtually all of it diminishes to natural levels in less than 300 years. In Egypt, The Hot Laboratories and Waste Management Center has been established since 1983, as a waste management facility for LLW and ILW and the disposal site licensed for preoperational in 2005. The site accepts the low level waste generated on site and off site and unwanted radioactive sealed sources with half-life less than 30 years for disposal and all types of sources for interim storage prior to the final disposal. Operational requirements at the low-level (LLRW) disposal site are listed in the National Center for Nuclear Safety and Radiation Control NCNSRC guidelines. Additional procedures are listed in the Low-Level Radioactive Waste Disposal Facility Standards Manual. The following describes the current operations at the LLRW disposal site. (authors)

Performance of asphalt mixture incorporating recycled waste

Science.gov (United States)

Hamid, Nor Baizura; Abdullah, Mohd Ezree; Sanik, Mohd Erwan; Mokhtar, Mardiha; Kaamin, Masiri; Raduan, Rasyidah; Ramli, Mohd Zakwan

2017-12-01

Nowadays, the amount of premix waste was increased every year, especially at the batching plants. Normally, the waste materials will be discarded without doing any innovative and effective research about those materials. This situation has become one of the global concerns due to the increasing number of premix waste produced every year. Therefore, the aim of this study is to evaluate the performance of hot mix asphalt (HMA) using premix waste on improving asphalt mixture fatigue behaviour. The method used in this study was Superpave mix design method. The sample conducted in this study were 0%, 10%, 20%, 30%, and 100% of premix waste respectively. For a binder test, the laboratory test conducted were penetration test, softening test and thin film oven test while for the performance test were resilient modulus test and indirect tensile fatigue test. From the laboratory test, the resilient modulus test was conducted with two different temperature which was 25°C and 40°C. The result from that test was 20% of premix waste had higher resilient modulus at that two different temperatures compared to another samples. From that test also shown that the sample at the lower temperature which was 25°C has higher resilient modulus compared to the temperature of 40°C. Indirect tensile fatigue test showed that the 30% of premix waste sample was suitable for the modified asphalt mixture with referring to the maximum deformation and strain for comparison control, 10%,20%, and 100% of premix waste samples. So, it can be concluded that premix waste inhibits great potential as road construction material and suitable for repeated traffic loading.

Annual report on operation, utilization and technical development of hot laboratories. From April 1, 1996 to March 31, 1997

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-02-01

This report describes activities, in fiscal year 1996, of the Reactor Fuel Examination Facility (RFEF), the Research Hot Laboratory (RHL) and the Waste Safety Testing Facility (WASTEF) which belong to the Department of Hot laboratories. In the RFEF, Post-Irradiation Examinations (PIEs) of PWR fuel assemblies irradiated in the Takahama Unit 3, a BWR fuel assembly irradiated in the Fukusima Daini Unit have been performed. Also, PIEs of assembly materials irradiated in the Fugen Reactor have been carried out. To support R and D works in JAERI, refabrication of segmented fuel rods have been done using irradiated LWR fuel rods for pulse irradiation in the NSRR and re-irradiation tests in the JMTR. PIEs have been performed on high burnup fuel rods and ROX fuel rods. For the RHL, PIEs have been performed on segment fuels irradiated in the NSRR, fuels and materials for HTTR, standard fuels for JRR-3M and materials for nuclear fusion reactor. In addition, a monitoring test of fuel elements in accordance with the surveillance program of the Magnox reactor of the Japan Atomic Power Corporation has been continued. In the WASTEF, leaching tests on TRU in simulated glass forms and a low flow rate tests on glass waste forms have been carried out. The examinations of alpha damage acceleration for the Synroc waste forms have also been performed. (author)

Hot conditioning equipment conceptual design report

International Nuclear Information System (INIS)

Bradshaw, F.W.

1996-01-01

This report documents the conceptual design of the Hot Conditioning System Equipment. The Hot conditioning System will consist of two separate designs: the Hot Conditioning System Equipment; and the Hot Conditioning System Annex. The Hot Conditioning System Equipment Design includes the equipment such as ovens, vacuum pumps, inert gas delivery systems, etc.necessary to condition spent nuclear fuel currently in storage in the K Basins of the Hanford Site. The Hot Conditioning System Annex consists of the facility of house the Hot Conditioning System. The Hot Conditioning System will be housed in an annex to the Canister Storage Building. The Hot Conditioning System will consist of pits in the floor which contain ovens in which the spent nuclear will be conditioned prior to interim storage

Hot conditioning equipment conceptual design report

Energy Technology Data Exchange (ETDEWEB)

Bradshaw, F.W., Westinghouse Hanford

1996-08-06

This report documents the conceptual design of the Hot Conditioning System Equipment. The Hot conditioning System will consist of two separate designs: the Hot Conditioning System Equipment; and the Hot Conditioning System Annex. The Hot Conditioning System Equipment Design includes the equipment such as ovens, vacuum pumps, inert gas delivery systems, etc.necessary to condition spent nuclear fuel currently in storage in the K Basins of the Hanford Site. The Hot Conditioning System Annex consists of the facility of house the Hot Conditioning System. The Hot Conditioning System will be housed in an annex to the Canister Storage Building. The Hot Conditioning System will consist of pits in the floor which contain ovens in which the spent nuclear will be conditioned prior to interim storage.

Recovering energy and materials from hazardous waste

Energy Technology Data Exchange (ETDEWEB)

Anon.

2003-12-01

The tannery industry faces growing environmental concerns because of the high hazardous metal content of its process waste. The formation, during the tanning process, of the highly toxic hexavalent chromium precludes the use of conventional thermal incineration processes. Borge Tannery in Norway, which processes 600 cattle hides per day, has solved the problem by using new PyroArc technology. The PyroArc waste processing plant can treat all of the tannery's production wastes, transforming them into useful products such as fuel gas and re-usable metal. The fuel gas consists mainly of carbon monoxide, hydrogen and nitrogen, and has a calorific value of about 4 MJ/Nm{sub 3}. About 65-70% of the energy content of the source material (waste or biomass) is recovered in the gas, and this is used to produce steam and/or electricity in a gas engine with a capacity of 580 kW. A further 20-25% of the initial energy content is recovered as heat or low-pressure steam. The plant is designed to be self-sufficient in energy (1.5 MW) and to meet the tannery's maximum requirements for hot water and steam. (UK)

IEN low-level radioactive waste management

International Nuclear Information System (INIS)

Rocha, A.C.S. da; Pina, J.L.S.; Silva, S. da; Silva, J.J.G.

1986-09-01

The low-level radioactive waste produced in Instituto de Engenharia Nuclear is generated basically from three distinct modes: a particle accelerator (CV-28 Cyclotron), radiochemistry laboratories and the operation of a nuclear research reactor (Argonaut type). In the Cyclotron unit, all water flow from hot labs as well as from the decontamination laundry is retained in special tank with homogenizing system and a remote control, that signalizes when the tank gets a pre-specified level. Samples homogenized from the tank are colected for previous analysis. (Author) [pt

Friction welded closures of waste canisters

International Nuclear Information System (INIS)

Klein, R.F.

1987-01-01

Liquid radioactive waste presently stored in underground tanks is to undergo a vitrifying process which will immobilize it into a solid form. This solid waste will be contained in a stainless steel canister. The canister opening requires a positive-seal weld, the properties and thickness of which must be at least equal to those of the canister material. All studies and tests performed in the work discussed in this paper have the inertia friction welding concept to be highly feasible in this application. This paper describes the decision to investigate the inertia friction welding process, the inertia friction welding process itself, and a proposed equipment design concept. This system would provide a positive, reliable, inspectable, and full-thickness seal weld while utilizing easily maintainable equipment. This high-quality weld can be achieved even in highly contaminated hot cell

Cleanout and decontamination of radiochemical hot cells

International Nuclear Information System (INIS)

Surma, J.E.; Holton, L.K. Jr.; Katayama, Y.B.; Gose, J.E.; Haun, F.E.; Dierks, R.D.

1990-01-01

The Pacific Northwest Laboratory is developing and employing advanced remote and contact technologies in cleaning out and decontaminating six radiochemical hot cells at Hanford under the Department of Energy's Surplus Facilities Management Program. The program is using a series of remote and contact decontamination techniques to reduce costs and to significantly lower radiation doses to workers. Refurbishment of the cover blocks above the air lock trench reduced radiation exposure in the air lock and cleanout and decontamination of an analytical cell achieved a reduction in radioactive contamination. Nuclear Regulatory Commission-approved Type B burial boxes are also being used to reduce waste disposal costs and radiation doses. PNL is currently decommissioning its pilot-scale radioactive liquid-fed ceramic melter. Special tools have been developed and are being used to accomplish the world's first such effort. 4 refs., 5 figs

Value-added utilisation of recycled concrete in hot-mix asphalt.

Science.gov (United States)

Wong, Yiik Diew; Sun, Darren Delai; Lai, Dickson

2007-01-01

The feasibility of partial substitution of granite aggregate in hot-mix asphalt (HMA) with waste concrete aggregate was investigated. Three hybrid HMA mixes incorporating substitutions of granite fillers/fines with 6%, 45% untreated, and 45% heat-treated concrete were evaluated by the Marshall mix design method; the optimum binder contents were found to be 5.3%, 6.5% and 7.0% of grade Pen 60/70 bitumen, respectively. All three hybrid mixes satisfied the Marshall criteria of the Singapore Land Transport Authority (LTA) W3B wearing course specification. The hybrid mix with 6% concrete fillers gave comparable resilient modulus and creep resistance as the conventional W3B mix, while hybrid mixes with higher concrete substitutions achieved better performance. X-ray diffraction (XRD) showed the distinct presence of free lime in the heat-treated concrete, while the scanning electron microscope (SEM) provided an in-depth perspective of the concrete grains in the HMA matrix. The results suggest feasible use of waste concrete as partial aggregate substitution in HMA.

Biodegradation of ethyl acetate in radioactive liquid organic waste by bacterial communities

International Nuclear Information System (INIS)

Ferreira, Rafael V.P.; Sakata, Solange K.; Borba, Tania R.; Bellini, Maria H.; Marumo, Julio T.; Dutra, Fernando

2009-01-01

The research and development program in reprocessing of low burn-up spent fuel elements began in Brazil in 70's, originating the lab -scale hot cell, known as CELESTE located at IPEN-CNEN/SP. The program was ended at the beginning of 90's and part of the radioactive waste generated mainly from the analytical laboratories is stored at the Waste Management Laboratory. Among various types of radioactive waste generated, the organic liquid represents a major problem for its management, because it can not be directly solidified with cement. The objective of this work is to develop a pretreatment methodology to degrade the ethyl acetate present in organic liquid waste so that it can subsequently be immobilized in cement. This work was divided into two parts: selection and adaptation of three bacterial communities for growth in medium containing ethyl acetate and degradation experiments of ethyl acetate present in radioactive organic liquid waste. The results showed that from bacterial communities the highest biodegradation level observed was 77%. (author)

Operating experience and radiation protection problems in the working of the radio-metallurgy hot cell facilities at BARC

International Nuclear Information System (INIS)

Janardhanan, S.; Watamwar, S.B.; Mehta, S.K.; Pillai, P.M.B.; John, Jacob; Kutty, K.N.

1977-01-01

The Bhabha Atomic Research Centre at Bombay has six hot cell facilities for radiometallurgical investigations of irradiated/failed fuel elements. The hot cell facilities have been provided with certain built-in safety features, a ventilation system, radiation monitoring instruments for various purposes, a centralised air monitoring system and a central panel for display of various alarms. Procedures adopted for radiation protection and contamination control include : (1) radiation leak test for cells and filter efficiency evaluation before cell activation, (2) practices to be followed by frog suit personnel while working in hot cell areas, (3) receipt and handling of irradiated fuel elements, (4) cell filter change operation, (5) checks on high level drains and (6) effluent discharge and waste shipments. Operating experience in the working of these facilities along with radiation accident incidents is described. Data regarding release of activity during normal cell operations, dose rates during various metallurgical operations and personnel exposures are presented. (M.G.B.)

Structure modification of natural zeolite for waste removal application

Science.gov (United States)

Widayatno, W. B.

2018-03-01

Tremendous industrialization in the last century has led to the generation of huge amount of waste. One of the recent hot research topics is utilizing any advance materials and methods for waste removal. Natural zeolite as an inexpensive porous material with a high abundance holds a key for efficient waste removal owing to its high surface area. However, the microporous structure of natural zeolite hinders the adsorption of waste with a bigger molecular size. In addition, the recovery of natural zeolite after waste adsorption into its pores should also be considered for continuous utilization of this material. In this study, the porosity of natural zeolite from Tasikmalaya, Indonesia, was hydrothermally-modified in a Teflon-lined autoclave filled with certain pore directing agent such as distilled water, KOH, and NH4OH to obtain hierarchical pore structure. After proper drying process, the as-treated natural zeolite is impregnated with iron cation and heat-treated at specified temperature to get Fe-embedded zeolite structure. XRD observation is carried out to ensure the formation of magnetic phase within the zeolite pores. The analysis results show the formation of maghemite phase (γ-Fe2O3) within the zeolite pore structure.

Process for reclaiming tungsten from a hazardous waste

International Nuclear Information System (INIS)

Scheithauer, R.A.; MacInnis, M.B.; Miller, M.J.; Vanderpool, C.D.

1984-01-01

A process is disclosed wherein tungsten is recovered from hazardous waste material containing said tungsten, arsenic, and other impurities which can consist of magnesium, phosphorus, and silicon and the resulting waste is treated to render it nonhazardous according to EPA standards for arsenic. Said process involves digesting said hazardous waste material in an aqueous solution of an alkali metal hydroxide, adjusting the pH of the resulting solution to about 11.0 to about 13.0 with NaOH to precipitate essentially all of the magnesium and silicon species, filtering the digestion mix to remove the solids from said resulting solution which contains about 80 to about 100% of said tungsten and essentially none of said magnesium and said silicon, slurrying the hazardous solids in hot water, and adding to the slurry a ferric salt solution to precipitate ferric hydroxide, filtering this mixture to give a solid which passes the EPA standard test for solids with respect to arsenic

10 CFR 431.102 - Definitions concerning commercial water heaters, hot water supply boilers, and unfired hot water...

Science.gov (United States)

2010-01-01

... supply boilers, and unfired hot water storage tanks. 431.102 Section 431.102 Energy DEPARTMENT OF ENERGY... Water Heaters, Hot Water Supply Boilers and Unfired Hot Water Storage Tanks § 431.102 Definitions concerning commercial water heaters, hot water supply boilers, and unfired hot water storage tanks. The...

Techniques for sampling nuclear waste tank contents and in situ measurement of activity

International Nuclear Information System (INIS)

Lawrence, R.C.

1978-04-01

A study was conducted to develop suitable sampling equipment and techniques for characterizing the mechanical properties of nuclear wastes; identifying effective means of measuring radiation levels, temperatures, and neutron fluxes in situ in wastes; and developing a waste core sampler. A portable, stainless steel probe was developed which is placed in the tank through a riser. This probe is built for the insertion of instrumentation that can measure the contents of the tank at any level and take temperature, radiation, and neutron activation readings with reliable accuracy. A simple and reliable instrument for the in situ extraction of waste materials ranging from liquid to concrete-like substances was also developed. This portable, stainless steel waste core sampler can remove up to one liter of radioactive waste from tanks for transportation to hot cell laboratories for analysis of hardness, chemical form, and isotopic content. A cask for transporting the waste samples from the tanks to the laboratory under radiation-protected conditions was also fabricated. This cask was designed with a ''boot'' or inner-seal liner to contain any radioactive wastes that might remain on the outside of the waste core sampling device

Powder technological vitrification of simulated high-level waste

International Nuclear Information System (INIS)

Gahlert, S.

1988-03-01

High-level waste simulate from the reprocessing of light water reactor and fast breeder fuel was vitrified by powder technology. After denitration with formaldehyde, the simulated HLW is mixed with glass frit and simultaneously dried in an oil-heated mixer. After 'in-can calcination' for at least 24 hours at 850 or 950 K (depending on the type of waste and glass), the mixture is hot-pressed in-can for several hours at 920 or 1020 K respectively, at pressures between 0.4 and 1.0 MPa. The technology has been demonstrated inactively up to diameters of 30 cm. Leach resistance is significantly enhanced when compared to common borosilicate glasses by the utilization of glasses with higher silicon and aluminium content and lower sodium content. (orig.) [de

Electropolishing decontamination system for high-level waste canisters

International Nuclear Information System (INIS)

Larson, D.E.; Berger, D.N.; Allen, R.P.; Bryan, G.H.; Place, B.G.

1988-10-01

As part of a US Department of Energy (DOE) project agreement with the Federal Ministry for Research and Technology (BMFT) in the Federal Republic of Germany (FRG). The Nuclear Waste Treatment Program at the Pacific Northwest Laboratory (PNL) is preparing 30 radioactive canisters containing borosilicate glass for use in high-level waste repository related tests at the Asse Salt Mine. After filling, the canisters will be welded closed and decontaminated in preparation for shipping to the FRG. Electropolishing was selected as the primary decontamination approach, and an electropolishing system with associated canister inspection equipment has been designed and fabricated for installation in a large hot cell. This remote electropolishing system, which is currently undergoing preliminary testing, is described in this report. 3 refs., 3 figs., 1 tab

Repository waste-handling operations, 1998

International Nuclear Information System (INIS)

Cottam, A.E.; Connell, L.

