Waste processing system for product contaminated with radioactivity

International Nuclear Information System (INIS)

Sotoyama, Koichi; Takaya, Jun-ichi; Takahashi, Suehiro.

1987-01-01

Purpose: To enable to processing contaminated products while separating them into metals at high contamination level and non-metals at low contamination level. Constitution: Pulverized radioactive wastes conveyed on a conveyor belt are uniformly irradiated by a ring-illumination device and then they are picked-up by a television camera or the like. The picked-up signals are sent to an image processing device, applied with appropriate binarization and metal objects are separated by utilizing the light absorbing property of non-metal and light reflection property of metals. The graviational center for the metal object is calculated from the binarized image, the positional information is provided to a robot controller and the metal object is transferred to another position by a robot. Since only the metal object at high radioactive contamination level can be taken out separately, it is no more necessary to process the entire wastes as the high level decontamination products, to thereby provide an economical advantage. (Sekiya, K.)

Management of radioactive liquid waste at the Idaho Chemical Processing Plant

International Nuclear Information System (INIS)

Bendixsen, C.L.

1992-01-01

Highly radioactive liquid wastes (HLLW) are routinely produced during spent nuclear fuel processing at the Idaho Chemical Processing Plant (ICPP), located at the Idaho National Engineering Laboratory (INEL). This paper discusses the processes and safe practices for management of the radioactive process waste streams, which processes include collection, concentration, interim storage, calcination to granular solids, and long-term intermediate storage. Over four million gallons of HLLW have been converted to a recoverable granular solid form through waste liquid injection into a high-temperature, fluidized bed wherein the wastes are converted to their respective solid oxides. The development of a glass ceramic solid for the long-term permanent disposal of the high level waste (HLW) solids is also described

Method of processing radioactive metallic sodium with recycling alcohols

International Nuclear Information System (INIS)

Sakai, Takuhiko; Mitsuzuka, Norimasa.

1980-01-01

Purpose: To employ high safety alcohol procession and decrease the amount of wastes in the procession of radioactive metallic sodium discharged from LMFBR type reactors. Method: Radioactive metallic sodium containing long half-decay period nuclides such as cesium, strontium, barium, cerium, lanthanum or zirconium is dissolved in an alcohol at about 70% purity. After extracting the sodium alcoholate thus formed, gaseous hydrochloride is blown-in to separate the sodium alcoholate into alcohol and sodium chloride, and regenerated alcohol is used again for dissolving sodium metal. The sodium chloride thus separated is processed into solid wastes. (Furukawa, Y.)

Processing method for radioactive sludge

International Nuclear Information System (INIS)

Shoji, Yuichi; Kaneko, Masaaki.

1993-01-01

The concentration of radioactive sludges contained in a storage tank is controlled, thereafter, a filter is charged into a processing vessel to continuously conduct dewatering. Then, the radioactive sludges and an oxidizer are mixed by stirring using a stirring impeller and by vibrations using a vibrator. At the same time, thermic rays are irradiated by using infrared ray lamps to heat and decompose them. Since thermic rays reach the center of the radioactive sludges by the infrared ray lamps, ion exchange resins are sufficiently decomposed and carbonized into inorganic material. Then, a filling hardener such as mortar cement having a good flowability is charged to solidify the wastes. With such procedures, radioactive sludges can be stored under a stable condition for a long period of time by decomposing organic materials into inorganic materials and solidifying them. Further, an operator's radiation exposure dose can remarkably be reduced by applying a predetermined and a stabilization treatment in an identical processing vessel. (N.H.)

Radioactive waste processing container

International Nuclear Information System (INIS)

Ishizaki, Kanjiro; Koyanagi, Naoaki; Sakamoto, Hiroyuki; Uchida, Ikuo.

1992-01-01

A radioactive waste processing container used for processing radioactive wastes into solidification products suitable to disposal such as underground burying or ocean discarding is constituted by using cements. As the cements, calcium sulfoaluminate clinker mainly comprising calcium sulfoaluminate compound; 3CaO 3Al 2 O 3 CaSO 4 , Portland cement and aqueous blast furnace slug is used for instance. Calciumhydroxide formed from the Portland cement is consumed for hydration of the calcium sulfoaluminate clinker. According, calcium hydroxide is substantially eliminated in the cement constituent layer of the container. With such a constitution, damages such as crackings and peelings are less caused, to improve durability and safety. (I.N.)

Process and device for processing radioactive wastes

International Nuclear Information System (INIS)

1974-01-01

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

Site selection process for radioactive waste repository (radioactive facility) in Cuba as a fundamental safety criteria

International Nuclear Information System (INIS)

Vital, Jose Luis Peralta; Castillo, Reinaldo Gil; Chales Suarez, Gustavo; Rodriguez Reyes, Aymee

1999-01-01

The paper show the process of search carried out for the selection of the safest site in the National territory, in order to sitting the Facility (Repository) that will disposal the low and intermediate level radioactive wastes, as well as the possible Storage Facility for nuclear spent Fuel (radioactive wastes of high activity). We summarize the obtained Methodology and the Criterions of exclusion adopted for the development of the Process of site selection, as well as the current condition of the researches that will permit the obtaining of the nominative objectives. (author)

Development of very low-level radioactive waste sequestration process criteria

Energy Technology Data Exchange (ETDEWEB)

Chan, N.; Wong, P., E-mail: nicholas.chan@cnl.ca [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada)

2015-12-15

Segregating radioactive waste at the source and reclassifying radioactive waste to lower waste classes are the key activities to reduce the environmental footprint and long-term liability. In the Canadian Standards Association's radioactive waste classification system, there are 2 sub-classes within low-level radioactive waste: very short-lived radioactive waste and very low-level radioactive waste (VLLW). VLLW has a low hazard potential but is above the Canadian unconditional clearance criteria as set out in Schedule 2 of Nuclear Substances and Devices Regulations. Long-term waste management facilities for VLLW do not require a high degree of containment and isolation. In general, a relatively low-cost near-surface facility with limited regulatory control is suitable for VLLW. At Canadian Nuclear Laboratories' Chalk River Laboratories site an initiative, VLLW Sequestration, was implemented in 2013 to set aside potential VLLW for temporary storage and to be later dispositioned in the planned VLLW facility. As of May 2015, a total of 236m{sup 3} resulting in approximately $1.1 million in total savings have been sequestered. One of the main hurdles in implementing VLLW Sequestration is the development of process criteria. Waste Acceptance Criteria (WAC) are used as a guide or as requirements for determining whether waste is accepted by the waste management facility. Establishment of the process criteria ensures that segregated waste materials have a high likelihood to meet the VLLW WAC and be accepted into the planned VLLW facility. This paper outlines the challenges and various factors which were considered in the development of interim process criteria. (author)

Japan's Siting Process for the Geological Disposal of High-level Radioactive Waste - An International Peer Review