1986-04-01

The Civilian Radioactive Waste Management Program Mission Plan and the Generic Requirements for a Mined Geologic Disposal System state that beginning in 1998, commercial spent fuel not exceeding 70,000 metric tons of heavy metal, or a quantity of solidified high-level radioactive waste resulting from the reprocessing of such a quantity of spent fuel, will be shipped to a deep geologic repository for permanent storage. The development of a waste-handling system that can process 3000 metric tons of heavy metal annually will require the adoption of a fully automated approach. The safety and minimum exposure of personnel will be the prime goals of the repository waste handling system. A man-out-of-the-loop approach will be used in all operations including the receipt of spent fuel in shipping casks, the inspection and unloading of the spent fuel into automated hot-cell facilities, the disassembly of spent fuel assemblies, the consolidation of fuel rods, and the packaging of fuel rods into heavy-walled site-specific containers. These containers are designed to contain the radionuclides for up to 1000 years. The ability of a repository to handle more than 6000 pressurized water reactor spent-fuel rods per day on a production basis for approximately a 23-year period will require that a systems approach be adopted that combines space-age technology, robotics, and sophisticated automated computerized equipment. New advanced inspection techniques, maintenance by robots, and safety will be key factors in the design, construction, and licensing of a repository waste-handling facility for 1998

Decommissioning program and future plan for research hot laboratory (2)

International Nuclear Information System (INIS)

Koya, Toshio; Nozawa, Yukio; Hanada, Yasushi; Ono, Katsuto; Kanazawa, Hiroyuki; Nihei, Yasuo; Owada, Isao

2010-01-01

The Research Hot Laboratory (RHL) in Japan Atomic Energy Agency (JAEA) was constructed in 1961, as the first one in JAPAN, to perform the examinations of irradiated fuels and materials. RHL consists of 10 heavy concrete cells and 38 lead cells, which had been contributed to research and development program in or out of JAEA for the investigation of irradiation behavior for fuels and nuclear materials. However, RHL is the one of target as the rationalization program for decrepit facilities in former Tokai institute. Therefore the decommissioning works of RHL have been started on April 2003. The decommissioning work will be progressing, dismantling the lead cells and decontamination of concrete caves then release in the regulation of controlled area. The 18 lead cells (including semi-hot cell and junior-cell) had been dismantled. Removal of the applause from the cells, survey of the contamination revel in the lead cells and prediction of radio active waste have been finished as the preparing work for dismantling of the remained 20 lead cells. The future plan of decommissioning work has been prepared to incarnate the basic vision and dismantling procedure. (author)

Radiological Characterization Methodology for INEEL-Stored Remote-Handled Transuranic (RH TRU) Waste from Argonne National Laboratory-East

International Nuclear Information System (INIS)

Kuan, P.; Bhatt, R.N.

2003-01-01

An Acceptable Knowledge (AK)-based radiological characterization methodology is being developed for RH TRU waste generated from ANL-E hot cell operations performed on fuel elements irradiated in the EBR-II reactor. The methodology relies on AK for composition of the fresh fuel elements, their irradiation history, and the waste generation and collection processes. Radiological characterization of the waste involves the estimates of the quantities of significant fission products and transuranic isotopes in the waste. Methods based on reactor and physics principles are used to achieve these estimates. Because of the availability of AK and the robustness of the calculation methods, the AK-based characterization methodology offers a superior alternative to traditional waste assay techniques. Using the methodology, it is shown that the radiological parameters of a test batch of ANL-E waste is well within the proposed WIPP Waste Acceptance Criteria limits

Strategic Planning for Hot Cell Closure

International Nuclear Information System (INIS)

LANGSTAFF, D.C.

2001-01-01

The United States Department of Energy (DOE) and its contractor were remediating a large hot cell complex to mitigate the radiological hazard. A Resource Conservation and Recovery Act (RCRA) closure unit was determined to be located within the complex. The regulator established a challenge to develop an acceptable closure plan on a short schedule (four months). The scope of the plan was to remove all excess equipment and mixed waste from the closure unit, establish the requirements of the legally binding Closure Plan and develop an acceptable schedule. The complex has several highly radioactive tanks, tank vaults, piping, and large hot cells containing complex chemical processing equipment. Driven by a strong need to develop an effective strategy to meet cleanup commitments, three principles were followed to develop an acceptable plan: (1) Use a team approach, (2) Establish a buffer zone to support closure, and (3) Use good practice when planning the work sequence. The team was composed of DOE, contractor, and Washington State Department of Ecology (Regulator) staff. The team approach utilized member expertise and fostered member involvement and communication. The buffer zone established an area between the unregulated parts of the building and the areas that were allegedly not in compliance with environmental standards. Introduction of the buffer zone provided simplicity, clarity, and flexibility into the process. Using good practice means using the DOE Integrated Safety Management Core Functions for planning and implementing work safely. Paying adequate attention to detail when the situation required contributed to the process credibility and a successful plan

A supramolecular structure insight for conversion property of cellulose in hot compressed water: Polymorphs and hydrogen bonds changes.

Science.gov (United States)

Wang, Yan; Lian, Jie; Wan, Jinquan; Ma, Yongwen; Zhang, Yingshi

2015-11-20

Waste paper samples with different cellulose supramolecular structure were treated in hot compressed water (HCW) at 375°C and 22.5MPa within 200s to evaluate the specific effect mechanism of cellulose supramolecular structure on the conversion of waste paper to reusable resource. Although the distribution of liquid products and the oligosaccharides were related to reaction time, depolymerization and decrystallization of the cellulose, the characteristics absorption peak of cellulose from FTIR analysis and crystal structure of the cellulose detected in the residues with hydrolysis rate up 96.5% indicated crystal structure was the dominant factor that affect conversion behavior of waste paper. The conversion of cellulose Iβ to cellulose Iα or cellulose I(α+β) in HCW demonstrated that the recrystallization occurred during the decrystallization of cellulose through the rearrangement of hydrogen bonds. Copyright © 2015 Elsevier Ltd. All rights reserved.

Waste treatment at the Radiochemical Engineering Development Center

International Nuclear Information System (INIS)

Brunson, R.R.; Bond, W.D.; Chattin, F.R.; Collins, R.T.; Sullivan, G.R.; Wiles, R.H.

1997-01-01

At the Radiochemical Engineering Development Center (REDC) irradiated targets are processed for the recovery of valuable radioisotopes, principally transuranium nuclides. A system was recently installed for treating the various liquid alkaline waste streams for removal of excess radioactive contaminants at the REDC. Radionuclides that are removed will be stored as solids and thus the future discharge of radionuclides to liquid low level waste tank storage will be greatly reduced. The treatment system is of modular design and is installed in a hot cell (Cubicle 7) in Building 7920 at the REDC where preliminary testing is in progress. The module incorporates the following: (1) a resorcinol-formaldehyde resin column for Cs removal, (2) a cross flow filtration unit for removal of rare earths and actinides as hydroxide, and (3) a waste solidification unit. Process flowsheets for operation of the module, key features of the module design, and its computer-assisted control system are presented. Good operability of the cross flow filter system is mandatory to the successful treatment of REDC wastes. Results of tests to date on the operation of the filter in its slurry collection mode and its slurry washing mode are presented. These tests include the effects of entrained organic solvent in the waste stream feed to the filter

Fluidized-bed incineration plant equipped with waste heat boilers. Developed for mid-size municipalities

Energy Technology Data Exchange (ETDEWEB)

Handa, Hitoshi

1988-01-20

A fluidized bed incineration plant with a waste heat boiler was installed to dispose wastes in Sakura City on March, 1987 and has waste disposing capacity of 120tons/d. Sands are fluidized in the furnace at 700-800/sup 0/C and wastes are burned completely for a short time. The waste heat boiler is used to utilize waste heat to send steam to aquiculturing farms and hot water to the community plaza and further supplies steam to two 90kW back pressure turbines for driving forced draft fan used for the incineration plant. Harmful gases in waste gas are removed by the harmful gas eliminator to lower HCl to 120ppm or less and K value of SOx to 9.0 or less and then cleaned gas is exhausted through the electostatic precipitator and the chimney. Dust and fly ash are transferred to a reservior through a superior seal tight air transportation system, pelletized and disposed for land fill. Bulk waste disposing capacity is 50 tons/d and harmful wastes, magnetic materials, unburnable and burnable wastes are classified and separated. Separated iron purity is 95% or more. (4 figs, 2 photos)

Hot springs in Hokuriku District

Energy Technology Data Exchange (ETDEWEB)

Sato, K. (Hot Springs Research Center, Japan)

1971-01-01

In the Hokuriku district including Toyama, Ishikawa, and Fukui Prefectures, hot springs of more than 25/sup 0/C were investigated. In the Toyama Prefecture, there are 14 hot springs which are located in an area from the Kurobe River to the Tateyama volcano and in the mountainous area in the southwest. In Ishikawa Prefecture there are 16 hot springs scattered in Hakusan and its vicinity, the Kaga mountains, and in the Noto peninsula. In northern Fukui Prefecture there are seven hot springs. The hot springs in Shirakawa in Gifu Prefecture are characterized as acid springs producing exhalations and H/sub 2/S. These are attributed to the Quaternary volcanoes. The hot springs of Wakura, Katayamazu, and Awara in Ishikawa Prefecture are characterized by a high Cl content which is related to Tertiary andesite. The hot springs of Daishoji, Yamanaka, Yamashiro, Kuritsu, Tatsunokuchi, Yuwaku, and Yunotani are characterized by a low HCO/sub 3/ content. The Ca and SO/sub 4/ content decreases from east to west, and the Na and Cl content increases from west to east. These fluctuations are related to the Tertiary tuff and rhyolite. The hot springs of Kuronagi, Kinshu, and Babadani, located along the Kurobe River are characterized by low levels of dissolved components and high CO/sub 2/ and HCO/sub 3/ content. These trends are related to late Paleozoic granite. Hot springs resources are considered to be connected to geothermal resources. Ten tables, graphs, and maps are provided.

Decontamination and dismantlement of the old hot laundry, CFA-669. Final report

International Nuclear Information System (INIS)

Smith, D.L.; Perry, E.F.

1995-01-01

This final report describes the decontamination and dismantlement (D ampersand D) of the old hot laundry, located at the Idaho National Engineering Laboratory (INEL) Central Facilities Area (CFA). The report describes the site before and after D ampersand D, processes used, cost and duration, and waste volume generated. In addition, lessons learned are presented. Pre-D ampersand D characterization indicated gross alpha concentrations in the building ranged from 6 to 310 pCi/g and gross beta measurements from 6 to 15,000 pCi/g. Gamma spectrum analysis identified cobalt-60, cesium-137, antimony-125, europium-152, europium-154, and niobium-94

Agricultural waste as a source for the production of silica nanoparticles.

Science.gov (United States)

Vaibhav, Vineet; Vijayalakshmi, U; Roopan, S Mohana

2015-03-15

The major interest of the paper deals with the extraction of silica from four natural sources such as rice husk, bamboo leaves, sugarcane bagasse and groundnut shell. These waste materials in large quantities can create a serious environmental problem. Hence, there is a need to adopt proper strategy to reduce the waste. In the present investigation, all the waste materials are subjected to moisture removal in a hot plate and sintered at 900°C for 7 h. The sintered powder was treated with 1 M NaOH to form sodium silicate and then with 6M H2SO4 to precipitate silica. The prepared silica powders were characterized by FT-IR, XRD and SEM-EDAX analysis. The silica recovered from different sources was found to vary between 52% and 78%. Magnesium substituted silica was formed from the groundnut waste and further treatment is required to precipitate silica. Copyright © 2014 Elsevier B.V. All rights reserved.

Detection of Hot Halo Gets Theory Out of Hot Water

Science.gov (United States)

2006-02-01

Scientists using NASA's Chandra X-ray Observatory have detected an extensive halo of hot gas around a quiescent spiral galaxy. This discovery is evidence that galaxies like our Milky Way are still accumulating matter from the gradual inflow of intergalactic gas. "What we are likely witnessing here is the ongoing galaxy formation process," said Kristian Pedersen of the University of Copenhagen, Denmark, and lead author of a report on the discovery. Chandra observations show that the hot halo extends more than 60,000 light years on either side of the disk of the galaxy known as NGC 5746. The detection of such a large halo alleviates a long-standing problem for the theory of galaxy formation. Spiral galaxies are thought to form from enormous clouds of intergalactic gas that collapse to form giant, spinning disks of stars and gas. Chandra X-ray Image of NGC 5746 Chandra X-ray Image of NGC 5746 One prediction of this theory is that large spiral galaxies should be immersed in halos of hot gas left over from the galaxy formation process. Hot gas has been detected around spiral galaxies in which vigorous star formation is ejecting matter from the galaxy, but until now hot halos due to infall of intergalactic matter have not been detected. "Our observations solve the mystery of the missing hot halos around spiral galaxies," said Pedersen. "The halos exist, but are so faint that an extremely sensitive telescope such as Chandra is needed to detect them." DSS Optical Image of NGC 5746 DSS Optical Image of NGC 5746 NGC 5746 is a massive spiral galaxy about a 100 million light years from Earth. Its disk of stars and gas is viewed almost edge-on. The galaxy shows no signs of unusual star formation, or energetic activity from its nuclear region, making it unlikely that the hot halo is produced by gas flowing out of the galaxy. "We targeted NGC 5746 because we thought its distance and orientation would give us the best chance to detect a hot halo caused by the infall of

Development of a desalination system driven by solar energy and low grade waste heat

International Nuclear Information System (INIS)

Elminshawy, Nabil A.S.; Siddiqui, Farooq R.; Sultan, Gamal I.

2015-01-01

Highlights: • Productivity increases significantly up to critical waste gas flow rate. • Productivity decreases for waste gas flow rate higher than critical flow rate. • Increasing evaporator inlet waste gas temperature increases productivity. • The proposed system coupled with combined cycle has a fuel saving 1844 kg/h. • The cost of potable water produced is 0.014 USD/L. - Abstract: Various thermal power systems emit flue gases containing significant amount of waste energy. The aim of this research is to recover a valuable amount of this energy to develop an efficient desalination system coupled with solar energy. Experiments were performed in the month of June 2014 at Al-Qassim, Saudi Arabia (26°4′53″N, 43°58′32″E) for different hot air (waste gas) flow rates and evaporator inlet water temperature to study the effect on daily potable water productivity. The proposed setup comprised an evaporator, condenser, air blower, electric heaters, storage tank and evacuated tube solar collectors. It was found that increasing the hot air flow rate increases the water productivity up to the critical flow rate after which the productivity decreases. Analytical model was developed for this desalination setup and the results were compared to that obtained from experiments. The overall daily (9 AM–5 PM) potable water productivity of the proposed system is about 50 L for corresponding useful waste heat varying from 130 to 180 MJ/day and a global solar radiation on a horizontal surface ranging from 15 to 29 MJ/m 2 /day. Water is produced at the cost of 0.014 USD/L and the fuel saving equal to 1844 kg/h is achieved for the proposed desalination system

Hot Gas Halos in Galaxies

Science.gov (United States)

Mulchaey, John

Most galaxy formation models predict that massive low-redshift disk galaxies are embedded in extended hot halos of externally accreted gas. Such gas appears necessary to maintain ongoing star formation in isolated spirals like the Milky Way. To explain the large population of red galaxies in rich groups and clusters, most galaxy evolution models assume that these hot gas halos are stripped completely when a galaxy enters a denser environment. This simple model has been remarkably successful at reproducing many observed properties of galaxies. Although theoretical arguments suggest hot gas halos are an important component in galaxies, we know very little about this gas from an observational standpoint. In fact, previous observations have failed to detect soft X-ray emission from such halos in disk galaxies. Furthermore, the assumption that hot gas halos are stripped completely when a galaxy enters a group or cluster has not been verified. We propose to combine proprietary and archival XMM-Newton observations of galaxies in the field, groups and clusters to study how hot gas halos are impacted by environment. Our proposed program has three components: 1) The deepest search to date for a hot gas halo in a quiescent spiral galaxy. A detection will confirm a basic tenet of disk galaxy formation models, whereas a non-detection will seriously challenge these models and impose new constraints on the growth mode and feedback history of disk galaxies. 2) A detailed study of the hot gas halos properties of field early-type galaxies. As environmental processes such as stripping are not expected to be important in the field, a study of hot gas halos in this environment will allow us to better understand how feedback and other internal processes impact hot gas halos. 3) A study of hot gas halos in the outskirts of groups and clusters. By comparing observations with our suite of simulations we can begin to understand what role the stripping of hot gas halos plays in galaxy

Progress report on safety research of high-level waste management for the period April 1987 to March 1988

International Nuclear Information System (INIS)

Nakamura, Haruto; Tashiro, Shingo

1988-10-01

Researches on high-level waste management at the High Level Waste Management Laboratory and the Waste Safety Testing Facility Operation Division of the Japan Atomic Energy Research Institute in the fiscal year of 1987 are reviewed in the three sections of the report. The topics are as follows: 1) On performance and durability of waste forms and engineered barrier materials, accelerated alpha radiation stability of glass form and Synroc has been investigated and stress corrosion cracking of canister materials was examined under simulated conditions. 2) Sorption of 237 Np on granite samples and behavior of iron during weathering of granites were studied with respect to safety evaluation for geological disposal. 3) Actual waste was transported from the Tokai Reprocessing Plant and hot operation using the actual waste was initiated at WASTEF. (author)

Seismic modifications to the hot suspect repair area Argonne National Laboratory, West

International Nuclear Information System (INIS)

Malik, L.E.; Harris, B.G.

1993-01-01

The ANL-W WIPP Waste Facility Enhancement Project required substantial remodeling and upgrades to the Hot Fuels Examination Facility (HFEF) building, including the Hot and Suspect Repair Area (HSRA). The HSRA is an enclosed single-stoned area inside the HFEF. It is separated into several compartments, some of which are used for handling radioactive materials. The HSRA roof consists of 18 GA steel Robertson Q decking with 1.5 in. concrete topping, and is utilized for storage. Braced steel frames support the HSRA roof, except at the north side, where the steel beams arc connected to the HFEF columns. The HSRA has hollow block masonry perimeter and interior walls. Seismic evaluations concluded that the HSRA did not have a competent seismic force resisting system. The structure was upgraded by decoupling it from the HFEF framing for N/S motions, modifying two existing braced frames, adding a now braced frame that can be removed temporarily during maintenance and strengthening the roof diaphragm by a unique modification consisting of special epoxy grout and steel plates installed over the existing concrete roof

Hot flashes and sleep in women.

Science.gov (United States)

Moe, Karen E

2004-12-01

Sleep disturbances during menopause are often attributed to nocturnal hot flashes and 'sweats' associated with changing hormone patterns. This paper is a comprehensive critical review of the research on the relationship between sleep disturbance and hot flashes in women. Numerous studies have found a relationship between self-reported hot flashes and sleep complaints. However, hot flash studies using objective sleep assessment techniques such as polysomnography, actigraphy, or quantitative analysis of the sleep EEG are surprisingly scarce and have yielded somewhat mixed results. Much of this limited evidence suggests that hot flashes are associated with objectively identified sleep disruption in at least some women. At least some of the negative data may be due to methodological issues such as reliance upon problematic self-reports of nocturnal hot flashes and a lack of concurrent measures of hot flashes and sleep. The recent development of a reliable and non-intrusive method for objectively identifying hot flashes during the night should help address the need for substantial additional research in this area. Several areas of clinical relevance are described, including the effects of discontinuing combined hormone therapy (estrogen plus progesterone) or estrogen-only therapy, the possibility of hot flashes continuing for many years after menopause, and the link between hot flashes and depression.

Waste minimization through high-pressure microwave digestion of soils for gross α/β analyses

International Nuclear Information System (INIS)

Yaeger, J.S.; Smith, L.L.

1995-04-01

As a result of the U.S. Department of Energy's (DOE) environmental restoration and waste management activities, laboratories receive numerous analytical requests for gross α/β analyses. Traditional sample preparation methods for gross α/β analysis of environmental and mixed waste samples require repetitive leaching, which is time consuming and generates large volumes of secondary wastes. An alternative to leaching is microwave digestion. In the past. microwave technology has had limited application in the radiochemical laboratory because of restrictions on sample size resulting from vessel pressure limitations. However, new microwave vessel designs allow for pressures on the order of 11 MPa (1500 psi). A procedure is described in which microwave digestion is used to prepare environmental soil samples for gross α/β analysis. Results indicate that the described procedure meets performance requirements for several soil types and is equivalent to traditional digestion techniques. No statistical differences at the 95% confidence interval exist between the measurement on samples prepared from the hot plate and microwave digestion procedures for those soils tested. Moreover, microwave digestion allows samples to be prepared in a fraction of the time with significantly less acid and with lower potential of cross-contamination. In comparison to the traditional hot plate method, the waste volumes required for the microwave procedure are a factor of 10 lower, while the analyst time for sample processing is at least a factor of three lower

Supplement analysis of transuranic waste characterization and repackaging activities at the Idaho National Engineering Laboratory in support of the Waste Isolation Pilot Plant test program

International Nuclear Information System (INIS)

1991-03-01

This supplement analysis has been prepared to describe new information relevant to waste retrieval, handling, and characterization at the Idaho National Engineering Laboratory (INEL) and to evaluate the need for additional documentation to satisfy the National Environmental Policy Act (NEPA). The INEL proposes to characterize and repackage contact-handled transuranic waste to support the Waste Isolation Pilot Plant (WIPP) Test Phase. Waste retrieval, handling and processing activities in support of test phase activities at the WIPP were addressed in the Supplemental Environmental Impact Statement (SEIS) for the WIPP. To ensure that test-phase wastes are properly characterized and packaged, waste containers would be retrieved, nondestructively examined, and transported from the Radioactive Waste Management Complex (RWMC) to the Hot-Fuel Examination Facility for headspace gas analysis, visual inspections to verify content code, and waste acceptance criteria compliance, then repackaging into WIPP experimental test bins or returned to drums. Following repackaging the characterized wastes would be returned to the RWMC. Waste characterization would help DOE determine WIPP compliance with US Environmental Protection Agency regulations governing disposal of transuranic waste and hazardous waste. Additionally, this program supports onsite compliance with Resource Conservation and Recovery Act (RCRA) requirements, compliance with the terms of the No-Migration Variance at WIPP, and provides data to support future waste shipments to WIPP. This analysis will help DOE determine whether there have been substantial changes made to the proposed action at the INEL, or if preparation of a supplement to the WIPP Final Environmental Impact Statement (DOE, 1980) and SEIS (DOE, 1990a) is required. This analysis is based on current information and includes details not available to the SEIS

Development of low grade waste heat thermoelectric power generator

Directory of Open Access Journals (Sweden)

Suvit Punnachaiya

2010-07-01

Full Text Available This research aimed to develop a 50 watt thermoelectric power generator using low grade waste heat as a heat source,in order to recover and utilize the excess heat in cooling systems of industrial processes and high activity radioisotope sources. Electricity generation was based on the reverse operation of a thermoelectric cooling (TEC device. The TEC devices weremodified and assembled into a set of thermal cell modules operating at a temperature less than 100°C. The developed powergenerator consisted of 4 modules, each generating 15 watts. Two cascade modules were connected in parallel. Each modulecomprised of 96 TEC devices, which were connected in series. The hot side of each module was mounted on an aluminumheat transfer pipe with dimensions 12.212.250 cm. Heat sinks were installed on the cold side with cooling fans to provideforced air cooling.To test electricity generation in the experiment, water steam was used as a heat source instead of low grade waste heat.The open-circuit direct current (DC of 250 V and the short-circuit current of 1.2 A was achieved with the following operatingconditions: a hot side temperature of 96°C and a temperature difference between the hot and cold sides of 25°C. The DC poweroutput was inverted to an AC power source of 220 V with 50 Hz frequency, which can continuously supply more than 50 wattsof power to a resistive load as long as the heat source was applied to the system. The system achieved an electrical conversionefficiency of about 0.47 percent with the capital cost of 70 US$/W.

Web technology in the separation of strontium and cesium from INEL-ICPP radioactive acid waste (WM-185)

International Nuclear Information System (INIS)

Bray, L.A.; Brown, G.N.

1995-01-01

Strontium and cesium were successfully removed from radioactive acidic waste (WM-185) at the Idaho National Engineering Laboratory, Idaho Chemical Processing Plant (ICPP), with web technology from 3M and IBC Advanced Technologies, Inc. (IBC). A technical team from Pacific Northwest Laboratory, ICPP, 3M and IBC conducted a very successful series of experiments from August 15 through 18, 1994. The ICPP, Remote Analytical Laboratory, Idaho Falls, Idaho, provided the hot cell facilities and staff to complete these milestone experiments. The actual waste experiments duplicated the initial 'cold' simulated waste results and confirmed the selective removal provided by ligand-particle web technology

Waste energy boosts tomato industry at distillery

Energy Technology Data Exchange (ETDEWEB)

McColl, J

1989-04-01

A trial project aimed at using waste hot water from the cooling process at a Scottish whisky distillery to heat a glasshouse for tomato production is described. Later developments have involved the installation of a waste heat boiler to make use of the heat from the still burner flue gases. Steam from the boiler is used within the distillery and to supplement the glasshouse system. The payback within the distillery industry has been excellent, but tomato production, though continuing, was adversely affected by severe cutbacks in distillery production in the early eighties. Recently further significant savings have been made in the distillery industry by the installation of a regenerative burner in one of the stills and thermo-compressors in the cooling tower condensers to produce low pressure steam which can be fed back into the system. (U.K.).

Treatment of solid waste highly contaminated by alpha emitters: Low-temperature impact crushing, leaching and incineration

International Nuclear Information System (INIS)

Bertolotti, G.; Vigreux, B.; Caillol, A.; Koehly, G.

1987-01-01

Reprocessing plants, hot laboratories and fuel fabrication plants produce solid wastes containing residual amounts of plutonium and uranium in nitrate and oxide form at concentrations up to several tens of grams per m/sup 3/. Dismantling of nuclear facilities having handled these radioelements also generates large volumes of solid wastes highly contaminated with alpha emitters. It is desirable to process these alpha wastes to recover valuable fissile materials and/or permit surface storage. Solid waste treatment by low-temperature impact crushing and then leaching, after minimal sorting and classifying at the sites of production, meets the corresponding requirements for high volume reduction plus fissile material recovery or waste decontamination. Additional volume reduction of crushed wastes containing mainly combustible materials can be obtained by incineration. This is facilitated by the low fissile material content after low-temperature impact crushing and leaching. Sorted wastes can also be leached or incinerated directly after, in most cases, crushing by more conventional techniques

Update on the KELT Transit Survey: Hot Planets around Hot Stars

Science.gov (United States)

Gaudi, B. Scott; Stassun, Keivan G.; Pepper, Joshua; KELT Collaboration

2018-01-01

The KELT Transit Survey consists of a pair of small-aperture, wide-angle automated telescopes located at Winer Observatory in Sonoita, Arizona and the South African Astronomical Observatory (SAAO) in Sutherland, South Africa. Together, they are surveying roughly 70% of the sky for transiting planets. By virtue of their small apertures (42 mm) and large fields-of-view (26 degrees x 26 degrees), KELT is most sensitive to hot Jupiters transiting relatively bright (V~8-11), and thus relatively hot stars. I will provide an update on the planets discovered by KELT, focusing in detail on our recent discoveries of very hot planets transiting several bright A and early F stars.

Hot subluminous star: HDE 283048

International Nuclear Information System (INIS)

Laget, M.; Vuillemin, A.; Parsons, S.B.; Henize, K.G.; Wray, J.D.

1978-01-01

The star HDE 283048, located at α = 3/sup h/50/sup m/.3, delta = +25 0 36', shows a strong ultraviolet continuum. Ground-based observations indicate a hot-dominated composite spectrum. Several lines of evidence suggest that the hot component is a hot subdwarf. 2 figures

Hot Isostatic Pressing of Engineered Forms of I-AgZ

Energy Technology Data Exchange (ETDEWEB)

Jubin, Robert Thomas [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Watkins, Thomas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bruffey, Stephanie H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jordan, Jacob A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Parks, Mackenzie L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-11-01

Hot isostatic pressing (HIP) is being considered for direct conversion of ¹²⁹I-bearing materials to a radiological waste form. The removal of volatile radioactive ¹²⁹I from the off-gas of a nuclear fuel reprocessing facility will be necessary to comply with regulatory requirements regarding reprocessing facilities sited within the United States, and any iodine-containing media or solid sorbents generated by offgas abatement will require disposal. Zeolite minerals such as silver-exchanged mordenite (AgZ) have been studied as potential iodine sorbents and will contain ¹²⁹I as chemisorbed AgI. Oak Ridge National Laboratory (ORNL) has conducted several recent studies on the HIP of both iodine-loaded AgZ (I-AgZ) and other iodine-bearing zeolite minerals. The goal of these research efforts is to achieve a stable, highly leach resistant material that is reduced in volume as compared to bulk iodine-loaded I-AgZ. Through the use of HIP, it may be possible to achieve this with the addition of little or no additional materials (waste formers). Other goals for the process include that the waste form will be tolerant to high temperatures and pressures, not chemically hazardous, and that the process will result in minimal secondary waste generation. This document describes the preparation of 27 samples that are distinct from previous efforts in that they are prepared exclusively with an engineered form of AgZ that is manufactured using a binder. Iodine was incorporated solely by chemisorption. This base material is expected to be more representative of an operational system than were samples prepared previously with pure minerals.

Second generation waste package design and storage concept for the Yucca Mountain Repository

International Nuclear Information System (INIS)

Armijo, Joseph Sam; Kar, Piyush; Misra, Manoranjan

2006-01-01

The reference waste package design and operating mode to be used in the Yucca Mountain Repository is reviewed. An alternate (second generation) operating concept and waste package design is proposed to reduce the risk of localized corrosion of waste packages and to reduce repository costs. The second generation waste package design and storage concept is proposed for implementation after the initial licensing and operation of the reference repository design. Implementation of the second generation concept at Yucca Mountain would follow regulatory processes analogous to those used successfully to extend the design life and uprate the power of commercial light water nuclear reactors in the United States. The second generation concept utilizes the benefits of hot dry storage to minimize the potential for localized corrosion of the waste package by liquid electrolytes. The second generation concept permits major reductions in repository costs by increasing the number of fuel assemblies stored in each waste package, by eliminating the need for titanium drip shields and by fabricating the outer container from corrosion resistant low alloy carbon steel

Waste canister closure welding using the inertia friction welding process

International Nuclear Information System (INIS)

Klein, R.F.; Siemens, D.H.; Kuruzar, D.L.

1986-02-01

Liquid radioactive waste presently stored in underground tanks is to undergo a vitrifying process which will immobilize it in a solid form. This solid waste will be contained in a stainless steel canister. The canister opening requires a positive seal weld, the properties and thickness of which are at least equal to those of the canister material. This paper describes the inertia friction welding process and a proposed equipment design concept that will provide a positive, reliable, inspectable, and full thickness seal weld while providing easily maintainable equipment, even though the weld is made in a highly contaminated hot cell. All studies and tests performed have shown the concept to be highly feasible. 2 refs., 6 figs

The development of a mobile hot cell facility for the conditioning of spent high activity radioactive sources (SHARS)

International Nuclear Information System (INIS)

Liebenberg, G.R.; Al-Mughrabi, M.

2008-01-01

The International Atomic Energy Agency (IAEA) Waste Technology Section with additional support from the U.S. National Nuclear Security Agency (NNSA) through the IAEA Nuclear Security Fund has funded the design, fabrication, evaluation, and testing of a portable hot cell intended to address the problem of disused SHARS in obsolete irradiation devices such as teletherapy heads and dry irradiators. The project is initially targeting the African continent but expected soon to expand to Latin America and Asia. This hot cell would allow source removal, characterization, consolidation, repackaging in modern storage shields, and secure storage of high risk SHARS at single sites in each IAEA Member State. The mobile hot cell and related equipment is transported in two shipping containers to a specific country where the following process takes place: 1-) Assembly of hot cell; 2-) Removal of SHARS from working shields, encapsulation into a stainless steel capsule and placement into a long term storage shield; 3-) Conditioning of any other spent sources the country may require; 4-) Dismantling of the hot cell; 5-) Shipping equipment out of country. The operation in a specific country is planned to be executed over a three week period. This presentation will discuss the development of the mobile hot cell facility as well as the demonstration of the state of readiness of the system for manipulation of SHARS and the planned execution of the conditioning operations. As a result of this project, excess SHARS could be managed safely and securely and possibly be more easily repatriated to their country of origin for appropriate final disposition. (author)

Development of a ceramic waste form for high-level waste disposal

International Nuclear Information System (INIS)

Esh, D. W.

1998-01-01

A ceramic waste form is being developed by Argonne National Laboratory (ANL) as part of the demonstration of the electrometallurgical treatment of spent nuclear fuel. The halide, alkaline earth, alkali, transuranic, and rare earth fission products are stabilized in zeolite which is combined with glass and processed in a hot isostatic press (HIP) to form a ceramic composite. The mineral sodalite is formed in the HIP from the zeolite precursor. The process, from starting materials to final product, is relatively simple. An overview of the processing operations is given. The metrics that have been developed to measure the success or completion of processing operations are developed and discussed. The impact of variability in processing metrics on the durability of the final product is presented

A study on decontamination and decommissioning of experimental DUPIC equipment at PIEF 9405 hot cell

International Nuclear Information System (INIS)

Lee, H. H.; Park, J. J.; Shin, J. M.; Yang, M. S.; Lee, H. S.; Lee, E. P.