International Nuclear Information System (INIS)

Brassinnes, Stephane; Fabbri, Olivier; Rubenstone, James; Seppaelae, Timo; Siemann, Michael; ); Kwong, Gloria; )

2016-01-01

The Nuclear Energy Agency carried out an independent peer review of Japan's siting process and criteria for the geological disposal of high-level radioactive waste in May 2016. The review concluded that Japan's site screening process is generally in accordance with international practices. As the goal of the siting process is to locate a site - that is both appropriate and accepted by the community - to host a geological disposal facility for high-level radioactive waste, the international review team emphasises in this report the importance of maintaining an open dialogue and interaction between the regulator, the implementer and the public. Dialogue should begin in the early phases and continue throughout the siting process. The international review team also underlines the importance of taking into account feasibility aspects when selecting a site for preliminary investigations, but suggests that it would be inappropriate to set detailed scientific criteria for nationwide screening at this stage. The team has provided extensive advisory remarks in the report as opportunities for improvement, including the recommendation to use clear and consistent terminology in defining the site screening criteria as it is a critical factor in a successful siting process. (authors)

Disposal of high level and intermediate level radioactive wastes

International Nuclear Information System (INIS)

Flowers, R.H.

1991-01-01

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

Method of processing low-level radioactive liquid wastes

International Nuclear Information System (INIS)

Matsunaga, Ichiro; Sugai, Hiroshi.

1984-01-01

Purpose: To effectively reduce the radioactivity density of low-level radioactive liquid wastes discharged from enriched uranium conversion processing steps or the likes. Method: Hydrazin is added to low-level radioactive liquid wastes, which are in contact with iron hydroxide-cation exchange resins prepared by processing strongly acidic-cation exchange resins with ferric chloride and aqueous ammonia to form hydrorizates of ferric ions in the resin. Hydrazine added herein may be any of hydrazine hydrate, hydrazine hydrochloride and hydranine sulfate. The preferred addition amount is more than 100 mg per one liter of the liquid wastes. If it is less than 100 mg, the reduction rate for the radioactivety density (procession liquid density/original liquid density) is decreased. This method enables to effectively reduce the radioactivity density of the low-level radioactive liquid wastes containing a trace amount of radioactive nucleides. (Yoshihara, H.)

Method of processing radioactive wastes

International Nuclear Information System (INIS)

Katada, Katsuo.

1986-01-01

Purpose: To improve the management for radioactive wastes containers thereby decrease the amount of stored matters by arranging the radioactive wastes containers in the order of their radioactivity levels. Method: The radiation doses of radioactive wastes containers arranged in the storing area before volume-reducing treatment are previously measured by a dosemeter. Then, a classifying machine is actuated to hoist the containers in the order to their radiation levels and the containers are sent out passing through conveyor, surface contamination gage, weight measuring device and switcher to a volume-reducing processing machine. The volume-reduced products are packed each by several units to the storing containers. Thus, the storing containers after stored for a certain period of time can be transferred in an assembled state. (Kawakami, Y.)

Analogues to features and processes of a high-level radioactive waste repository proposed for Yucca Mountain, Nevada

Science.gov (United States)

Simmons, Ardyth M.; Stuckless, John S.; with a Foreword by Abraham Van Luik, U.S. Department of Energy

2010-01-01

Natural analogues are defined for this report as naturally occurring or anthropogenic systems in which processes similar to those expected to occur in a nuclear waste repository are thought to have taken place over time periods of decades to millennia and on spatial scales as much as tens of kilometers. Analogues provide an important temporal and spatial dimension that cannot be tested by laboratory or field-scale experiments. Analogues provide one of the multiple lines of evidence intended to increase confidence in the safe geologic disposal of high-level radioactive waste. Although the work in this report was completed specifically for Yucca Mountain, Nevada, as the proposed geologic repository for high-level radioactive waste under the U.S. Nuclear Waste Policy Act, the applicability of the science, analyses, and interpretations is not limited to a specific site. Natural and anthropogenic analogues have provided and can continue to provide value in understanding features and processes of importance across a wide variety of topics in addressing the challenges of geologic isolation of radioactive waste and also as a contribution to scientific investigations unrelated to waste disposal. Isolation of radioactive waste at a mined geologic repository would be through a combination of natural features and engineered barriers. In this report we examine analogues to many of the various components of the Yucca Mountain system, including the preservation of materials in unsaturated environments, flow of water through unsaturated volcanic tuff, seepage into repository drifts, repository drift stability, stability and alteration of waste forms and components of the engineered barrier system, and transport of radionuclides through unsaturated and saturated rock zones.

Process for final storage of high level radioactive fission product solution

International Nuclear Information System (INIS)

Ellis, J.R.B.; Fries, B.A.

1984-01-01

In this process for the storage of an aqueous solution of radioactive nuclides, the solution is diluted with system water, which is obtained from a reservoir below the bottom of the sea in a porous geological stratum. After dilution, the diluted solution is injected into the same geological stratum under the bottom of the sea. (orig.) [de

Separating and stabilizing phosphate from high-level radioactive waste: process development and spectroscopic monitoring.

Science.gov (United States)

Lumetta, Gregg J; Braley, Jenifer C; Peterson, James M; Bryan, Samuel A; Levitskaia, Tatiana G

2012-06-05

Removing phosphate from alkaline high-level waste sludges at the Department of Energy's Hanford Site in Washington State is necessary to increase the waste loading in the borosilicate glass waste form that will be used to immobilize the highly radioactive fraction of these wastes. We are developing a process which first leaches phosphate from the high-level waste solids with aqueous sodium hydroxide, and then isolates the phosphate by precipitation with calcium oxide. Tests with actual tank waste confirmed that this process is an effective method of phosphate removal from the sludge and offers an additional option for managing the phosphorus in the Hanford tank waste solids. The presence of vibrationally active species, such as nitrate and phosphate ions, in the tank waste processing streams makes the phosphate removal process an ideal candidate for monitoring by Raman or infrared spectroscopic means. As a proof-of-principle demonstration, Raman and Fourier transform infrared (FTIR) spectra were acquired for all phases during a test of the process with actual tank waste. Quantitative determination of phosphate, nitrate, and sulfate in the liquid phases was achieved by Raman spectroscopy, demonstrating the applicability of Raman spectroscopy for the monitoring of these species in the tank waste process streams.

Radioactive gas processing device

International Nuclear Information System (INIS)

Kita, Kaoru; Minemoto, Masaki; Takezawa, Kazuaki; Okazaki, Akira; Kumagaya, Koji.