2000-09-01

The characterization experiment for powder and sintered fuel had been performed using about 1 kg-U spent PWR fuel at No. 9405 hot-cell in PIEF(Post Irradiated Experiment Facility) since early in 1999. Currently, The experiments in PIEF have been completed. It is supposed to dismantle and decontaminate the installed equipment by the end of year 2000. Since all of DUPIC equipment in hot-cell are contaminated by high radioactive material, the decontamination and dismantlement must br performed remotely by M/S manipulator. During the radioactive waste packing and transportation, the reduction method of radiation exposure has to be considered. Firstly, This report describes the basic plan for dismantlement/decontamination of the characterization equipment(power and sintered fuel). And methods of measurement/packing/ transportation, method of dismantlement/decontamination of the experimental apparatus and the reduction method of radiation dose exposure, etc. are explained in order

Silica based gel as a potential waste form for high level waste from fuel reprocessing

International Nuclear Information System (INIS)

Ford, C.E.; Dempster, T.J.; Melling, P.J.

1983-10-01

To assess the feasibility of safe disposal of high-level radioactive waste as synthetic clay, or material that would react with ground water to form clay, experiments have been carried out to determine the hydrothermal crystallisation and leaching behaviour of silica based gels fired at 900 deg C. Crystallisation rates at a pressure of 500 bars and at temperatures below 400 deg C are negligible and this more or less precludes pre-disposal production of synthetic clay on the scale required. Leaching experiments suggest that the leach rates of Cs from gels by distilled water are higher than those of boro-silicate glasses and SYNROC at the lower temperatures that would be preferred for geological storage. However, amounts of bulk dissolution of gels may be lower than those of boro-silicate glasses. The initial leaching behaviour of gels might be considerably improved by hot compaction at 900 to 1000 deg C. Consideration of likely waste form dissolution behaviour in a repository environment suggests that gels of appropriate composition might perform as well as, or better than, boro-silicate glasses. A novel hypothetical plant is described that could produce the gel waste form on the scale required on a more or less continuous basis. (author)

Statistical hot spot analysis of reactor cores

International Nuclear Information System (INIS)

Schaefer, H.

1974-05-01

This report is an introduction into statistical hot spot analysis. After the definition of the term 'hot spot' a statistical analysis is outlined. The mathematical method is presented, especially the formula concerning the probability of no hot spots in a reactor core is evaluated. A discussion with the boundary conditions of a statistical hot spot analysis is given (technological limits, nominal situation, uncertainties). The application of the hot spot analysis to the linear power of pellets and the temperature rise in cooling channels is demonstrated with respect to the test zone of KNK II. Basic values, such as probability of no hot spots, hot spot potential, expected hot spot diagram and cumulative distribution function of hot spots, are discussed. It is shown, that the risk of hot channels can be dispersed equally over all subassemblies by an adequate choice of the nominal temperature distribution in the core

Characterization of Sludge from the Process of Steel Tubes Chemical Treatment for Hot Galvanizing

Directory of Open Access Journals (Sweden)

Sofilić, U.

2009-10-01

Full Text Available Inadequate industrial waste management in Croatia is reflected in the non-sanitary waste disposal, low recycling levels, negligible share of waste processing technologies, insufficient control of its flows, etc.Generated industrial wastes are most frequently disposed of at producers’ own, mostly illegal landfills. There are many such landfills on the Croatian territory, and the disposed types of waste often include those that can be hazardous and represent a considerable source of environmental pollution.Past waste management in all industrial branches can be characterized in this way, which at the same time may result in the harmful impact on human health and the environment. It also represents economic loss due to low utilisation of material and energy potential of some industrial wastes. The metallurgical industry collects its production waste separately. Only a part of the generated waste is returned to the production process and some waste is occasionally used by other industries as secondary raw materials, but the largest part of it ends at producers' own landfills on site. Hazardous wastes (dust containing heavy metals, waste oils etc. are mostly disposed of in a controlled and lawful manner. Past handling of metallurgical waste was unacceptable both from the environmental and economic point of view. Therefore a systematic resolving of this important issue was initiated at the beginning of this decade. Sisak Steelworks galvanized steel pipes in the hot-dip galvanizing procedure by immersing in molten zinc. Between 1970 and 2000 Sisak Steelworks produced approximately 900 000 tonnes of galvanized pipes this way and generated around 70 000 m3 of neutralisation sludge, which was subsequently disposed of in the landfill on site. The paper presents the results of examination of physical-chemical properties of neutralisation sludge generated as waste material in the process of neutralisation of waste sulphate acid bath used in Sisak

PACCOM: A nuclear waste packaging facility cost model: Draft technical report

International Nuclear Information System (INIS)

Dippold, D.G.; Tzemos, S.; Smith, D.J.

1985-05-01

PACCOM is a computerized, parametric model used to estimate the capital, operating, and decommissioning costs of a variety of nuclear waste packaging facility configurations. The model is based upon a modular waste packaging facility concept from which functional components of the overall facility have been identified and their design and costs related to various parameters such as waste type, waste throughput, and the number of operational shifts employed. The model may be used to either estimate the cost of a particular waste packaging facility configuration or to explore the cost tradeoff between plant capital and labor. That is, one may use the model to search for the particular facility sizes and associated cost which when coupled with a particular number of shifts, and thus staffing level, leads to the lowest overall total cost. The functional components which the model considers include hot cells and their supporting facilities, transportation, cask handling facilities, transuranic waste handling facilities, and administrative facilities such as warehouses, security buildings, maintenance buildings, etc. The cost of each of these functional components is related either directly or indirectly to the various independent design parameters. Staffing by shift is reported into direct and indirect support labor. These staffing levels are in turn related to the waste type, waste throughput, etc. 2 refs., 11 figs., 3 tabs

Management of Low-Level Radioactive Waste from Research, Hospitals and Nuclear Medical Centers in Egypt - 13469

Energy Technology Data Exchange (ETDEWEB)

Hasan, M.A.; Selim, Y.T.; Lasheen, Y.F. [Hot Labs and Waste Management Center, Atomic Energy Authority, 3 Ahmed El-Zomor St., El-Zohour District, Naser City, 11787, Cairo (Egypt)

2013-07-01

The application of radioisotopes and radiation sources in medical diagnosis and therapy is an important issue. Physicians can use radioisotopes to diagnose and treat diseases. Methods of treatment, conditioning and management of low level radioactive wastes from the use of radiation sources and radioisotopes in hospitals and nuclear medicine application, are described. Solid Radioactive waste with low-level activity after accumulation, minimization, segregation and measurement, are burned or compressed in a compactor according to the international standards. Conditioned drums are transported to the interim storage site at the Egyptian Atomic Energy Authority (EAEA) represented in Hot Labs and Waste Management Center (HLWMC) for storage and monitoring. (authors)

MIS hot electron devices for enhancement of surface reactivity by hot electrons

DEFF Research Database (Denmark)

Thomsen, Lasse Bjørchmar

A Metal-Insulator-Semiconductor (MIS) based device is developed for investigation of hot electron enhanced chemistry. A model of the device is presented explaining the key concepts of the functionality and the character- istics. The MIS hot electron emitter is fabricated using cleanroom technology...... and the process sequence is described. An Ultra High Vacuum (UHV) setup is modified to facilitate experiments with electron emission from the MIS hot electron emitters and hot electron chemistry. Simulations show the importance of keeping tunnel barrier roughness to an absolute minimum. The tunnel oxide...... to be an important energy loss center for the electrons tunneling through the oxide lowering the emission e±ciency of a factor of 10 for a 1 nm Ti layer thickness. Electron emission is observed under ambient pressure conditions and in up to 2 bars of Ar. 2 bar Ar decrease the emission current by an order...

Electronic archive system for the management of historic radioactive waste

International Nuclear Information System (INIS)

Calin, M. R.; Garlea, C.; Petre, A. R.; Serbina, L.

2005-01-01

The development of nuclear activities in Magurele, Ilfov, during the last decades has led to the accumulation of an important quantity of radioactive waste. In addition to this, there is also a large number of former radioactive sources, now shut and removed from use, currently stored at IFIN-HH. This project deals with the discharge of historic waste storages belonging to the following nuclear units: - the WWR-S nuclear reactor - (the main reactor hall, the pump hall, the hot cells and annexes); - C.P.R. - used filters storage; - S.T.D.R. - storage for both historic radioactive waste and used filters; - shut sources in the storehouses in the 'Texas Bunker' building and annexes. For a modern management, including a proper system of quality insurance, an archiving system became needed. The electronic archive is based on several informational streams: the activity of storing historic radioactive waste; - the activity of locating historic radioactive waste; - the radiological descriptions of the storehouses and their influence areas; - the determination of the waste's composition. So as to reach these objectives, information regarding the following is necessary: the storehouse's inventory, the historic radioactive waste's characteristics and proprieties, the neighbors of this facility, the way in which the environment and the personnel involved in the operations are being influenced, the preparing of discharge operations, semblances. The data base conceived to tackle the problems of data related to nuclear waste management has been programmed in Microsoft Access (Microsoft Office). (authors)

Recent trend of administration on hot springs

Energy Technology Data Exchange (ETDEWEB)

Okubo, Shigeru [Environment Agency, Tokyo (Japan)

1989-01-01

The Environmental Agency exercises jurisdiction over Hot Spring Act, and plans to protect the source of the hot spring and to utilize it appropriately. From the aspect of utilization, hot springs are widely used as a means to remedy chronic diseases and tourist spots besides places for recuperation and repose. Statistics on Japanese hot springs showed that the number of hot spring spots and utilized-fountainhead increased in 1987, compared with the number in 1986. Considering the utilized-headspring, the number of naturally well-out springs has stabilized for 10 years while power-operated springs have increased. This is because the demand of hot springs has grown as the number of users has increased. Another reason is to keep the amount of hot water by setting up the power facility as the welled-out amount has decreased. Major point of recent administration on the hot spring is to permit excavation and utilization of hot springs. Designation of National hot spring health resorts started in 1954 in order to ensure the effective and original use of hot springs and to promote the public use of them, for the purpose of arranging the sound circumstances of hot springs. By 1988, 76 places were designated. 4 figs., 3 tabs.

Computational Prediction of Hot Spot Residues

Science.gov (United States)

Morrow, John Kenneth; Zhang, Shuxing

2013-01-01

Most biological processes involve multiple proteins interacting with each other. It has been recently discovered that certain residues in these protein-protein interactions, which are called hot spots, contribute more significantly to binding affinity than others. Hot spot residues have unique and diverse energetic properties that make them challenging yet important targets in the modulation of protein-protein complexes. Design of therapeutic agents that interact with hot spot residues has proven to be a valid methodology in disrupting unwanted protein-protein interactions. Using biological methods to determine which residues are hot spots can be costly and time consuming. Recent advances in computational approaches to predict hot spots have incorporated a myriad of features, and have shown increasing predictive successes. Here we review the state of knowledge around protein-protein interactions, hot spots, and give an overview of multiple in silico prediction techniques of hot spot residues. PMID:22316154

Integrated biomass gasification using the waste heat from hot slags: Control of syngas and polluting gas releases

International Nuclear Information System (INIS)

Sun, Yongqi; Seetharaman, Seshadri; Liu, Qianyi; Zhang, Zuotai; Liu, Lili; Wang, Xidong

2016-01-01

In this study, the thermodynamics of a novel strategy, i.e., biomass/CO 2 gasification integrated with heat recovery from hot slags in the steel industry, were systemically investigated. Both the target syngas yield and the polluting gas release were considered where the effect of gasifying conditions including temperature, pressure and CO 2 reacted was analyzed and then the roles of hot slags were further clarified. The results indicated that there existed an optimum temperature for the maximization of H 2 production. Compared to blast furnace slags, steel slags remarkably increased the CO yield at 600–1400 °C due to the existence of iron oxides and decreased the S-containing gas releases at 400–700 °C, indicating potential desulfurizing ability. The identification of biomass/CO 2 gasification thermodynamics in presence of slags could thus provide important clues not only for the deep understanding of biomass gasification but also for the industrial application of this emerging strategy from the viewpoint of syngas optimization and pollution control. - Highlights: • Biomass/CO 2 gasification was integrated with the heat recovery from hot slags. • Both syngas yield and polluting gas release during gasification were determined. • There existed an optimum temperature for the maximization of H 2 production. • Steel slags increased CO yield at 600–1400 °C due to the existence of iron oxides. • Steel slags remarkably decreased the releases of S-containing gas at 400–700 °C.

Hot Spots and Hot Moments of Nitrogen in a Riparian Corridor

Science.gov (United States)

Dwivedi, Dipankar; Arora, Bhavna; Steefel, Carl I.; Dafflon, Baptiste; Versteeg, Roelof

2018-01-01

We use 3-D high-resolution reactive transport modeling to investigate whether the spatial distribution of organic-carbon-rich and chemically reduced sediments located in the riparian zone and temporal variability in groundwater flow direction impact the formation and distribution of nitrogen hot spots (regions that exhibit higher reaction rates when compared to other locations nearby) and hot moments (times that exhibit high reaction rates as compared to longer intervening time periods) within the Rifle floodplain in Colorado. Groundwater flows primarily toward the Colorado River from the floodplain but changes direction at times of high river stage. The result is that oxic river water infiltrates the Rifle floodplain during these relatively short-term events. Simulation results indicate that episodic rainfall in the summer season leads to the formation of nitrogen hot moments associated with Colorado River rise and resulting river infiltration into the floodplain. The results further demonstrate that the naturally reduced zones (NRZs) present in sediments of the Rifle floodplain have a higher potential for nitrate removal, approximately 70% greater than non-NRZs for typical hydrological conditions. During river water infiltration, nitrate reduction capacity remains the same within the NRZs, however, these conditions impact non-NRZs to a greater extent (approximately 95% less nitrate removal). Model simulations indicate chemolithoautotrophs are primarily responsible for the removal of nitrate in the Rifle floodplain. These nitrogen hot spots and hot moments are sustained by microbial respiration and the chemolithoautotrophic oxidation of reduced minerals in the riparian zone.

Hot char-catalytic reforming of volatiles from MSW pyrolysis

International Nuclear Information System (INIS)

Wang, Na; Chen, Dezhen; Arena, Umberto; He, Pinjing

2017-01-01

Highlights: • Volatile from MSW pyrolysis is reformed with hot char from the same pyrolysis process. • The yields of syngas increase evidently with H 2 being the main contributor and the major component of the syngas. • Pyrolysis oil becomes light and its composition distribution is narrowed. • The HHV, volatile elements and alkali metals contents in the char decrease. • The emissions including SO 2 , NO, NO 2 and HCN changed after reforming process. - Abstract: Volatile products obtained from pyrolysis of municipal solid waste (MSW), including syngas and pyrolysis oil, were forced to contact the hot char from the same pyrolysis process at 500–600 °C in a fixed bed reactor to be reformed. The yields and properties of syngas, char and pyrolysis liquid were investigated; and the energy re-distribution among the products due to char reforming was quantified. The preliminary investigation at lab scale showed that hot char-catalytic reforming of the volatiles leads to an increase in the dry syngas yield from 0.25 to 0.37 N m 3 kg −1 MSW at 550 °C. Accordingly, the carbon conversion ratio into syngas increases from 29.6% to 35.0%; and the MSW chemical energy transferred into syngas increased from 41.8% to 47.4%. The yield of pyrolysis liquid products, including pyrolysis oil and water, decreased from 27.3 to 16.5 wt%, and the molecular weight of the oil becoming lighter. Approximately 60% of the water vapour contained in the volatiles converted into syngas. After reforming, the concentrations of SO 2 and HCN in the syngas decreases, while those of NO and NO 2 increase. The char concentrations of N, H, C and alkali metal species decreased and its higher heating value decreased too.

Economic assessment of greenhouse gas reduction through low-grade waste heat recovery using organic Rankine cycle (ORC)

Energy Technology Data Exchange (ETDEWEB)

Imran, Muhammad; Park, Byung Sik; Kim, Hyouck Ju; Usman, Muhammad [University of Science and Technology, Daejeon (Korea, Republic of); Lee, Dong Hyun [Korea Institute of Energy Research, Daejeon (Korea, Republic of)

2015-02-15

Low-grade waste heat recovery technologies reduce the environmental impact of fossil fuels and improve overall efficiency. This paper presents the economic assessment of greenhouse gas (GHG) reduction through waste heat recovery using organic Rankine cycle (ORC). The ORC engine is one of the mature low temperature heat engines. The low boiling temperature of organic working fluid enables ORC to recover low-temperature waste heat. The recovered waste heat is utilized to produce electricity and hot water. The GHG emissions for equivalent power and hot water from three fossil fuels-coal, natural gas, and diesel oil-are estimated using the fuel analysis approach and corresponding emission factors. The relative decrease in GHG emission is calculated using fossil fuels as the base case. The total cost of the ORC system is used to analyze the GHG reduction cost for each of the considered fossil fuels. A sensitivity analysis is also conducted to investigate the effect of the key parameter of the ORC system on the cost of GHG reduction. Throughout the 20-year life cycle of the ORC plant, the GHG reduction cost for R245fa is 0.02 $/kg to 0.04 $/kg and that for pentane is 0.04 $/kg to 0.05 $/kg. The working fluid, evaporation pressure, and pinch point temperature difference considerably affect the GHG emission.

Hot laboratory in Saclay. Equipment and radio-metallurgy technique of the hot lab in Saclay. Description of hot cell for handling of plutonium salts. Installation of an hot cell; Laboratoire a tres haute activite de Saclay. Equipement et techniques radiometallurgiques du laboratoire a haute activite de Saclay. Description de cellules pour manipulation de sels de plutonium. Amenagement d'une cellule du laboratoire de haute activite

Energy Technology Data Exchange (ETDEWEB)

Bazire, R; Blin, J; Cherel, G; Duvaux, Y; Cherel, G; Mustelier, J P; Bussy, P; Gondal, G; Bloch, J; Faugeras, P; Raggenbass, A; Raggenbass, P; Fufresne, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1959-07-01

Describes the conception and installation of the hot laboratory in Saclay (CEA, France). The construction ended in 1958. The main aim of this laboratory is to examine fuel rods of EL2 and EL3 as well as nuclear fuel studies. It is placed in between both reactors. In a first part, the functioning and specifications of the hot lab are given. The different hot cells are described with details of the ventilation and filtration system as well as the waste material and effluents disposal. The different safety measures are explained: description of the radiation protection, decontamination room and personnel monitoring. The remote handling equipment is composed of cutting and welding machine controlled with manipulators. Periscopes are used for sight control of the operation. In a second part, it describes the equipment of the hot lab. The unit for an accurate measurement of the density of irradiated uranium is equipped with an high precision balance and a thermostat. The equipment used for the working of irradiated uranium is described and the time length of each operation is given. There is also an installation for metallographic studies which is equipped with a manipulation bench for polishing and cleaning surfaces and a metallographic microscope. X-ray examination of uranium pellets will also be made and results will be compared with those of metallography. The last part describes the hot cells used for the manipulation of plutonium salts. The plutonium comes from the reprocessing plant and arrived as a nitric solution. Thus these cells are used to study the preparation of plutonium fluorides from nitric solution. The successive operations needed are explained: filtration, decontamination and extraction with TBP, purification on ion exchangers and finally formation of the plutonium fluorides. Particular attention has been given to the description of the specifications of the different gloveboxes and remote handling equipment used in the different reaction steps and

Hot laboratory in Saclay. Equipment and radio-metallurgy technique of the hot lab in Saclay. Description of hot cell for handling of plutonium salts. Installation of an hot cell; Laboratoire a tres haute activite de Saclay. Equipement et techniques radiometallurgiques du laboratoire a haute activite de Saclay. Description de cellules pour manipulation de sels de plutonium. Amenagement d'une cellule du laboratoire de haute activite

Energy Technology Data Exchange (ETDEWEB)

Bazire, R.; Blin, J.; Cherel, G.; Duvaux, Y.; Cherel, G.; Mustelier, J.P.; Bussy, P.; Gondal, G.; Bloch, J.; Faugeras, P.; Raggenbass, A.; Raggenbass, P.; Fufresne, J. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1959-07-01

Describes the conception and installation of the hot laboratory in Saclay (CEA, France). The construction ended in 1958. The main aim of this laboratory is to examine fuel rods of EL2 and EL3 as well as nuclear fuel studies. It is placed in between both reactors. In a first part, the functioning and specifications of the hot lab are given. The different hot cells are described with details of the ventilation and filtration system as well as the waste material and effluents disposal. The different safety measures are explained: description of the radiation protection, decontamination room and personnel monitoring. The remote handling equipment is composed of cutting and welding machine controlled with manipulators. Periscopes are used for sight control of the operation. In a second part, it describes the equipment of the hot lab. The unit for an accurate measurement of the density of irradiated uranium is equipped with an high precision balance and a thermostat. The equipment used for the working of irradiated uranium is described and the time length of each operation is given. There is also an installation for metallographic studies which is equipped with a manipulation bench for polishing and cleaning surfaces and a metallographic microscope. X-ray examination of uranium pellets will also be made and results will be compared with those of metallography. The last part describes the hot cells used for the manipulation of plutonium salts. The plutonium comes from the reprocessing plant and arrived as a nitric solution. Thus these cells are used to study the preparation of plutonium fluorides from nitric solution. The successive operations needed are explained: filtration, decontamination and extraction with TBP, purification on ion exchangers and finally formation of the plutonium fluorides. Particular attention has been given to the description of the specifications of the different gloveboxes and remote handling equipment used in the different reaction steps and

Decommissioning of the Risoe hot cell facility

International Nuclear Information System (INIS)

Carlsen, H.

1992-02-01

Concise descriptions of actions taken in relation to the decommissioning of the hot cell facility at Risoe National Laboratory are presented. The removal of fissile material, of large contaminated equipment from the concrete cell line and a separate shielded storage facility, and the removal of large contaminated facilities such as out cell parts of a tube transport system between a concrete cell and a lead shielded steel box and a remotely operated Reichert Telatom microscope housed in a lead shielded glove box is described in addition to the initial mapping of radiation levels related to the decontamination of concrete cells. The dose commitment of 17.7 mSv was ascribed to 12 persons in the 2nd half of 1991. The work resulting in these doses was mainly handling of waste together with the frogman entrances in order to repair the in-cell crane and power manipulator. The overall time schedule for the project still appears to be applicable. (AB)

The Distinction of Hot Herbal Compress, Hot Compress, and Topical Diclofenac as Myofascial Pain Syndrome Treatment.

Science.gov (United States)

Boonruab, Jurairat; Nimpitakpong, Netraya; Damjuti, Watchara

2018-01-01

This randomized controlled trial aimed to investigate the distinctness after treatment among hot herbal compress, hot compress, and topical diclofenac. The registrants were equally divided into groups and received the different treatments including hot herbal compress, hot compress, and topical diclofenac group, which served as the control group. After treatment courses, Visual Analog Scale and 36-Item Short Form Health survey were, respectively, used to establish the level of pain intensity and quality of life. In addition, cervical range of motion and pressure pain threshold were also examined to identify the motional effects. All treatments showed significantly decreased level of pain intensity and increased cervical range of motion, while the intervention groups exhibited extraordinary capability compared with the topical diclofenac group in pressure pain threshold and quality of life. In summary, hot herbal compress holds promise to be an efficacious treatment parallel to hot compress and topical diclofenac.

Hot particles

International Nuclear Information System (INIS)

Merwin, S.E.; Moeller, M.P.

1989-01-01

Nuclear Regulatory Commission (NRC) licensees are required to assess the dose to skin from a hot particle contamination event at a depth of skin of7mg/cm 2 over an area of 1 cm 2 and compare the value to the current dose limit for the skin. Although the resulting number is interesting from a comparative standpoint and can be used to predict local skin reactions, comparison of the number to existing limits based on uniform exposures is inappropriate. Most incidents that can be classified as overexposures based on this interpretation of dose actually have no effect on the health of the worker. As a result, resources are expended to reduce the likelihood that an overexposure event will occur when they could be directed toward eliminating the cause of the problem or enhancing existing programs such as contamination control. Furthermore, from a risk standpoint, this practice is not ALARA because some workers receive whole body doses in order to minimize the occurrence of hot particle skin contaminations. In this paper the authors suggest an alternative approach to controlling hot particle exposures

Seismic modifications to the Hot and Suspect Repair area Argone National Laboratory - West

International Nuclear Information System (INIS)

Malik, L.E.; Harris, B.G.

1993-01-01

The ANL-W WIPP Waste Facility Enhancement Project required substantial remodeling and upgrades to the Hot Fuels Examination Facility (HFEF) building, including the Hot and Suspect Repair Area (HSRA). The HSRA is an enclosed single-storied area inside the HFEF. It is separated into several compartments, some of which are used for handling radioactive materials. The HSRA roof consists of 18 GA steel Robertson Q decking with 1.5 in. concrete topping, and is utilized for storage. Braced steel frames support the HSRA roof, except at the north side, where the steel beams are connected to the HFEF columns. The HSRA has hollow block masonry perimeter and interior walls. Seismic evaluations concluded that the HSRA did not have a competent seismic force resisting system. The structure was upgraded by decoupling it from the HFEF framing for N/S motions, modifying two existing braced frames, adding a new braced frame that can be removed temporarily during maintenance and strengthening the roof diaphragm by a unique modification consisting of special epoxy grout and steel plates installed over the existing concrete roof

Hot-Film and Hot-Wire Anemometry for a Boundary Layer Active Flow Control Test

Science.gov (United States)

Lenahan, Keven C.; Schatzman, David M.; Wilson, Jacob Samuel

2013-01-01

Unsteady active flow control (AFC) has been used experimentally for many years to minimize bluff-body drag. This technology could significantly improve performance of rotorcraft by cleaning up flow separation. It is important, then, that new actuator technologies be studied for application to future vehicles. A boundary layer wind tunnel was constructed with a 1ft-x-3ft test section and unsteady measurement instrumentation to study how AFC manipulates the boundary layer to overcome adverse pressure gradients and flow separation. This unsteady flow control research requires unsteady measurement methods. In order to measure the boundary layer characteristics, both hot-wire and hot-film Constant Temperature Anemometry is used. A hot-wire probe is mounted in the flow to measure velocity while a hot-film array lays on the test surface to measure skin friction. Hot-film sensors are connected to an anemometer, a Wheatstone bridge circuit with an output that corresponds to the dynamic flow response. From this output, the time varying flow field, turbulence, and flow reversal can be characterized. Tuning the anemometers requires a fan test on the hot-film sensors to adjust each output. This is a delicate process as several variables drastically affect the data, including control resistance, signal input, trim, and gain settings.

Recov'Heat: An estimation tool of urban waste heat recovery potential in sustainable cities

Science.gov (United States)

Goumba, Alain; Chiche, Samuel; Guo, Xiaofeng; Colombert, Morgane; Bonneau, Patricia

2017-02-01

Waste heat recovery is considered as an efficient way to increase carbon-free green energy utilization and to reduce greenhouse gas emission. Especially in urban area, several sources such as sewage water, industrial process, waste incinerator plants, etc., are still rarely explored. Their integration into a district heating system providing heating and/or domestic hot water could be beneficial for both energy companies and local governments. EFFICACITY, a French research institute focused on urban energy transition, has developed an estimation tool for different waste heat sources potentially explored in a sustainable city. This article presents the development method of such a decision making tool which, by giving both energetic and economic analysis, helps local communities and energy service companies to make preliminary studies in heat recovery projects.

All-inorganic perovskite nanocrystal assisted extraction of hot electrons and biexcitons from photoexcited CdTe quantum dots.

Science.gov (United States)

Mondal, Navendu; De, Apurba; Samanta, Anunay

2018-01-03

Excitation of semiconductor quantum dots (QDs) by photons possessing energy higher than the band-gap creates a hot electron-hole pair, which releases its excess energy as waste heat or under certain conditions (when hν > 2E g ) produces multiple excitons. Extraction of these hot carriers and multiple excitons is one of the key strategies for enhancing the efficiency of QD-based photovoltaic devices. However, this is a difficult task as competing carrier cooling and relaxation of multiple excitons (through Auger recombination) are ultrafast processes. Herein, we study the potential of all-inorganic perovskite nanocrystals (NCs) of CsPbX 3 (X = Cl, Br) as harvesters of these short-lived species from photo-excited CdTe QDs. The femtosecond transient absorption measurements show CsPbX 3 mediated extraction of both hot and thermalized electrons of the QDs (under a low pump power) and (under a high pump fluence) extraction of multiple excitons prior to their Auger assisted recombination. A faster timescale of thermalized electron transfer (∼2 ps) and a higher extraction efficiency of hot electrons (∼60%) are observed in the presence of CsPbBr 3 . These observations demonstrate the potential of all-inorganic perovskite NCs in the extraction of these short-lived energy rich species implying that complexes of the QDs and perovskite NCs are better suited for improving the efficiency of QD-sensitized solar cells.

SHOSPA-MOD, Hot Spot Factors for Fuel and Clad, Hot Channel Factors

International Nuclear Information System (INIS)

Amendola, A.

1982-01-01

1 - Nature of the physical problem solved: SHOSPA evaluates the hot spot factors for fuel and cladding as well as the hot channel factor as a function of the confidence level. Moreover, it evaluates the probability on n hot subassemblies. The code has been developed with emphasis on sodium cooled fast reactors, but it is applicable to any type of reactors constituted of bundled fuel rods with single phase coolant. An option for plotting is available in this version. 2 - Restrictions on the complexity of the problem: This code is applicable to any type of reactors constituted of fuel rods with single phase coolant

Hot test of a TALSPEAK procedure for separation of actinides and lanthanides using recirculating DTPA-lactic acid solution

International Nuclear Information System (INIS)

Persson, G.; Svantesson, I.; Wingefors, S.; Liljenzin, J.O.

1984-01-01

Results are reported from a hot test of a TALSPEAK type process for separation of higher actinides (Am, Cm) from lanthanides. Actinides and lanthanides are extracted by 1 M HDEHP and separated by selective strip of the actinides, using a mixture of DTPA and lactic acid (reversed TALSPEAK process). In order to minimize the generation of secondary waste, a procedure using recirculating DTPA-Lactic acid solution has been developed. A separation factor between Am and Eu of 132 was achieved. In regard to separations of Am and Cm from commercial HLLW (high level liquid wastes), the factor corresponds to 1.5% of the lanthanide group remaining with the actinides. The loss of Am was about 0.2%. 9 figures, 3 tables

Solidification of high level liquid waste (HLLW) into ceramics by sintering process

International Nuclear Information System (INIS)

Masuda, Sumio; Oguino, Naohiko; Tsunoda, Naomi; O-oka, Kazuo; Ohta, Takao.

1979-01-01

One of the alternatives to vitrified solid which is acceptable and well characterized for storing radioactive HLLW with desirable long-term stability is ceramics. On the other hand, the solidification process of highly radioactive wastes should be simple and suitable for continuous production. On the above described basis, the authors have made preliminary study on the production of sintered ceramics by the addition of several oxides to HLLW. The simulated waste and additive oxides were pressed in a mold to make the preforms of 50 mm diameter and 10 to 15 mm thick. The preforms were then normally sintered at temperature from 1000 to 1400 deg C for 2 to 4 hours. The characterization of the sintered solids revealed the following facts. (1) X-ray diffraction analysis showed that the expected crystals were formed by normal-sintering as well as by hot-pressing. (2) The bulk density of the ceramics by normal-sintering was around 90 to 95% of the assumed theoretical values. (3) The leach-rate of the solids was affected by the bulk density. (4) Other properties of the solids, such as thermal expansion or thermal conductivity, are dominantly determined by those of main crystals in the solids. Sintering process is generally simple and productive as far as normal sintering is concerned. However, hot-pressing is an intermittent and time consuming process. From this fact, the authors intended to adopt the normal sintering process for the ceramic solidification of high level liquid wastes. (Wakatsuki, Y.)

Pectic oligosacharides from lemon peel wastes: production, purification, and chemical characterization.

Science.gov (United States)

Gómez, Belén; Gullón, Beatriz; Yáñez, Remedios; Parajó, Juan C; Alonso, Jose L

2013-10-23

Lemon peel wastes were extracted with water to remove free sugars and other soluble compounds, and the insoluble solid was employed as a substrate for the manufacture of pectin-derived oligosaccharides by processing with hot, compressed water. When water-extracted lemon peel wastes were treated with water at 160 °C, the oligomer concentration reached the maximum value (31 g/L). Autohydrolysis liquors were subjected to two membrane filtration stages (diafiltration followed by concentration), yielding a refined product containing about 98 wt % of oligomers at a global yield of 14 kg/100 kg oven-dry lemon peel. The concentrate contained oligogalacturonides (with DP in the range of 2-18) and arabinooligosaccharides (with DP in the range of 2-8).

Application of Cyclone to Removal of Hot Particulate in Hot Cell

International Nuclear Information System (INIS)

Kim, Gye Nam; Lee, Sung Yeol; Won, Hui Jun; Jung, Chong Hun; Oh, Won Zin

2005-01-01

The size and main ingredient of hot particulate generated during the nuclide experiment in hot cells of nuclear facilities were 0.5300 μm and UO 2 . A cyclone filter equipment which consists of a cyclone and Bag/HEPA filter was devised to remove hot particulate generated during the nuclide experiment in hot cells of nuclear facilities. The experimental conditions to maximize the collection efficiency of hot particulate were suggested through experiments done with the cyclone filter equipment. With the large size of simulated particulate, the collection efficiency of the particulate was high. When the size of simulated particulate was more than 5 μm, the collection efficiency of the particulate was more than 80% and when the size of simulated particulate was less than 1.0 μm, the collection efficiency decreased by less than. If the inflow velocity of simulated particulate was increased, the collection efficiency of the particulate was also increased. When the inflow velocity of simulated particulate was more than 12 m/sec, the collection efficiency was higher than , but after 17 m/sec inflow velocity, no change observed. The collection efficiency of the simulated particulate can be enhanced with the length of vortex finder inside the chamber. With the length of vortex finder, 7.2 cm, the observed collection efficiency of the particulate was the maximum. Moreover, when the sub-cone was attached under the cyclone, the collection efficiency of cyclone increased 2%. It was found that effect by attachment of sub-cone was not serious.

Ionospheric hot spot at high latitudes

International Nuclear Information System (INIS)

Schunk, R.W.; Sojka, J.J.