1982-01-01

Purpose: To simplify the structure of a gas processing system which has hitherto been much complicated by the recyclic use of molecular sieve regeneration gas, by enabling to release the regeneration gas to outside in a once-through manner. Constitution: The system comprises a cooler for receiving and cooling gases to be processed containing radioactive rare gases, moisture-removing pipelines each connected in parallel to the exit of the cooler and having switching valves and a moisture removing column disposed between the valves and a charcoal absorber in communication with the moisture removing pipelines. Pipelines for flowing regeneration heating gases are separately connected to the moisture removing columns, and molecular sieve is charged in the moisture removing column by the amount depending on the types of the radioactive rare gases. (Aizawa, K.)

Production of adsorbent from palm shell for radioactive iodine scrubbing process

International Nuclear Information System (INIS)

Mohamad Azman Che Mat Isa; Ku Halim Ku Hamid; Muhd Noor Muhd Yunus; Mohamad Puad Abu; Abdul Halim Badaruddin; Mohammad Nizammudin Abd Aziz; Muhd Ridwan Abdul Rahim

2010-01-01

The biggest biomass source in Malaysia comes from oil palm industry. According to the statistic of year 2004, Malaysia produced 40 million tones per year of biomass which 30 million tones of biomass originated from the oil palm industries. Therefore, the biomass waste such as palm kernel shell can be used to produce granular adsorbent for radioactive materials. For that reason, a newly system, called Rocking Kiln - Fluidized Bed (RK - FB) was developed to utilize large amount of the biomass to produce high value added product. Charcoal or chemically produced activated carbon could be produced by using the kiln. Washing process was introduced to remove particles, minerals and volatile matters from charcoal produced and then would create more surface area in the adsorbent by creating more active sites. In this research, the adsorbent produced was used to scrub iodine 131. In nuclear power reactor, iodine isotope 131 is produced during nuclear fission, and this elementary radioactive iodine may pollute exhaust air streams that could cause thyroid cancer. For removal of radioactive iodine, normally a potassium iodide - impregnated activated carbon (KI - AC) is used. Thus, a process will be developed to produce KI - AC and this product will be used to calculate the efficiency to remove the radioactive iodine 131.The results obtain show that adsorbent produced has a high potential to be used in radioactive adsorbing and likely more economics. This paper will elaborate further the experimental set-up of in Kiln - Fluidized Bed (RK - FB), adsorbent quality and radioactive scrubbing process. (author)

Radioactive waste processing method

International Nuclear Information System (INIS)

Kikuchi, Makoto; Kamiya, Kunio; Yusa, Hideo.

1976-01-01

Object: To form radioactive wastes into a pellet-like solid body having high strength. Structure: Liquid waste containing a radioactive material is heated into a powdery body. Granular solid matter such as sand greater in diameter than grain size of the powdery body are mixed into the powdery body, and thereafter the mixture is formed by a granulator into a pellet-like solid body. The thus formed material is introduced into a drum can, into which a thermoplastic material such as asphalt is poured into the can and cooled so that the asphalt is impregnated inside the pellet to obtain a solid having high strength. (Furukawa, Y.)

Amersham's high radioactivity

Energy Technology Data Exchange (ETDEWEB)

Caulkin, S

1984-11-01

An account is given of the history, organisation and work of Amersham International, to produce radioactive and other products for use in medical diagnosis and therapy, in research in the life sciences, and in industrial processes and control systems. The account covers the developments from the war-time work of Thorium Ltd., on naturally occurring radioactive materials, through the post-war expansion into the field of artificial radioisotopes, as the Radiochemical Centre (part of the UK Atomic Energy Authority), to the recent reorganisation and privatization. The width of the range of activities and products available is emphasised, with examples.

Amersham's high radioactivity

International Nuclear Information System (INIS)

Caulkin, S.

1984-01-01

An account is given of the history, organisation and work of Amersham International, to produce radioactive and other products for use in medical diagnosis and therapy, in research in the life sciences, and in industrial processes and control systems. The account covers the developments from the war-time work of Thorium Ltd., on naturally occurring radioactive materials, through the post-war expansion into the field of artificial radioisotopes, as the Radiochemical Centre (part of the UK Atomic Energy Authority), to the recent reorganisation and privatization. The width of the range of activities and products available is emphasised, with examples. (U.K.)

Microwave processing of radioactive materials-I

International Nuclear Information System (INIS)

White, T.L.; Berry, J.B.

1989-01-01

This paper is the first of two papers that reviews the major past and present applications of microwave energy for processing radioactive materials, with particular emphasis on processing radioactive wastes. Microwave heating occurs through the internal friction produced inside a dielectric material when its molecules vibrate in response to an oscillating microwave field. For this presentation, we shall focus on the two FCC-approved microwave frequencies for industrial, scientific, and medical use, 915 and 2450 MHz. Also, because of space limitations, we shall postpone addressing plasma processing of hazardous wastes using microwave energy until a later date. 13 refs., 4 figs

Method and equipment of processing radioactive laundry wastes

International Nuclear Information System (INIS)

Shirai, Takamori; Suzuki, Takeo; Tabata, Masayuki; Takada, Takao; Yamaguchi, Shin-ichi; Noda, Tetsuya.

1985-01-01

Purpose: To effectively process radioactive laundry wastes generated due to water-washing after dry-cleaning of protective clothings which have been put on in nuclear facilities. Method: Dry cleaning soaps and ionic radioactive materials contained in radioactive laundry wastes are selectively adsorbed to decontaminate by adsorbents. Then, the adsorbents having adsorbed dry cleaning soaps and ionic radioactive materials are purified by being removed with these radioactive materials. The purified adsorbents are re-used. (Seki, T.)

Radioactive Dry Process Material Treatment Technology Development

Energy Technology Data Exchange (ETDEWEB)

Park, J. J.; Hung, I. H.; Kim, K. K. (and others)

2007-06-15

The project 'Radioactive Dry Process Material Treatment Technology Development' aims to be normal operation for the experiments at DUPIC fuel development facility (DFDF) and safe operation of the facility through the technology developments such as remote operation, maintenance and pair of the facility, treatment of various high level process wastes and trapping of volatile process gases. DUPIC Fuel Development Facility (DFDF) can accommodate highly active nuclear materials, and now it is for fabrication of the oxide fuel by dry process characterizing the proliferation resistance. During the second stage from march 2005 to February 2007, we carried out technology development of the remote maintenance and the DFDF's safe operation, development of treatment technology for process off-gas, and development of treatment technology for PWR cladding hull and the results was described in this report.

Radioactive liquid waste processing system

International Nuclear Information System (INIS)

Noda, Tetsuya; Kuramitsu, Kiminori; Ishii, Tomoharu.

1997-01-01

The present invention provides a system for processing radioactive liquid wastes containing laundry liquid wastes, shower drains or radioactive liquid wastes containing chemical oxygen demand (COD) ingredients and oil content generated from a nuclear power plant. Namely, a collecting tank collects radioactive liquid wastes. A filtering device is connected to the exit of the collective tank. A sump tank is connected to the exit of the filtering device. A powdery active carbon supplying device is connected to the collecting tank. A chemical fluid tank is connected to the collecting tank and the filtering device by way of chemical fluid injection lines. Backwarding pipelines connect a filtered water flowing exit of the filtering device and the collecting tank. The chemical solution is stored in the chemical solution tank. Then, radioactive materials in radioactive liquid wastes generated from a nuclear power plant are removed by the filtering device. The water quality standard specified in environmental influence reports can be satisfied. In the filtering device, when the filtering flow rate is reduced, the chemical fluid is supplied from the chemical fluid tank to the filtering device to recover the filtering flow rate. (I.S.)

Disposal of radioactive wastes

International Nuclear Information System (INIS)

Dlouhy, Z.

1982-01-01

This book provides information on the origin, characteristics and methods of processing of radioactive wastes, as well as the philosophy and practice of their storage and disposal. Chapters are devoted to the following topics: radioactive wastes, characteristics of radioactive wastes, processing liquid and solid radioactive wastes, processing wastes from spent fuel reprocessing, processing gaseous radioactive wastes, fixation of radioactive concentrates, solidification of high-level radioactive wastes, use of radioactive wastes as raw material, radioactive waste disposal, transport of radioactive wastes and economic problems of radioactive wastes disposal. (C.F.)

Remote ignitability analysis of high-level radioactive waste

International Nuclear Information System (INIS)

Lundholm, C.W.; Morgan, J.M.; Shurtliff, R.M.; Trejo, L.E.

1992-09-01

The Idaho Chemical Processing Plant (ICPP), was used to reprocess nuclear fuel from government owned reactors to recover the unused uranium-235. These processes generated highly radioactive liquid wastes which are stored in large underground tanks prior to being calcined into a granular solid. The Resource Conservation and Recovery Act (RCRA) and state/federal clean air statutes require waste characterization of these high level radioactive wastes for regulatory permitting and waste treatment purposes. The determination of the characteristic of ignitability is part of the required analyses prior to calcination and waste treatment. To perform this analysis in a radiologically safe manner, a remoted instrument was needed. The remote ignitability Method and Instrument will meet the 60 deg. C. requirement as prescribed for the ignitability in method 1020 of SW-846. The method for remote use will be equivalent to method 1020 of SW-846

Processing method of radioactive metal wastes

International Nuclear Information System (INIS)

Uetake, Naoto; Urata, Megumu; Sato, Masao.

1985-01-01

Purpose: To reduce the volume and increase the density of radioactive metal wastes easily while preventing scattering of radioactivity and process them into suitable form to storage and treatment. Method: Metal wastes mainly composed of zirconium are discharged from nuclear power plants or fuel re-processing plants, and these metals such as zirconium and titanium vigorously react with hydrogen and rapidly diffuse as hydrides. Since the hydrides are extremely brittle and can be pulverized easily, they can be volume-reduced. However, since metal hydrides have no ductility, dehydrogenation is applied for the molding fabrication in view of the subsequent storage and processing. The dehydrogenation is easy like the hydrogenation and fine metal pieces can be molded in a small compression device. For the dehydrogenation, a temperature is slightly increased as compared with that in the hydrogenation, pressure is reduced through the vacuum evacuation system and the removed hydrogen is purified for reuse. The upper limit for the temperature of the hydrogenation is 680 0 C in order to prevent the scttering of radioactivity. (Kamimura, M.)

Cementification for radioactive waste including high-concentration sodium sulfate and high-concentration radioactive nuclide

International Nuclear Information System (INIS)

Miyamoto, Shinya; Sato, Tatsuaki; Sasoh, Michitaka; Sakurai, Jiro; Takada, Takao

2005-01-01

For the cementification of radioactive waste that has large concentrations of sodium sulfate and radioactive nuclide, a way of fixation for sulfate ion was studied comprising the pH control of water in contact with the cement solid, and the removal of the excess water from the cement matrix to prevent hydrogen gas generation with radiolysis. It was confirmed that the sulfate ion concentration in the contacted water with the cement solid is decreased with the formation of ettringite or barium sulfate before solidification, the pH value of the pore water in the cement solid can control less than 12.5 by the application of zeolite and a low-alkali cement such as alumina cement or fly ash mixed cement, and removal of the excess water from the cement matrix by heating is possible with aggregate addition. Consequently, radioactive waste including high-concentration sodium sulfate and high-concentration radioactive nuclide can be solidified with cementitious materials. (author)

Handbook of high-level radioactive waste transportation

International Nuclear Information System (INIS)

Sattler, L.R.

1992-10-01

The High-Level Radioactive Waste Transportation Handbook serves as a reference to which state officials and members of the general public may turn for information on radioactive waste transportation and on the federal government's system for transporting this waste under the Civilian Radioactive Waste Management Program. The Handbook condenses and updates information contained in the Midwestern High-Level Radioactive Waste Transportation Primer. It is intended primarily to assist legislators who, in the future, may be called upon to enact legislation pertaining to the transportation of radioactive waste through their jurisdictions. The Handbook is divided into two sections. The first section places the federal government's program for transporting radioactive waste in context. It provides background information on nuclear waste production in the United States and traces the emergence of federal policy for disposing of radioactive waste. The second section covers the history of radioactive waste transportation; summarizes major pieces of legislation pertaining to the transportation of radioactive waste; and provides an overview of the radioactive waste transportation program developed by the US Department of Energy (DOE). To supplement this information, a summary of pertinent federal and state legislation and a glossary of terms are included as appendices, as is a list of publications produced by the Midwestern Office of The Council of State Governments (CSG-MW) as part of the Midwestern High-Level Radioactive Waste Transportation Project

Key scientific challenges in geological disposal of high level radioactive waste

International Nuclear Information System (INIS)

Wang Ju

2007-01-01

The geological disposal of high radioactive waste is a challenging task facing the scientific and technical world. This paper introduces the latest progress of high level radioactive disposal programs in the latest progress of high level radioactive disposal programs in the world, and discusses the following key scientific challenges: (1) precise prediction of the evolution of a repository site; (2) characteristics of deep geological environment; (3) behaviour of deep rock mass, groundwater and engineering material under coupled con-ditions (intermediate to high temperature, geostress, hydraulic, chemical, biological and radiation process, etc); (4) geo-chemical behaviour of transuranic radionuclides with low concentration and its migration with groundwater; and (5) safety assessment of disposal system. Several large-scale research projects and several hot topics related with high-level waste disposal are also introduced. (authors)

Recovering method for high level radioactive material

International Nuclear Information System (INIS)

Fukui, Toshiki

1998-01-01

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

Hydrological performance assessment on siting the high level radioactive waste repository

International Nuclear Information System (INIS)

Guo Yonghai; Liu Shufen; Wang Ju; Wang Zhiming; Su Rui; Lv Chuanhe; Zong Zihua

2007-01-01

Based on the research experiences in China and some developed countries in the world, the processes and methods on hydrological performance assessment for the siting of high radioactive repository are discussed in this paper. The methods and contents of hydrological performance assessment are discussed respectively for region, area and site hydrological investigation stages. At the same time, the hydrological performance assessment of the potential site for high level radioactive waste in China is introduced. (authors)

Radioactive sealed sources production process for industrial radiography

International Nuclear Information System (INIS)

Santos, Paulo de S.; Ngunga, Daniel M.G.; Camara, Julio R.; Vasquez, Pablo A.S.