1982-01-01

A hot spot (or spots) can occur in the high-latitude ionosphere depending on the plasma convection pattern. The hot spot corresponds to a small magnetic local time-magnetic latitude region of elevated ion temperatures located near the dusk and/or dawn meridians. For asymmetric convection electric field patterns, with enhanced flow in either the dusk or dawn sector of the polar cap, a single hot spot should occur in association with the strong convection cell. However, on geomagnetically disturbed days, two strong convection cells can occur, and hence, two hot spots should exist. The hot spot should be detectable when the electric field in the strong convection cell exceeds about 40 mV m -1 . For electric fields of the order of 100 mV m -1 in the convection cell, the ion temperature in the hot spot is greatest at low altitudes, reaching 4000 0 K at 160 km, and decreases with altitude in the F-region. An ionospheric hot spot (or spots) can be expected at all seasons and for a wide range of solar cycle conditions

Method and device for incinerating radioactive wastes and preparing burnable wastes for non-polluting storage

International Nuclear Information System (INIS)

Hempelmann, W.

1975-01-01

An apparatus for incinerating radioactive wastes includes a furnace which has air inlet conduits and a flue gas outlet conduit and air heaters as well as blowers connected to the air inlets for forcing hot air into the furnace. The apparatus further has a feeding device connected to the charging end of the furnace for introducing liquid or solid wastes thereinto and a device which communicates with the discharge end of the furnace for removing solid reaction products from the furnace. In the flue gas conduit there is connected a plurality of flue gas filters each containing filter candles, a flue gas chamber and a mechanism for removing ashes from the flue gas chamber. The apparatus also includes a mixer section connected with the outlet of each flue gas filter and having a mechanism for mixing cool air with the flue gas filtered by the flue gas filters. Gas blowers connected to the output of the mixer section draw the gas from the apparatus. 18 Claims, 8 Drawing Figures

Application of solar energy to the supply of industrial process hot water. Aerotherm final report, 77-235. [Can washing in Campbell Soup plant

Energy Technology Data Exchange (ETDEWEB)

None

1977-01-01

The objectives of the Solar Industrial Process Hot Water Program are to design, test, and evaluate the application of solar energy to the generation and supply of industrial process hot water, and to provide an assessment of the economic and resource benefits to be gained. Other objectives are to stimulate and give impetus to the use of solar energy for supplying significant amounts of industrial process heat requirements. The plant selected for the design of a solar industrial process hot water system was the Campbell Soup facility in Sacramento, California. The total hot water demand for this plant varies between 500 and 800 gpm during regular production shifts, and hits a peak of over 1,000 gpm for approximately one hour during the cleanup shift. Most of the hot water is heated in the boiler room by a combination of waste heat recovery and low pressure (5 psi) steam-water heat exchangers. The hot water emerges from the boiler room at a temperature between 160/sup 0/F and 180/sup 0/F and is transported to the various process areas. Booster heaters in the process areas then use low pressure (5 psi) or medium pressure (20 psi) steam to raise the temperature of the water to the level required for each process. Hot water is used in several processes at the Campbell Soup plant, but the can washing process was selected to demonstrate the feasibility of a solar hot water system. A detailed design and economic analysis of the system is given. (WHK)

Actinide partitioning from high level liquid waste using the Diamex process

International Nuclear Information System (INIS)

Madic, C.; Blanc, P.; Condamines, N.; Baron, P.; Berthon, L.; Nicol, C.; Pozo, C.; Lecomte, M.; Philippe, M.; Masson, M.; Hequet, C.

1994-01-01

The removal of long-lived radionuclides, which belong to the so-called minor actinides elements, neptunium, americium and curium, from the high level nuclear wastes separated during the reprocessing of the irradiated nuclear fuels in order to transmute them into short-lived nuclides, can substantially decrease the potential hazards associated with the management of these nuclear wastes. In order to separate minor actinides from high-level liquid wastes (HLLW), a liquid-liquid extraction process was considered, based on the use of diamide molecules, which display the property of being totally burnable, thus they do not generate secondary solid wastes. The main extracting properties of dimethyldibutyltetradecylmalonamide (DMDBTDMA), the diamide selected for the development of the DIAMEX process, are briefly described in this paper. Hot tests of the DIAMEX process (using DMDBTDMA) related to the treatment of an mixed oxide fuels (MOX) type HLLW, were successfully performed. The minor actinide decontamination factors of the HLLW obtained were encouraging. The main results of these tests are presented and discussed in this paper. (authors). 9 refs., 2 figs., 7 tabs

Determination of molybdenum, ruthenium, rhodium, and palladium in radioinactive simulated waste of the nuclear fuel cycle by solid sampling graphite furnace atomic absorption spectrometry (GFAAS)

International Nuclear Information System (INIS)

Schmiedel, G.; Mainka, E.; Ache, H.J.

1989-01-01

In relation with insoluble particles in the nuclear fuel cycle waste, the solid sampling GFAAS was used to determine molybdenum, ruthenium, rhodium, and palladium in such waste. Two methods for the direct determination of these elements are described. The samples must be handled in glove boxes or moreover in hot cells with a robot. The determination of the elements by the cup-in-tube technique needs a very sensitive balance (microbalance) for weighing in μg-range and the handling of this method is not practical in glove boxes and hot cells. An alternative technique of solid sampling GFAAS, which can be used without great problems in glove boxes and hot cells is the slurry technique. In this case two methods have been used. One method uses graphite powder as a diluter, the other is the direct suspension of the sample in a matrix modifier solution. In the case of slurry technique with predilution of the sample with graphite powder, recoveries between 91 and 102% and RSD between 4 and 8% were obtained, whereas in the case of slurry technique with direct suspension of the waste sample recoveries between 91 and 103% and RSD between 14 and 20% for the above mentioned elements were obtained. (orig.)

Menopausal Hot Flashes and White Matter Hyperintensities

Science.gov (United States)

Thurston, Rebecca C.; Aizenstein, Howard J.; Derby, Carol A.; Sejdić, Ervin; Maki, Pauline M.

2015-01-01

Objective Hot flashes are the classic menopausal symptom. Emerging data links hot flashes to cardiovascular disease (CVD) risk, yet how hot flashes are related to brain health is poorly understood. We examined the relationship between hot flashes - measured via physiologic monitor and self-report - and white matter hyperintensities (WMH) among midlife women. Methods Twenty midlife women ages 40-60 without clinical CVD, with their uterus and both ovaries, and not taking hormone therapy were recruited. Women underwent 24 hours of ambulatory physiologic and diary hot flash monitoring to quantify hot flashes; magnetic resonance imaging to assess WMH burden; 72 hours of actigraphy and questionnaires to quantify sleep; and a blood draw, questionnaires, and physical measures to quantify demographics and CVD risk factors. Test of a priori hypotheses regarding relations between physiologically-monitored and self-reported wake and sleep hot flashes and WMH were conducted in linear regression models. Results More physiologically-monitored hot flashes during sleep were associated with greater WMH, controlling for age, race, and body mass index [beta(standard error)=.0002 (.0001), p=.03]. Findings persisted controlling for sleep characteristics and additional CVD risk factors. No relations were observed for self-reported hot flashes. Conclusions More physiologically-monitored hot flashes during sleep were associated with greater WMH burden among midlife women free of clinical CVD. Results suggest that relations between hot flashes and CVD risk observed in the periphery may extend to the brain. Future work should consider the unique role of sleep hot flashes in brain health. PMID:26057822

Ferrocyanide tank safety program: Cesium uptake capacity of simulated ferrocyanide tank waste. Final report

International Nuclear Information System (INIS)

Burgeson, I.E.; Bryan, S.A.

1995-07-01

The objective of this project is to determine the capacity for 137 Cs uptake by mixed metal ferrocyanides present in Hanford Site waste tanks, and to assess the potential for aggregation of these 137 Cs-exchanged materials to form ''hot-spots'' in the tanks. This research, performed at Pacific Northwest Laboratory (PNL) for Westinghouse Hanford Company, stems from concerns regarding possible localized radiolytic heating within the tanks. After ferrocyanide was added to 18 high-level waste tanks in the 1950s, some of the ferrocyanide tanks received considerable quantities of saltcake waste that was rich in 137 Cs. If radioactive cesium was exchanged and concentrated by the nickel ferrocyanide present in the tanks, the associated heating could cause tank temperatures to rise above the safety limits specified for the ferrocyanide-containing tanks, especially if the supernate in the tanks is pumped out and the waste becomes drier

Single Shell Tank Waste Characterization Project for Tank B-110, Core 9 - data package and PNL validation summary report

International Nuclear Information System (INIS)

Pool, K.N.; Jones, T.E.; McKinley, S.G.; Tingey, J.M.; Longaker, T.M.; Gibson, J.A.

1990-01-01

This Data Package contains results obtained by Pacific Northwest Laboratory (PNL) staff in the characterization and analyses of Core 9 segments taken from the Single-Shell Tank (SST) 110B. The characterization and analysis of Core 9 segments are outlined in the Waste Characterization Plan for Hanford Site Single-Shell Tanks and in the Pacific Northwest Laboratory (PNL) Single-Shell Tank Waste Characterization Support FY 89/90 Statement of Work (SOW), Rev. 1 dated March, 1990. Specific analyses for each sub-sample taken from a segment are delineated in Test Instructions prepared by the PNL Single-Shell Tank Waste Characterization Project Management Office (SST Project) in accordance with procedures contained in the SST Waste Characterization Procedure Compendium (PNL-MA-599). Analytical procedures used in the characterization activities are also included in PNL-MA-599. Core 9 included five segments although segment 1 did not have sufficient material for characterization. The five samplers were received from Westinghouse Hanford Company (WHC) on 11/21-22/89. Each segment was contained in a sampler and was enclosed in a shipping cask. The shipping cask was butted up to the 325-A hot cell and the sampler moved into the hot cell. The material in the sampler (i.e., the segment) was extruded from the sampler, limited physical characteristics assessed, and photographed. At this point samples were taken for particle size and volatile organic analyses. Each segment was then homogenized. Sub-samples were taken for required analyses as delineated in the appropriate Test Instruction. Table 1 includes sample numbers assigned to Core 9 segment materials being transferred from 325-A Hot Cell. Sample numbers 90-0298, 90-0299, 90-0302, and 90-0303 were included in Table 1 although no analyses were requested for these samples. Table 2 lists Core 9 sub-sample numbers per sample preparation method

Long-term high-level waste technology. Composite quarterly technical report: April-June 1981

International Nuclear Information System (INIS)

Cornman, W.R.

1981-12-01

This series of reports summarizes research and development studies on the immobilization of high-level wastes from the chemical reprocessing of nuclear reactor fuels. The reports are grouped under the following tasks: (1) program management and support; (2) waste preparation; (3) waste fixation; and (4) final handling. Some of the highlights are: leaching properties were obtained for titanate and tailored ceramic materials being developed at ICPP to immobilize zirconia calcine; comparative leach tests, hot-cell tests, and process evaluations were conducted of waste form alternatives to borosilicate glass for the immobilization of SRP high-level wastes, experiments were run at ANL to qualify neutron activation analysis and radioactive tracers for measuring leach rates from simulated waste glasses; comparative leach test samples of SYNROC D were prepared, characterized, and tested at LLNL; encapsulation of glass marbles with lead or lead alloys was demonstrated on an engineering scale at PNL; a canister for reference Commercial HLW was designed at PNL; a study of the optimization of salt-crete was completed at SRL; a risk assessment showed that an investment for tornado dampers in the interim storage building of the DWPF is unjustified

Use of waste heat from a dairy for heating of a community house

Energy Technology Data Exchange (ETDEWEB)

Rehn, C

1976-01-01

In a dairy, a lot of cooling capacity is needed. This article describes how this waste heat can be used for heating a community house including a sport establishment and producing hot water for that house. Four different technical solutions are discussed; (1) floor heat, (2) heat transfer connected to the ventilation, (3) regenerative heat exchanger, and (4) use of heat pumps.

Hot tearing studies in AA5182

Science.gov (United States)

van Haaften, W. M.; Kool, W. H.; Katgerman, L.

2002-10-01

One of the major problems during direct chill (DC) casting is hot tearing. These tears initiate during solidification of the alloy and may run through the entire ingot. To study the hot tearing mechanism, tensile tests were carried out in semisolid state and at low strain rates, and crack propagation was studied in situ by scanning electron microscopy (SEM). These experimentally induced cracks were compared with hot tears developed in an AA5182 ingot during a casting trial in an industrial research facility. Similarities in the microstructure of the tensile test specimens and the hot tears indicate that hot tearing can be simulated by performing tensile tests at semisolid temperatures. The experimental data were compared with existing hot tearing models and it was concluded that the latter are restricted to relatively high liquid fractions because they do not take into account the existence of solid bridges in the crack.

Applicability of Related Data, Algorithms, and Models to the Simulation of Ground-Coupled Residential Hot Water Piping in California

International Nuclear Information System (INIS)

Warner, J.L.; Lutz, J.D.

2006-01-01

Residential water heating is an important consideration in California?s building energy efficiency standard. Explicit treatment of ground-coupled hot water piping is one of several planned improvements to the standard. The properties of water, piping, insulation, backfill materials, concrete slabs, and soil, their interactions, and their variations with temperature and over time are important considerations in the required supporting analysis. Heat transfer algorithms and models devised for generalized, hot water distribution system, ground-source heat pump and ground heat exchanger, nuclear waste repository, buried oil pipeline, and underground electricity transmission cable applications can be adapted to the simulation of under-slab water piping. A numerical model that permits detailed examination of and broad variations in many inputs while employing a technique to conserve computer run time is recommended.

Do scientists trace hot topics?

Science.gov (United States)

Wei, Tian; Li, Menghui; Wu, Chensheng; Yan, Xiao-Yong; Fan, Ying; Di, Zengru; Wu, Jinshan

2013-01-01

Do scientists follow hot topics in their scientific investigations? In this paper, by performing analysis to papers published in the American Physical Society (APS) Physical Review journals, it is found that papers are more likely to be attracted by hot fields, where the hotness of a field is measured by the number of papers belonging to the field. This indicates that scientists generally do follow hot topics. However, there are qualitative differences among scientists from various countries, among research works regarding different number of authors, different number of affiliations and different number of references. These observations could be valuable for policy makers when deciding research funding and also for individual researchers when searching for scientific projects.

Preliminary analysis of treatment strategies for transuranic wastes from reprocessing plants

International Nuclear Information System (INIS)

Ross, W.A.; Schneider, K.J.; Swanson, J.L.; Yasutake, K.M.; Allen, R.P.

1985-07-01

This document provides a comparison of six treatment options for transuranic wastes (TRUW) resulting from the reprocessing of commercial spent fuel. Projected transuranic waste streams from the Barnwell Nuclear Fuel Plant (BNFP), the reference fuel reprocessing plant in this report, were grouped into the five categories of hulls and hardware, failed equipment, filters, fluorinator solids, and general process trash (GPT) and sample and analytical cell (SAC) wastes. Six potential treatment options were selected for the five categories of waste. These options represent six basic treatment objectives: (1) no treatment, (2) minimum treatment (compaction), (3) minimum number of processes and products (cementing or grouting), (4) maximum volume reduction without decontamination (melting, incinerating, hot pressing), (5) maximum volume reduction with decontamination (decontamination, treatment of residues), and (6) noncombustible waste forms (melting, incinerating, cementing). Schemes for treatment of each waste type were selected and developed for each treatment option and each type of waste. From these schemes, transuranic waste volumes were found to vary from 1 m 3 /MTU for no treatment to as low as 0.02 m 3 /MTU. Based on conceptual design requirements, life-cycle costs were estimated for treatment plus on-site storage, transportation, and disposal of both high-level and transuranic wastes (and incremental low-level wastes) from 70,000 MTU. The study concludes that extensive treatment is warranted from both cost and waste form characteristics considerations, and that the characteristics of most of the processing systems used are acceptable. The study recommends that additional combinations of treatment methods or strategies be evaluated and that in the interim, melting, incineration, and cementing be further developed for commercial TRUW. 45 refs., 9 figs., 32 tabs

Preliminary analysis of treatment strategies for transuranic wastes from reprocessing plants

Energy Technology Data Exchange (ETDEWEB)

Ross, W.A.; Schneider, K.J.; Swanson, J.L.; Yasutake, K.M.; Allen, R.P.

1985-07-01

This document provides a comparison of six treatment options for transuranic wastes (TRUW) resulting from the reprocessing of commercial spent fuel. Projected transuranic waste streams from the Barnwell Nuclear Fuel Plant (BNFP), the reference fuel reprocessing plant in this report, were grouped into the five categories of hulls and hardware, failed equipment, filters, fluorinator solids, and general process trash (GPT) and sample and analytical cell (SAC) wastes. Six potential treatment options were selected for the five categories of waste. These options represent six basic treatment objectives: (1) no treatment, (2) minimum treatment (compaction), (3) minimum number of processes and products (cementing or grouting), (4) maximum volume reduction without decontamination (melting, incinerating, hot pressing), (5) maximum volume reduction with decontamination (decontamination, treatment of residues), and (6) noncombustible waste forms (melting, incinerating, cementing). Schemes for treatment of each waste type were selected and developed for each treatment option and each type of waste. From these schemes, transuranic waste volumes were found to vary from 1 m/sup 3//MTU for no treatment to as low as 0.02 m/sup 3//MTU. Based on conceptual design requirements, life-cycle costs were estimated for treatment plus on-site storage, transportation, and disposal of both high-level and transuranic wastes (and incremental low-level wastes) from 70,000 MTU. The study concludes that extensive treatment is warranted from both cost and waste form characteristics considerations, and that the characteristics of most of the processing systems used are acceptable. The study recommends that additional combinations of treatment methods or strategies be evaluated and that in the interim, melting, incineration, and cementing be further developed for commercial TRUW. 45 refs., 9 figs., 32 tabs.