2017-01-01

providing products and services to the private and governmental Brazilian users of industrial radiography and nucleonic control systems. Radioactive sealed sources are commonly used in nondestructive tests as radiography to make inspections and verify the internal structure and integrity of materials and in nucleonic gauges to control level, density, viscosity, etc. in on-line industrial processes. One of the most important activities carried out by this laboratory is related to the inspection of source projectors devices used in industrial radiography and its constituent parts as well as remote handle control assembly drive cable and guide tube systems. The laboratory also provide for the users iridium-192, cobalt-60 and selenium-75 sealed sources and performs quality control tests replacing spent or contaminated radiative sources. All discard of radioactive source is treated as radioactive waste. Additionally, administrative and commercial processes and protocols for exportation and transport of radioactive material are developed by specialized departments. In this work are presented the mean processes and procedures used by the Sealed Source Production Laboratory such as the arrival of the radioactive material to the laboratory and the source projectors, mechanical inspections, source loading, source leaking tests, etc. (author)

Radioactive sealed sources production process for industrial radiography

Energy Technology Data Exchange (ETDEWEB)

Santos, Paulo de S.; Ngunga, Daniel M.G.; Camara, Julio R.; Vasquez, Pablo A.S., E-mail: psantos@ipen.br, E-mail: hobeddaniel@gmail.com, E-mail: jrcamara@ipen.br, E-mail: pavsalva@ipen.br [Instituto de Pesquisas Energética s e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil)

2017-07-01

providing products and services to the private and governmental Brazilian users of industrial radiography and nucleonic control systems. Radioactive sealed sources are commonly used in nondestructive tests as radiography to make inspections and verify the internal structure and integrity of materials and in nucleonic gauges to control level, density, viscosity, etc. in on-line industrial processes. One of the most important activities carried out by this laboratory is related to the inspection of source projectors devices used in industrial radiography and its constituent parts as well as remote handle control assembly drive cable and guide tube systems. The laboratory also provide for the users iridium-192, cobalt-60 and selenium-75 sealed sources and performs quality control tests replacing spent or contaminated radiative sources. All discard of radioactive source is treated as radioactive waste. Additionally, administrative and commercial processes and protocols for exportation and transport of radioactive material are developed by specialized departments. In this work are presented the mean processes and procedures used by the Sealed Source Production Laboratory such as the arrival of the radioactive material to the laboratory and the source projectors, mechanical inspections, source loading, source leaking tests, etc. (author)

Method of processing radioactive wastes

International Nuclear Information System (INIS)

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

1979-01-01

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

Decoding Environmental Processes Using Radioactive Isotopes for the Post-Radioactive Contamination Recovery Assessment

Science.gov (United States)

Yasumiishi, Misa; Nishimura, Taku; Osawa, Kazutoshi; Renschler, Chris

2017-04-01

The continual monitoring of environmental radioactive levels in Fukushima, Japan following the nuclear plant accident in March 2011 provides our society with valuable information in two ways. First, the collected data can be used as an indicator to assess the progress of decontamination efforts. Secondly, the collected data also can be used to understand the behavior of radioactive isotopes in the environment which leads to further understanding of the landform processes. These two aspects are inseparable for us to understand the effects of radioactive contamination in a dynamic environmental system. During the summer of 2016, 27 soil core samples were collected on a farmer's land (rice paddies and forest) in Fukushima, about 20 km northwest of the nuclear plant. Each core was divided into 2.0 - 3.0 cm slices for the Cs-134, Cs-137, and I-131 level measurement. The collected data is being analyzed from multiple perspectives: temporal, spatial, and geophysical. In the forest area, even on the same hillslope, multiple soil types and horizon depths were observed which indicates the challenges in assessing the subsurface radioactive isotope movements. It appears that although highly humic soils show higher or about the same level of radioactivity in the surface layers, as the depth increased, the radioactivity decreased more in those samples compared with more sandy soils. With regard to the direction a slope faces and the sampling altitudes, the correlation between those attributes and radioactivity levels is inconclusive at this moment. The altitude might have affected the fallout level on a single hillslope-basis. However, to determine the correlation, further sampling and the detailed analysis of vegetation and topography might be necessary. Where the surface soil was scraped and new soil was brought in, former rice paddy surface layers did show three-magnitude levels lower of radioactivity in the top layer when compared with forest soils. At the foot of forest

Social acceptance process model for ensuring the high-level radioactive waste disposal site

International Nuclear Information System (INIS)

Wada, Ryutaro; Tanaka, Satoru; Nagasaki, Shinya

2009-01-01

Generally speaking, a vast, advanced and unfamiliar science and technology are unacceptable to the public for fear of their unknown nature. Here, the social acceptance process model was examined on the basis of the analysis of the cause phenomenon and numerical grounds, by referring to the problems on the application of literature documentation for location examination of a high-level radioactive waste disposal site in Toyo town in Kochi Pref. in April 2007. In analyzing the Toyo town case, we have found a possibility that the majority of local residents knew very little about the object opposed by the fringe route processing. To ensure a healthy decision making by the public, it is vital to convey fundamental information using sufficient wide-area PR media before the issue becomes actual. After the issue becomes actual, dialog with residents through a careful technology assessment is indispensable. The authors focus attention on the decision-making process of human beings from the social and psychological viewpoints, and point out that it is desirable for promoting social acceptance by adopting two approaches: a direct approach aiming at better intelligibility for the different resident layers and a deductive approach in technological essence. (author)

Radioactive waste processing and disposal

International Nuclear Information System (INIS)

1980-01-01

This compilation contains 4144 citations of foreign and domestic reports, journal articles, patents, conference proceedings, and books pertaining to radioactive waste processing and disposal. Five indexes are provided: Corporate Author, Personal Author, Subject, Contract Number, and Report Number

Demonstration of the Defense Waste Processing Facility vitrification process for Tank 42 radioactive sludge -- Glass preparation and characterization

International Nuclear Information System (INIS)

Bibler, N.E.; Fellinger, T.L.; Marshall, K.M.; Crawford, C.L.; Cozzi, A.D.; Edwards, T.B.

1999-01-01

The Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS) is currently processing and immobilizing the radioactive high level waste sludge at SRS into a durable borosilicate glass for final geological disposal. The DWPF has recently finished processing the first radioactive sludge batch, and is ready for the second batch of radioactive sludge. The second batch is primarily sludge from Tank 42. Before processing this batch in the DWPF, the DWPF process flowsheet has to be demonstrated with a sample of Tank 42 sludge to ensure that an acceptable melter feed and glass can be made. This demonstration was recently completed in the Shielded Cells Facility at SRS. An earlier paper in these proceedings described the sludge composition and processes necessary for producing an acceptable melter fee. This paper describes the preparation and characterization of the glass from that demonstration. Results substantiate that Tank 42 sludge after mixing with the proper amount of glass forming frit (Frit 200) can be processed to make an acceptable glass

Experimental study on pyrolysis incineration process for radioactive wastes

International Nuclear Information System (INIS)

Ma Mingxie; Qiu Mingcai; Wang Peiyi; Zhou Lianquan; Liu Xiaoqin

1993-01-01

In order to treat combustible radioactive wastes containing plastics and rubber in a considerable amount, a pyrolysis incineration process has been developed. Laboratory study and pilot test for the technology were performed. The results obtained in pilot test show that the waste containing a larger amount of plastics and rubber can be burnt perfectly in given technologic conditions, with a high volume-reduction factor obtained, and the process is easy to control

Melt processing of radioactive waste: A technical overview

International Nuclear Information System (INIS)

Schlienger, M.E.; Buckentin, J.M.; Damkroger, B.K.

1997-01-01

Nuclear operations have resulted in the accumulation of large quantities of contaminated metallic waste which are stored at various DOE, DOD, and commercial sites under the control of DOE and the Nuclear Regulatory Commission (NRC). This waste will accumulate at an increasing rate as commercial nuclear reactors built in the 1950s reach the end of their projected lives, as existing nuclear powered ships become obsolete or unneeded, and as various weapons plants and fuel processing facilities, such as the gaseous diffusion plants, are dismantled, repaired, or modernized. For example, recent estimates of available Radioactive Scrap Metal (RSM) in the DOE Nuclear Weapons Complex have suggested that as much as 700,000 tons of contaminated 304L stainless steel exist in the gaseous diffusion plants alone. Other high-value metals available in the DOE complex include copper, nickel, and zirconium. Melt processing for the decontamination of radioactive scrap metal has been the subject of much research. A major driving force for this research has been the possibility of reapplication of RSM, which is often very high-grade material containing large quantities of strategic elements. To date, several different single and multi-step melting processes have been proposed and evaluated for use as decontamination or recycling strategies. Each process offers a unique combination of strengths and weaknesses, and ultimately, no single melt processing scheme is optimum for all applications since processes must be evaluated based on the characteristics of the input feed stream and the desired output. This paper describes various melt decontamination processes and briefly reviews their application in developmental studies, full scale technical demonstrations, and industrial operations

Description of a Multipurpose Processing and Storage Complex for the Hanford Site's radioactive material

International Nuclear Information System (INIS)

Nyman, D.H.; Wolfe, B.A.; Hoertkorn, T.R.

1993-05-01

The mission of the US Department of Energy's (DOE) Hanford Site has changed from defense nuclear materials production to that of waste management/disposal and environmental restoration. ne Multipurpose Processing and Storage Complex (MPSC) is being designed to process discarded waste tank internal hardware contaminated with mixed wastes, failed melters from the vitrification plant, and other Hanford Site high-level solid waste. The MPSC also will provide interim storage of other radioactive materials (irradiated fuel, canisters of vitrified high-level waste [HLW], special nuclear material [SNM], and other designated radioactive materials)

The management of high-level radioactive wastes

International Nuclear Information System (INIS)

Lennemann, Wm.L.

1979-01-01

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

Defense Waste Processing Facility -- Radioactive operations -- Part 3 -- Remote operations

International Nuclear Information System (INIS)

Barnes, W.M.; Kerley, W.D.; Hughes, P.D.

1997-01-01

The Savannah River Site's Defense Waste Processing Facility (DWPF) near Aiken, South Carolina is the nation's first and world's largest vitrification facility. Following a ten year construction period and nearly three years of non-radioactive testing, the DWPF began radioactive operations in March 1996. Radioactive glass is poured from the joule heated melter into the stainless steel canisters. The canisters are then temporarily sealed, decontaminated, resistance welded for final closure, and transported to an interim storage facility. All of these operations are conducted remotely with equipment specially designed for these processes. This paper reviews canister processing during the first nine months of radioactive operations at DWPF. The fundamental design consideration for DWPF remote canister processing and handling equipment are discussed as well as interim canister storage

Low and intermediate level radioactive waste processing in plasma reactor

International Nuclear Information System (INIS)

Sauchyn, V.; Khvedchyn, I.; Van Oost, G.

2013-01-01

Methods of low and intermediate level radioactive waste processing comprise: cementation, bituminization, curing in polymer matrices, combustion and pyrolysis. All these methods are limited in their application in the field of chemical, morphological, and aggregate composition of material to be processed. The thermal plasma method is one of the universal methods of RAW processing. The use of electric-arc plasma with mean temperatures 2000 - 8000 K can effectively carry out the destruction of organic compounds into atoms and ions with very high speeds and high degree of conversion. Destruction of complex substances without oxygen leads to a decrease of the volume of exhaust gases and dimension of gas cleaning system. This paper presents the plasma reactor for thermal processing of low and intermediate level radioactive waste of mixed morphology. The equipment realizes plasma-pyrolytic conversion of wastes and results in a conditioned product in a single stage. As a result, the volume of conditioned waste is significantly reduced (more than 10 times). Waste is converted into an environmentally friendly form that suits long-term storage. The leaching rate of macro-components from the vitrified compound is less than 1.10 -7 g/(cm 2 .day). (authors)

Process of liquid radioactive waste treatment in nuclear power plant and development trend

International Nuclear Information System (INIS)

Liu Jiean; Wang Xin; Liu Dan; Zhu Laiye; Chen Bin

2014-01-01

The popular liquid radioactive waste treatment methods in nuclear power plants (NPP) are Chemical precipitation, evaporation, ion exchange, membrane treatment, chemical coagulation and activated carbon absorption and so on. 'Filter + activated carbon absorption (Chemical coagulation) + ion exchange' has a good prospect for development, as its simple process, high decontamination factor, low energy consumption and smaller secondary wastes. Also the process is used in Sanmen and Haiyang Projects. The severe incident in NPP set an even higher demand on liquid radioactive waste treatment. The new type treatment materials, optimization of the existed treatment, combination of treatment and the mobile treatment facility is the development trend in liquid radioactive waste treatment in NPP. (authors)

Process for cleaning radioactively contaminated metal surfaces

International Nuclear Information System (INIS)

Mihram, R.G.; Snyder, G.A.