'Hot' cognition in major depressive disorder

DEFF Research Database (Denmark)

Miskowiak, Kamilla W; Carvalho, Andre F

2014-01-01

Major depressive disorder (MDD) is associated with significant cognitive dysfunction in both 'hot' (i.e. emotion-laden) and 'cold' (non-emotional) domains. Here we review evidence pertaining to 'hot' cognitive changes in MDD. This systematic review searched the PubMed and PsycInfo computerized......-limbic network with hyper-activity in limbic and ventral prefrontal regions paired with hypo-activity of dorsal prefrontal regions subserve these abnormalities. A cross-talk of 'hot' and 'cold' cognition disturbances in MDD occurs. Disturbances in 'hot cognition' may also contribute to the perpetuation......' cognition deficits in healthy relatives of patients with MDD. Taken together, these findings suggest that abnormalities in 'hot' cognition may constitute a candidate neurocognitive endophenotype for depression....

Titanate ceramics for immobilisation of uranium-rich radioactive wastes arising from {sup 99}Mo production

Energy Technology Data Exchange (ETDEWEB)

Carter, M.L.; Li, H. [Institute of Materials Engineering, Australian Nuclear Science and Technology Organisation, PMB 1, Menai, Sydney, NSW 2232 (Australia); Zhang, Y. [Institute of Materials Engineering, Australian Nuclear Science and Technology Organisation, PMB 1, Menai, Sydney, NSW 2232 (Australia)], E-mail: yzx@ansto.gov.au; Vance, E.R.; Mitchell, D.R.G. [Institute of Materials Engineering, Australian Nuclear Science and Technology Organisation, PMB 1, Menai, Sydney, NSW 2232 (Australia)

2009-02-28

Uranium-rich liquid wastes arising from UO{sub 2} targets which have been neutron-irradiated to generate medical radioisotopes such as {sup 99m}Tc require immobilisation. A pyrochlore-rich hot isostatically pressed titanate ceramic can accommodate at least 40 wt% of such waste expressed on an oxide basis. In this paper, the baseline waste form composition (containing 40 wt% UO{sub 2}) was adjusted in two ways: (a) varying the UO{sub 2} loading with constant precursor oxide materials, (b) varying the precursor composition with constant waste loading of UO{sub 2}. This resulted in the samples having a similar phase assemblage but the amounts of each phase varied. The oxidation states of U in selected samples were determined using diffuse reflection spectroscopy (DRS) and electron energy loss spectroscopy (EELS). Leaching studies showed that there was no significant difference in the normalised elemental release rates and the normalised release rates are comparable with those from synroc-C. This demonstrates that waste forms based on titanate ceramics are robust and flexible for the immobilisation of U-rich waste streams from radioisotope processing.

Impacts of transportation on a test and evaluation facility for nuclear waste disposal: a systems analysis

International Nuclear Information System (INIS)

Varadarajan, R.V.; Peterson, R.W.; Joy, D.S.; Gibson, S.M.

1983-01-01

An essential element of the Test and Evaluation Facility (TEF) is a waste packaging facility capable of producing a small number Test and Evaluation Facility of packages consisting of several different waste forms. The study envisions three scenarios for such a packaging facility: (1) modify an existing hot cell facility such as the Engine Maintenance Assembly and Disassembly (EMAD) facility at the Nevada Test Site so that it can serve as a packaging facility for the TEF. This scenario is referred to as the EMAD Option. (2) Build a new generic packaging facility (GPF) at the site of the TEF. In other words, colocate the GPF and the TEF. This scenario is referred to as the GPF Option, and (3) utilize the EMAD facility in conjunction with a colocated GPF (of minimal size and scope) at the TEF. This scenario is referred to as the Split Option. The results of the system study clearly bring out the fact that transportation has a significant impact on the selection and siting of the waste packaging facility. Preliminary conclusions, subject to the assumptions of the study, include the following: (1) regardless of the waste form, the GPF option is preferable to the other two in minimizing both transportation costs and logistical problems, (2) for any given scenario and choice of waste forms, there exists a candidate TEF location for which the transportation costs are at a minimum compared to the other locations, (3) in spite of the increased transportation costs and logistical complexity, the study shows that the overall system costs favor modification of an existing hot cell facility for the particular case considered

A review on hot tearing of magnesium alloys

Directory of Open Access Journals (Sweden)

Jiangfeng Song

2016-09-01

Full Text Available Hot tearing is often a major casting defect in magnesium alloys and has a significant impact on the quality of their casting products. Hot tearing of magnesium alloys is a complex solidification phenomenon which is still not fully understood, it is of great importance to investigate the hot tearing behaviour of magnesium alloys. This review attempts to summarize the investigations on hot tearing of magnesium alloys over the past decades. The hot tearing criteria including recently developed Kou's criterion are summarized and compared. The numeric simulation and assessing methods of hot tearing, factors influencing hot tearing, and hot tearing susceptibility (HTS of magnesium alloys are discussed.

Ninth international symposium on hot atom chemistry. Abstracts

International Nuclear Information System (INIS)

1977-01-01

Abstracts of the papers presented at the Symposium are compiled. The topics considered were chemical dynamics of high energy reactions, hot atom chemistry in organic compounds of tritium, nitrogen, oxygen, and halogens, theory and chemical dynamics of hot atom reactions as determined by beam studies, solid state reactions of recoil atoms and implanted ions, hot atom chemistry in energy-related research, hot atom chemistry in inorganic compounds of oxygen and tritium, hot positronium chemistry, applied hot atom chemistry in labelling, chemical effects of radioactive decay, decay-induced reactions and excitation labelling, physical methods in hot atom chemistry, and hot atom reactions in radiation and stratospheric chemistry

Processing glass-pyrochlore composites for nuclear waste encapsulation

International Nuclear Information System (INIS)

Pace, S.; Cannillo, V.; Wu, J.; Boccaccini, D.N.; Seglem, S.; Boccaccini, A.R.

2005-01-01

Glass matrix composites have been developed as alternative materials to immobilize nuclear solid waste, in particular actinides. These composites are made of soda borosilicate glass matrix, into which particles of lanthanum zirconate pyrochlore are encapsulated in concentrations of 30 vol.%. The fabrication process involves powder mixing followed by hot-pressing. At the relatively low processing temperature used (620 deg. C), the pyrochlore crystalline structure of the zirconate, which is relevant for containment of radioactive nuclei, remains unaltered. The microstructure of the composites exhibits a homogeneous distribution of isolated pyrochlore particles in the glass matrix and strong bonding at the matrix-particle interfaces. Hot-pressing was found to lead to high densification (95% th.d.) of the composite. The materials are characterized by relatively high elastic modulus, flexural strength, hardness and fracture toughness. A numerical approach using a microstructure-based finite element solver was used in order to investigate the mechanical properties of the composites

Development of the Thyssen hot briquetting process and its industrial application

Energy Technology Data Exchange (ETDEWEB)

Hoeffken, E.; Pflipsen, H.D.; Seidelmann, L.; Auth, R.

1988-07-18

At Thyssen Stahl AG the converter waste gases of the basic oxygen steelmaking plant Bruckhausen are dedusted according to a new method , the CO included in the waste gas being recovered and used as fuel gas. The dry dusts produced in this process are dumped despite their high iron content. Considerations aimed at recovering this residues have led to the development of a hot briquetting process for converter dusts. Preliminary tests were followedby the construction of a pilot plant with a briqueting capacity of 2 to 3 t/h. As a result of the successful trials, a large-size plant for a throughput rate of 12 t/h was commissioned at mid-1987, which allows to briquet all the dusts produced in the steel plant. About 30000 t briquettes, which are charged as cooling ore and scrap in the steel plant, have been produced so far. In view of the high iron content of 60 to 70% and of the elimination of dumping costs, the process can be operated economically. The recycling of the dust has contributed, moreover, to the relief of the dumping ground and to environmental protection.

History of waste tank 11, 1955 through 1974

International Nuclear Information System (INIS)

Davis, T.L.; Tharin, D.W.; Lohr, D.R.

1978-10-01

Tank 11 was placed in service as a receiver of low heat waste (LW) in July 1955. In November 1961, the supernate was decanted from the sludge to prepare tank 11 for receipt of frame waste. In July 1962, the supernate was again decanted and tank 11 was used to receive fresh high heat waste (HW) from the enriched uranium process in Building 221-H. Again, the supernate was decanted in June 1965 and July 1967 to allow the tank to be reused for waste receipt. In order to use tank 11 for solid salt storage, a sludge removal operation was conducted in October 1969. The operation was unsuccessful. Tank 11 consequently received hot concentrated supernate from tank 10 blended with dilute waste and was later returned to service as a receiver of HW. A small, apparently inactive leak site was found in April 1974. Inspections have been made of the annulus and the interior of the tank both visually and with an optical periscope. Samples of the sludge, supernate, and tank vapor have been analyzed. Top-to-bottom profiles of radiation and temperature in the tank have been taken and measurements were made of deflections in the bottom knuckle plate due to changing liquid level. One horizontal and seventeen vertical cooling coils have failed, all within one month following the sludge removal operation. Several equipment modifications and various equipment repairs were made. 14 figures, 3 tables

Comparison of the corrosion behaviors of the glass-bonded sodalite ceramic waste form and reference HLW glasses

International Nuclear Information System (INIS)

Ebert, W. L.; Lewis, M. A.

1999-01-01

A glass-bonded sodalite ceramic waste form is being developed for the long-term immobilization of salt wastes that are generated during spent nuclear fuel conditioning activities. A durable waste form is prepared by hot isostatic pressing (HIP) a mixture of salt-loaded zeolite powders and glass frit. A mechanistic description of the corrosion processes is being developed to support qualification of the CWF for disposal. The initial set of characterization tests included two standard tests that have been used extensively to study the corrosion behavior of high level waste (HLW) glasses: the Material Characterization Center-1 (MCC-1) Test and the Product Consistency Test (PCT). Direct comparison of the results of tests with the reference CWF and HLW glasses indicate that the corrosion behaviors of the CWF and HLW glasses are very similar

Hot Fuel Examination Facility (HFEF)

Data.gov (United States)

Federal Laboratory Consortium — The Hot Fuel Examination Facility (HFEF) is one of the largest hot cells dedicated to radioactive materials research at Idaho National Laboratory (INL). The nation's...

Method to reduce the nitrogen oxide content of gaseous wastes

Energy Technology Data Exchange (ETDEWEB)

Klopp, G.; Sueto, J.; Szasz, K.; Szebenyi, I.; Winkler, G.; Machacs, M.; Palmai, G.

1980-11-13

The proposed process is suited for the denitrification of waste gases from nitric acid plants. It proceeds without an additional energy source with an integrated adsorption unit which guaranties the complete recirculation of the produced nitrogen oxides and allows the regeneration of the adsorbents by the use of the energy from the oxidation of nitrous oxide to nitric oxide. The desorption is carried out by the intermediate passage of the hot gases from the oxidizer through the adsorber.

Auto-radiographical study of activity distribution of high-dispersity hot particles from NPP

International Nuclear Information System (INIS)

Bonchev, Ts.; Andreev, Ts.; Todorov, N.; Andreev, P.; Kasabov, I.

1995-01-01

Autoradiograms of hot particles (HP) detected at Kozloduy NPP have been obtained with ordinary photographic film using direct contact. HP diameter and activity have been calculated by computer processing of the image.The distribution of HP weight and the precipitation time on surfaces has been estimated. Results on 37 HP found at the Kozloduy NPP units 1-4 and 9 HP found in special rooms containing waste are reported. The emitters are identified as 60 Co, 54 Mn and α emitters with specific activities 2.10 8 Bq/g, 2.10 7 Bq/g and 3.10 4 Bq/g respectively. 5 refs., 6 figs., 3 tabs

Nuclear facilities of EdF's operational hot base of Tricastin. 2009 annual report

International Nuclear Information System (INIS)

2010-01-01

This annual report is established on account of article 21 of the 2006-686 French law from June 13, 2006, relative to the transparency and safety in the nuclear domain. It describes, first, the nuclear facilities of the EdF operational hot base of Tricastin, then, the measures taken to ensure their safety (personnel radioprotection, actions implemented for nuclear safety improvement, organisation in crisis situation, external and internal controls, technical assessment of the facilities), and finally the procedures of management of radioactive wastes. A glossary and the viewpoint of the Committee of Hygiene, safety and working conditions about the content of the document conclude the report. (J.S.)

Treatment of solid waste highly contaminated by alpha emitters low-temperature impact crushing/leaching and incineration

International Nuclear Information System (INIS)

Carpentier, S.; Bertolotti, G.

1986-01-01

Reprocessing plants, hot laboratories, fuel fabrication plants all produce waste containing residual quantities of plutonium and uranium in oxide form which often reach some tens of grammes per m 3 . Appropriate treatment recovers an appreciable amount of fissile material, which could lead to the waste being declassified and able to be disposed of in near ground-level facilities. After a summary sorting at production level, waste can be sent to a low-temperature impact crushing/leaching unit which considerably reduces volumes to be stored. We call this process cryo-crushing/leaching. For burnable crushed particles, a further volume reduction may be obtained by incineration, an operation which is made easier by the low fissile material residue content. Incineration can, of course also be applied directly to burnable solid waste sorted at source and crushed following more conventional methods [fr

Hot Chemistry Laboratory decommissioning activities at IPEN/CNEN-SP, Brazil

International Nuclear Information System (INIS)

Camilo, Ruth L.; Lainetti, Paulo E.O.

2009-01-01

IPEN's fuel cycle activities were accomplished in laboratory and pilot plant scale and most facilities were built in the 70-80 years. Nevertheless, radical changes of the Brazilian nuclear policy in the beginning of 90's determined the interruption of several fuel cycle activities and facilities shutdown. Since then, IPEN has faced the problem of the pilot plants decommissioning considering that there was no experience/expertise in this field at all. In spite of this, some laboratory and pilot plant decommissioning activities have been performed in IPEN in the last years, even without previous experience and training support. One of the first decommissioning activities accomplished in IPEN involved the Hot Chemistry Laboratory. This facility was built in the beginning of the 80's with the proposal of supporting research and development in the nuclear chemistry area. It was decided to settle a new laboratory in the place where the Hot Chemistry Laboratory was installed, being necessary its total releasing from the radioactive contamination point of view. The previous work in the laboratory involved the manipulation of samples of irradiated nuclear fuel, besides plutonium-239 and uranium-233 standard solutions. There were 5 glove-boxes in the facility but only 3 were used with radioactive material. The glove-boxes contained several devices and materials, besides the radioactive compounds, such as: electric and electronic equipment, metallic and plastic pieces, chemical reagents, liquid and solid radioactive wastes, etc. The laboratory's decommissioning process was divided in 12 steps. This paper describes the procedures, problems faced and results related to the Hot Chemistry Laboratory decommissioning operations and its reintegration as a new laboratory of the Chemical and Environmental Technology Center (CQMA) - IPEN-CNEN/SP. (author)

The Properties of Polymer-Modified Asphalt Drived from Vacuum Bottom and Wastes of PE and SBR Plants

Directory of Open Access Journals (Sweden)

A. Yousefi

2008-12-01

Full Text Available The polymeric and non-polymeric petrochemical wastes are environmentally unfavorable materials. We have to found a way to reuse these wastes in a safe manner. In styrene-butadiene rubber (SBR production plant two major wastes are produced one is a fine rubber powder and the other is recycled N-methyl pyrrolidone (NMP. In high-density polyethylene (HDPE production plant also a low molecular weight polymer waste is produced which does not degrade in environment easily. On the other hand, bitumen is usually blended with thermoplastics, rubbers and an oil to produce polymermodified bitumens. Effect of SBR, PE and NMP wastes on vacuum bottom residue of crude oil distillation was investigated. The results of penetration, softening point and Frass tests revealed the advantages of SBR and PE wastes at high temperatures and suggest NMP waste as a good bitumen extender. The results of the Marshal test on the prepared polymer-modified hot-mix asphalt of the optimum formulation highlight the superiority of its strength and its lower flow as compared with those of standard 60/70 penetration grade bitumen.

Corrosion behaviors of a glass-bonded sodalite ceramic waste form and its constituents

International Nuclear Information System (INIS)

Lewis, M. A.; Ebert, W. L.; Morss, L.