1975-01-01

A process is described for removing radioactive scale from a ferrous metal surface, including the steps of initially preconditioning the surface by contacting it with an oxidizing solution (such as an aqueous solution of an alkali metal permanganate or hydrogen peroxide), then, after removal or decomposition of the oxidizing solution, the metallic surface is contacted with a cleaning solution which is a mixture of a mineral acid and a complexing agent (such as sulfuric acid and oxalic acid), and which preferably contains a corrosion inhibitor. A final step in the process is the treatment of the spent cleaning solution containing radioactive waste materials in solution by adding a reagent selected from the group consisting of calcium hydroxide or potassium permanganate and an alkali metal hydroxide to thereby form easily recovered metallic compounds containing substantially all of the dissolved metals and radioactivity. (auth)

Technical evaluation of proposed Ukrainian Central Radioactive Waste Processing Facility

International Nuclear Information System (INIS)

Gates, R.; Glukhov, A.; Markowski, F.

1996-06-01

This technical report is a comprehensive evaluation of the proposal by the Ukrainian State Committee on Nuclear Power Utilization to create a central facility for radioactive waste (not spent fuel) processing. The central facility is intended to process liquid and solid radioactive wastes generated from all of the Ukrainian nuclear power plants and the waste generated as a result of Chernobyl 1, 2 and 3 decommissioning efforts. In addition, this report provides general information on the quantity and total activity of radioactive waste in the 30-km Zone and the Sarcophagus from the Chernobyl accident. Processing options are described that may ultimately be used in the long-term disposal of selected 30-km Zone and Sarcophagus wastes. A detailed report on the issues concerning the construction of a Ukrainian Central Radioactive Waste Processing Facility (CRWPF) from the Ukrainian Scientific Research and Design institute for Industrial Technology was obtained and incorporated into this report. This report outlines various processing options, their associated costs and construction schedules, which can be applied to solving the operating and decommissioning radioactive waste management problems in Ukraine. The costs and schedules are best estimates based upon the most current US industry practice and vendor information. This report focuses primarily on the handling and processing of what is defined in the US as low-level radioactive wastes

Radioactive waste processing device

International Nuclear Information System (INIS)

Inaguma, Masahiko; Takahara, Nobuaki; Hara, Satomi.

1996-01-01

In a processing device for filtering laundry liquid wastes and shower drains incorporated with radioactive materials, a fiber filtration device is disposed and an activated carbon filtration device is also disposed subsequent to the fiber filtration device. In addition, a centrifugal dewatering device is disposed for dewatering spent granular activated carbon in the activated carbon filtration device, and a minute filtering device is disposed for filtering the separated dewatering liquid. Filtrates filtered by the minute filtration device are recovered in a collecting tank. Namely, at first, suspended solid materials in laundry liquid wastes and shower drains are captured, and then, ingredients concerning COD are adsorbed in the activated carbon filtration device. The radioactive liquid wastes of spent granular activated carbon in the activated carbon filtration device are reduced by dewatering them by the centrifugal dewatering device, and then the granular activated carbon is subjected to an additional processing. Further, it is separated by filtration using the minute filtration device and removed as cakes. Since the filtrates are recovered to the collecting tank and filtered again, the water quality of the drains is not degraded. (N.H.)

Current status of high level radioactive waste disposal in Japan and foreign countries

International Nuclear Information System (INIS)

Tanaka, Satoru; Tanabe, Hiromi; Inagaki, Yusuke; Ishida, Hisahiro; Kato, Osamu; Kurata, Mitsuyuki; Yamachika, Hidehiko

2002-01-01

At a time point of 2002, there is no country actually disposing high level radioactive wastes into grounds, but in most of countries legislative preparation and practicing agents are carried out and site selection is promoted together with energetic advancement of its R and Ds. As disposal methods of the high level radioactive wastes, various methods such as space disposal, oceanic bottom disposal, ice bed disposal, ground disposal, and so on have been examined. And, a processing technology called partitioning and transmutation technology separating long-lived radionuclides from liquid high level radioactive waste and transmutation into short-lived or harmless radionuclides has also been studied. Here was introduced their wrestling conditions in Japan and main foreign countries, as a special issue of the Current status of high level radioactive waste disposal in Japan and foreign countries'. The high level radioactive wastes (glassification solids or spent nuclear fuels) are wastes always formed by nuclear power generation and establishment of technologies is an important subject for nuclear fuel cycle. (G.K.)

40 CFR 227.30 - High-level radioactive waste.

Science.gov (United States)

2010-07-01

... 40 Protection of Environment 24 2010-07-01 2010-07-01 false High-level radioactive waste. 227.30...-level radioactive waste. High-level radioactive waste means the aqueous waste resulting from the operation of the first cycle solvent extraction system, or equivalent, and the concentrated waste from...

Application of reverse osmosis membrane technology for liquid radioactive waste processing

International Nuclear Information System (INIS)

Zhao Juan

2010-01-01

Liquid radioactive waste (LRW) processing should bear an acceptable level of residual radioactivity for discharge and meet the request of energy saving and waste minimization. Reverse osmosis (RO) membrane technology has been developed as a novel process for LRW processing. Five basic operating parameters of flux, recovery factor, rejection factor, concentration factor and decontamination factor were described, and the latter two parameters were the most important. Concentration factor and decontamination factor should be as high as possible and simultaneously the operating cost for membrane filtration should be low. Technical design considerations for membrane process were discussed and optimized from the aspects of pretreatment, membrane module choice and arrangement and membrane clear out. Application and investigation of RO membrane technology for LRW processing were introduced and it should be noted that the RO membrane technology has been introduced into overseas nuclear power plants for LRW processing and interiorly in the stage of investigation. (authors)

Thermal process for immobilization of radioactive wastes

International Nuclear Information System (INIS)

Brownell, L.E.; Isaacson, R.E.; Kupfer, M.J.; Schulz, W.W.

1971-01-01

The Thermalt process involves an exothermic, thermite-like reaction of aluminum metal with basalt, quartz sand, and radioactive waste. The resulting melt when solidified is a silicious stone-like material that is similar in chemical composition to basalt. The process utilizes low cost ingredients: basalt rock, which occurs naturally in the Hanford region, inexpensive aluminum metal such as aluminum scrap which need not be pure, and the waste which is predominately sodium nitrate salt. The waste itself along with the basalt provides the oxygen necessary for the reaction. The exothermic reaction provides the necessary heat to melt the ingredients thus eliminating the need for external heat sources such as furnaces which are necessary with most other melt methods. The final product is highly stable and essentially nonleachable; leach rates appear as low or lower than other melt products described in the literature. Initial studies indicate the process is effective for both low-level and high-level wastes. (U.S.)

Process for reducing radioactive contamination in phosphogypsum

International Nuclear Information System (INIS)

Palmer, J.W.; Gaynor, J.C.