1999-01-01

A ceramic waste form (CWF) of glass bonded sodalite is being developed as a waste form for the long-term immobilization of fission products and transuranic elements from the U.S. Department of Energy's activities on spent nuclear fuel conditioning. A durable waste form was prepared by hot isostatic pressing (HIP) a mixture of salt-loaded zeolite powders and glass frit. During HIP the zeolite is converted to sodalite, and the resultant CWF is been completed for durations of up to 182 days. Four dissolution modes were identified: dissolution of free salt, dissolution of the aluminosilicate matrix of sodalite and the accompanying dissolution of occluded salt, dissolution of the boroaluminosilicate matrix of the glass, and ion exchange. Synergies inherent to the CWF were identified by comparing the results of the tests with pure glass and sodalite with those of the composite CWF

Measuring hot flash phenomenonology using ambulatory prospective digital diaries

Science.gov (United States)

Fisher, William I.; Thurston, Rebecca C.

2016-01-01

Objective This study provides the description, protocol, and results from a novel prospective ambulatory digital hot flash phenomenon diary. Methods This study included 152 midlife women with daily hot flashes who completed an ambulatory electronic hot flash diary continuously for the waking hours of 3 consecutive days. In this diary, women recorded their hot flashes and accompanying characteristics and associations as the hot flashes occurred. Results Self-reported hot flash severity on the digital diaries indicated that the majority of hot flashes were rated as mild (41.3%) or moderate (43.7%). Severe (13.1%) and very severe (1.8%) hot flashes were less common. Hot flash bother ratings were rated as mild (43%), or moderate (33.5%), with fewer hot flashes reported bothersome (17.5%) or very bothersome (6%). The majority of hot flashes were reported as occurring on the on the face (78.9%), neck (74.7%), and chest (61.3%). Prickly skin was reported concurrently with 32% of hot flashes, 7% with anxiety and 5% with nausea. A novel finding, 38% of hot flashes were accompanied by a premonitory aura. Conclusion A prospective electronic digital hot flash diary allows for a more precise quantitation of hot flashes while overcoming many of the limitations of commonly employed retrospective questionnaires and paper diaries. Unique insights into the phenomenology, loci and associated characteristics of hot flashes were obtained using this device. The digital hot flash phenomenology diary is recommended for future ambulatory studies of hot flashes as a prospective measure of the hot flash experience. PMID:27404030

Performance of Hot Asphalt Mixtures Containing Plastic Bottles as Additive

Directory of Open Access Journals (Sweden)

Jan Hakeem

2017-01-01

Full Text Available This study focuses on evaluating the resistance of polymer modified asphalt mixes and the role played by asphalt in the realm of construction is undeniably important. Addition of polymers(PB as additives to asphalt helps to improve the strength and water repellent property of the mix and as well as helps environment in various ways and at the same time, analyzing its lower maintenance activities and service life is most important. The use of inexpensive polymers, in this case, waste polymers has without any doubt proven to be the most convenient way of reducing the cost of construction and at the same time maintaining quality. The main resolve for this research was to establish the effects of the use of plastic bottles on hot asphalt and its mixtures. In order to put this into perspective, varying percentages of asphalt mixtures were calculated and subjected to laboratory tests. The two-factor variance analysis (ANOVA was conducted to determine the significance at various confidence limits. The results indicate that the inclusion of Polyethylene Terephthalate (PET had a particularly substantial effect on the properties of asphalt. Consequently, it can encourage the re-utilization of waste in the manufacturing industry in an ecologically friendly and cost-effective way.

Computational prediction of protein hot spot residues.

Science.gov (United States)

Morrow, John Kenneth; Zhang, Shuxing

2012-01-01

Most biological processes involve multiple proteins interacting with each other. It has been recently discovered that certain residues in these protein-protein interactions, which are called hot spots, contribute more significantly to binding affinity than others. Hot spot residues have unique and diverse energetic properties that make them challenging yet important targets in the modulation of protein-protein complexes. Design of therapeutic agents that interact with hot spot residues has proven to be a valid methodology in disrupting unwanted protein-protein interactions. Using biological methods to determine which residues are hot spots can be costly and time consuming. Recent advances in computational approaches to predict hot spots have incorporated a myriad of features, and have shown increasing predictive successes. Here we review the state of knowledge around protein-protein interactions, hot spots, and give an overview of multiple in silico prediction techniques of hot spot residues.

Dimensionality of heavy metal distribution in waste disposal sites using nonlinear dynamics

International Nuclear Information System (INIS)

Modis, Kostas; Komnitsas, Kostas

2008-01-01

Mapping of heavy metal contamination in mining and waste disposal sites usually relies on geostatistical approaches and linear stochastic dynamics. The present paper aims to identify, using the Grassberger-Procaccia correlation dimension (CD) algorithm, the existence of a nonlinear deterministic and chaotic dynamic behaviour in the spatial pattern of arsenic, manganese and zinc concentration in a Russian coal waste disposal site. The analysis carried out yielded embedding dimension values ranging between 7 and 8 suggesting thus from a chaotic dynamic perspective that arsenic, manganese and zinc concentration in space is a medium dimensional problem for the regionalized scale considered in this study. This alternative nonlinear dynamics approach may complement conventional geostatistical studies and may be also used for the estimation of risk and the subsequent screening and selection of a feasible remediation scheme in wider mining and waste disposal sites. Finally, the synergistic effect of this study may be further elaborated if additional factors including among others presence of hot spots, density and depth of sampling, mineralogy of wastes and sensitivity of analytical techniques are taken into account

A Comparative Life Cycle Assessment of Hot Mixes Asphalt Containing Bituminous Binder Modified with Waste and Virgin Polymers

NARCIS (Netherlands)

Oliveira dos Santos, Joao Miguel; Cerezo, Veronique; Soudani, Khedoudja; Bressi, Sara

2018-01-01

This paper presents the results of a life cycle assessment undertaken to compare the potential environmental impacts associated with the use of asphalt surface mixtures produced with polymer modified bitumen with those of a conventional asphalt surface mixture. Seven types of hot mix asphalt

Uncertainty analysis for hot channel

International Nuclear Information System (INIS)

Panka, I.; Kereszturi, A.

2006-01-01

The fulfillment of the safety analysis acceptance criteria is usually evaluated by separate hot channel calculations using the results of neutronic or/and thermo hydraulic system calculations. In case of an ATWS event (inadvertent withdrawal of control assembly), according to the analysis, a number of fuel rods are experiencing DNB for a longer time and must be regarded as failed. Their number must be determined for a further evaluation of the radiological consequences. In the deterministic approach, the global power history must be multiplied by different hot channel factors (kx) taking into account the radial power peaking factors for each fuel pin. If DNB occurs it is necessary to perform a few number of hot channel calculations to determine the limiting kx leading just to DNB and fuel failure (the conservative DNBR limit is 1.33). Knowing the pin power distribution from the core design calculation, the number of failed fuel pins can be calculated. The above procedure can be performed by conservative assumptions (e.g. conservative input parameters in the hot channel calculations), as well. In case of hot channel uncertainty analysis, the relevant input parameters (k x, mass flow, inlet temperature of the coolant, pin average burnup, initial gap size, selection of power history influencing the gap conductance value) of hot channel calculations and the DNBR limit are varied considering the respective uncertainties. An uncertainty analysis methodology was elaborated combining the response surface method with the one sided tolerance limit method of Wilks. The results of deterministic and uncertainty hot channel calculations are compared regarding to the number of failed fuel rods, max. temperature of the clad surface and max. temperature of the fuel (Authors)

Waste management, waste resource facilities and waste conversion processes

International Nuclear Information System (INIS)

Demirbas, Ayhan

2011-01-01

In this study, waste management concept, waste management system, biomass and bio-waste resources, waste classification, and waste management methods have been reviewed. Waste management is the collection, transport, processing, recycling or disposal, and monitoring of waste materials. A typical waste management system comprises collection, transportation, pre-treatment, processing, and final abatement of residues. The waste management system consists of the whole set of activities related to handling, treating, disposing or recycling the waste materials. General classification of wastes is difficult. Some of the most common sources of wastes are as follows: domestic wastes, commercial wastes, ashes, animal wastes, biomedical wastes, construction wastes, industrial solid wastes, sewer, biodegradable wastes, non-biodegradable wastes, and hazardous wastes.

3-D thermoelastic analysis of the straight section of a PWR hot leg containing a hot spot using BEM

International Nuclear Information System (INIS)

Bains, R.S.; Sugimoto, J.

1995-01-01

A 3-D steady state thermoelastic analysis using the boundary element method has been successfully employed to investigate the structural response of the straight section of a pressurised water reactor hot leg containing a localised hot spot. With the present severe accident thermal boundary conditions, the analysis produces a nonuniform expansion across the hot leg thickness. This expansion was most predominant on the inner surface, especially at the hot spot location where surface swelling was obtained. Furthermore, the hot spot generates large tangential and axial tensile stresses on the outer surface. These could be detrimental to the integrity of the hot leg by acting as potential sites of crack initiation and subsequent propagation. (orig.)

Hot-pressing steatite bodies

International Nuclear Information System (INIS)

Aparicio Arroyo, E.

1967-01-01

Requirements for some special nuclear engineering ceramic shapes are: big size, impervious, dimensional accuracy and good mechanical and dielectric properties. Limitations of te conventional methods and advantages of te hot pressing techniques for the manufacturing of these shapes are discussed. Hot pressing characteristics of a certain steatite powder are studied. Occurrence of an optimum densification temperature just above the tale decomposition range is found. Experimental data show that the height/diameter ratio of the specimen has no effect on the sintering conditions. Increasing darkness from the graphite mould is detected above the optimum temperature. The hot-pressed steatite is compared with a fired dry-pressed sample of the same composition. (Author) 13 refs

HOT 2017

DEFF Research Database (Denmark)

Hannibal, Sara Stefansen

HOT er en kvalitativ undersøgelse, der hvert år diskuterer og undersøger en lille udvalgt skare af danskkyndige fagpersoners bud på, hvad de er optagede af på literacyområdet her og nu – altså hvilke emner, de vil vurdere som aktuelle at forholde sig til i deres nuværende praksis.......HOT er en kvalitativ undersøgelse, der hvert år diskuterer og undersøger en lille udvalgt skare af danskkyndige fagpersoners bud på, hvad de er optagede af på literacyområdet her og nu – altså hvilke emner, de vil vurdere som aktuelle at forholde sig til i deres nuværende praksis....

Report on best available techniques (BAT) at Finnish hot dip galvanizing plants in 2001; Selvitys parhaasta kaeytettaevissae olevasta tekniikasta (BAT) Suomen kuumasinkityslaitoksissa vuonna 2001

Energy Technology Data Exchange (ETDEWEB)

Hakala, I.

2002-02-01

Report presents hot dip (batch) galvanizing process and its emissions as well as measures to prevent and control them. Report has been prepared in cooperation with industry representatives and environmental authorities. Reference Document on Best Available Techniques in the Ferrous Metals Processing Industry and especially its part concerning batch galvanizing has been utilized. Also background information from the existing environmental permits as well as information collected by the questionnaires has been used. In year 2001 there were 18 hot dip (batch) galvanizing plants in Finland. Branch can be considered quite homogenous as there are only few variations in processes used. Waste management and recovery and as well as waste water treatment are well taken care of. At many plants treated process water is recycled back to the galvanizing process. Air emissions from hot dipping pot (dust, zinc, ammonium compounds, chlorides) are filtrated only at few plants. Therefore most of the environmental load at many plants is caused by air emissions as they can indirectly cause pollution also into soil and nearby water bodies. Instead of specific emission and consumption data and comparison between different techniques, report presents choice of measures and actions to be taken into account when applying best available techniques (BAT) at an installation level. Galvanizers can use the report as background information when assessing implementation of BAT at their own plants. Permitting authorities can utilize the report in the environmental permitting of the plants. (orig.)

Biodegradation of radioactive organic liquid waste from spent fuel reprocessing

International Nuclear Information System (INIS)

Ferreira, Rafael Vicente de Padua

2008-01-01

The research and development program in reprocessing of low burn-up spent fuel elements began in Brazil in 70's, originating the lab-scale hot cell, known as Celeste located at Nuclear and Energy Research Institute, IPEN - CNEN/SP. The program was ended at the beginning of 90's, and the laboratory was closed down. Part of the radioactive waste generated mainly from the analytical laboratories is stored waiting for treatment at the Waste Management Laboratory, and it is constituted by mixture of aqueous and organic phases. The most widely used technique for the treatment of radioactive liquid wastes is the solidification in cement matrix, due to the low processing costs and compatibility with a wide variety of wastes. However, organics are generally incompatible with cement, interfering with the hydration and setting processes, and requiring pre -treatment with special additives to stabilize or destroy them. The objective of this work can be divided in three parts: organic compounds characterization in the radioactive liquid waste; the occurrence of bacterial consortia from Pocos de Caldas uranium mine soil and Sao Sebastiao estuary sediments that are able to degrade organic compounds; and the development of a methodology to biodegrade organic compounds from the radioactive liquid waste aiming the cementation. From the characterization analysis, TBP and ethyl acetate were chosen to be degraded. The results showed that selected bacterial consortia were efficient for the organic liquid wastes degradation. At the end of the experiments the biodegradation level were 66% for ethyl acetate and 70% for the TBP. (author)

Progress report on safety research of high-level waste management for the period April 1986 to March 1987

International Nuclear Information System (INIS)

Nakamura, Haruto; Tashiro, Shingo

1987-08-01

Researches on high-level waste management at the High Level Waste Management Laboratory and the Waste Safety Testing Facility Operation Division of the Japan Atomic Energy Research Institute in the fiscal year of 1986 are reviewed in the report. Topics in the three sections are as follows: 1) Non-radioactive research has been continued on Synroc irradiation and modellings of waste form leaching. 2) Research results are described in the section of Safety Evaluation for Geological Disposal on engineered barriers, field tests, safety assessment models, migration, natural analogue, seabed disposal and conceptual design of a repository. 3) Adsorption behaviour of plutonium on leach-containers and migration of leached cesium in a rock column are described in the section of Safety Examination of Vitrified Forms in the Hot Cells of WASTEF. (author)

[History of hot spring bath treatment in China].

Science.gov (United States)

Hao, Wanpeng; Wang, Xiaojun; Xiang, Yinghong; Gu Li, A Man; Li, Ming; Zhang, Xin

2011-07-01

As early as the 7th century B.C. (Western Zhou Dynasty), there is a recording as 'spring which contains sulfur could treat disease' on the Wentang Stele written by WANG Bao. Wenquan Fu written by ZHANG Heng in the Easten Han Dynasty also mentioned hot spring bath treatment. The distribution of hot springs in China has been summarized by LI Daoyuan in the Northern Wei Dynasty in his Shuijingzhu which recorded hot springs in 41 places and interpreted the definition of hot spring. Bencao Shiyi (by CHEN Cangqi, Tang Dynasty) discussed the formation of and indications for hot springs. HU Zai in the Song Dynasty pointed out distinguishing hot springs according to water quality in his book Yuyin Conghua. TANG Shenwei in the Song Dynasty noted in Jingshi Zhenglei Beiji Bencao that hot spring bath treatment should be combined with diet. Shiwu Bencao (Ming Dynasty) classified hot springs into sulfur springs, arsenicum springs, cinnabar springs, aluminite springs, etc. and pointed out their individual indications. Geologists did not start the work on distribution and water quality analysis of hot springs until the first half of the 20th century. There are 972 hot springs in Wenquan Jiyao (written by geologist ZHANG Hongzhao and published in 1956). In July 1982, the First National Geothermal Conference was held and it reported that there were more than 2600 hot springs in China. Since the second half of the 20th century, hot spring sanatoriums and rehabilitation centers have been established, which promoted the development of hot spring bath treatment.

Nuclear Solid Waste Processing Design at the Idaho Spent Fuels Facility

International Nuclear Information System (INIS)

Dippre, M. A.

2003-01-01

A spent nuclear fuels (SNF) repackaging and storage facility was designed for the Idaho National Engineering and Environmental Laboratory (INEEL), with nuclear solid waste processing capability. Nuclear solid waste included contaminated or potentially contaminated spent fuel containers, associated hardware, machinery parts, light bulbs, tools, PPE, rags, swabs, tarps, weld rod, and HEPA filters. Design of the nuclear solid waste processing facilities included consideration of contractual, regulatory, ALARA (as low as reasonably achievable) exposure, economic, logistical, and space availability requirements. The design also included non-attended transfer methods between the fuel packaging area (FPA) (hot cell) and the waste processing area. A monitoring system was designed for use within the FPA of the facility, to pre-screen the most potentially contaminated fuel canister waste materials, according to contact- or non-contact-handled capability. Fuel canister waste materials which are not able to be contact-handled after attempted decontamination will be processed remotely and packaged within the FPA. Noncontact- handled materials processing includes size-reduction, as required to fit into INEEL permitted containers which will provide sufficient additional shielding to allow contact handling within the waste areas of the facility. The current design, which satisfied all of the requirements, employs mostly simple equipment and requires minimal use of customized components. The waste processing operation also minimizes operator exposure and operator attendance for equipment maintenance. Recently, discussions with the INEEL indicate that large canister waste materials can possibly be shipped to the burial facility without size-reduction. New waste containers would have to be designed to meet the drop tests required for transportation packages. The SNF waste processing facilities could then be highly simplified, resulting in capital equipment cost savings, operational