1983-01-01

In a process for reducing radioactive contamination of phosphogypsum, anhydrite crystals are obtained through dehydration of the phosphogypsum in strong sulfuric acid: a portion of the anhydrite crystals is converted to subtantially radiation free gypsum by crystallizing out on radiation free gypsum seed crystals. These coarse radiation free gypsum crystals are then separated from the small anhydrite crystal relics containing substantially all of the radioactive contamination

High level radioactive waste management facility design criteria

International Nuclear Information System (INIS)

Sheikh, N.A.; Salaymeh, S.R.

1993-01-01

This paper discusses the engineering systems for the structural design of the Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS). At the DWPF, high level radioactive liquids will be mixed with glass particles and heated in a melter. This molten glass will then be poured into stainless steel canisters where it will harden. This process will transform the high level waste into a more stable, manageable substance. This paper discuss the structural design requirements for this unique one of a kind facility. A special emphasis will be concentrated on the design criteria pertaining to earthquake, wind and tornado, and flooding

PROCESSING OF RADIOACTIVE WASTE

Science.gov (United States)

Johnson, B.M. Jr.; Barton, G.B.

1961-11-14

A process for treating radioactive waste solutions prior to disposal is described. A water-soluble phosphate, borate, and/or silicate is added. The solution is sprayed with steam into a space heated from 325 to 400 deg C whereby a powder is formed. The powder is melted and calcined at from 800 to 1000 deg C. Water vapor and gaseous products are separated from the glass formed. (AEC)

Radioactive waste immobilization in protective ceramic forms by the HIP method at high pressures

International Nuclear Information System (INIS)

Sayenko, S.Yu.; Kantsedal, V.P.; Tarasov, R.V.; Starchenko, V.A.; Lyubtsev, R.I.

1993-01-01

Intense research activities have been carried out in recent years at the Kharkov Institute of Physics and Technology (KIPT) to develop the method of hot isostatic pressing (HIP) for immobilizing radioactive (primarily, high-level) wastes. With this method, the radioactive material is immobilized in a matrix under the simultaneous action of high pressures (up to 6,000 atm) and appropriate temperatures. The process has 2 variants: (1) radioactive wastes are treated as powders of oxides resulting from calcination during chemical treatment of spent fuel. In this case the radioactive material enters into the crystalline structure of the immobilized matrix or is distributed in the matrix as a homogeneous mixture; (2) protective barrier layers are pressed on spent fuel rods or their pieces as radioactive wastes, by the HIP method (fuel rod encapsulation in a protective form). Based on numerous results from various studies, the authors suggest that various ceramic compositions should be used as protective materials. Here the authors report two trends of their investigations: (1) development of ecologically clean process equipments for radioactive waste treatment by the HIP method; (2) manufacture of promising protective ceramic compositions and investigation of their physico-mechanical properties

Plasma separation process: Disposal of PSP radioactive wastes

International Nuclear Information System (INIS)

1989-07-01

Radioactive wastes, in the form of natural uranium contaminated scrap hardware and residual materials from decontamination operations, were generated in the PSP facilities in buildings R1 and 106. Based on evaluation of the characteristics of these wastes and the applicable regulations, the various options for the processing and disposal of PSP radioactive wastes were investigated and recommended procedures were developed. The essential features of waste processing included: (1) the solidification of all liquid wastes prior to shipment; (2) cutting of scrap hardware to fit 55-gallon drums and use of inerting agents (diatomaceous earth) to eliminate pyrophoric hazards; and (3) compaction of soft wastes. All PSP radioactive wastes were shipped to the Hanford Site for disposal. As part of the waste disposal process, a detailed plan was formulated for handling and tracking of PSP radioactive wastes, from the point of generation through shipping. In addition, a waste minimization program was implemented to reduce the waste volume or quantity. Included in this document are discussions of the applicable regulations, the types of PSP wastes, the selection of the preferred waste disposal approach and disposal site, the analysis and classification of PSP wastes, the processing and ultimate disposition of PSP wastes, the handling and tracking of PSP wastes, and the implementation of the PSP waste minimization program. 9 refs., 1 fig., 8 tabs

Solidifying processing device for radioactive waste

International Nuclear Information System (INIS)

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

1990-01-01

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

Processing and monitoring liquid, radioactive effluents from the Wiederaufarbeitungsanlage Karlsruhe

International Nuclear Information System (INIS)

Hoehlein, G.; Huppert, K.L.; Winter, M.

1977-01-01

The Wiederaufarbeitungsanlage Karlsruhe (WAK) serves as a demonstration plant for the processing of highly-irradiated uranous oxide. The high active waste concentrates find interim storage at the WAK until they are solidified at a later stage. In contrast to this, the slightly- and the medium-active liquid wastes are transported to the decontamination facility of the Nuclear Research Centre Karlsruhe, where they are immediately processed. These liquid wastes contain about 1 per thousand of the activity inventary of the fuel elements processed. Monitoring of the radioactive waste water of the WAK is carried out by the Nuclear Research Centre's department radiation protection and safety. (orig.) [de

Alternatives evaluation of high activity radioactive wastes disposal

International Nuclear Information System (INIS)

Ciallella, N.R.; Petraitis, E.J.

1989-01-01

Different alternatives considered in the world to be used as barriers to isolate the high level radioactive from the environment wastes produced during the electric energy generation of nuclear origin are presented. Engineering and geologic barriers, are analyzed, considering nuclear fuel cycles with or without plutonium recycling; to that purpose the consideration of elements such as durability and resistance of the various engineering, availability of the fabrication processes, associated radiological impact, geological media apt to be used as geological barrier. Finally, the scopes of the Feasibility Study and Engineering draft are presented for the construction of a repository for high-level radioactive wastes, for the Argentine Nuclear Program needs, which contemplates the construction of six nuclear power plants with a potential installed towards the year 2000 GW( e ), with natural and/or lowly enriched uranium power plants and recycling of plutonium generated in the cycle. (Author) [es

Engineering materials for high level radioactive waste repository

International Nuclear Information System (INIS)

Wen Zhijian

2009-01-01

Radioactive wastes can arise from a wide range of human activities and have different physical and chemical forms with various radioactivity. The high level radioactive wastes (HLW)are characterized by nuclides of very high initial radioactivity, large thermal emissivity and the long life-term. The HLW disposal is highly concerned by the scientists and the public in the world. At present, the deep geological disposal is regarded as the most reasonable and effective way to safely dispose high-level radioactive wastes in the world. The conceptual model of HLW geological disposal in China is based on a multi-barrier system that combines an isolating geological environment with an engineering barrier system(EBS). The engineering materials in EBS include the vitrified HLW, canister, overpack, buffer materials and backfill materials. Referring to progress in the world, this paper presents the function, the requirement for material selection and design, and main scientific projects of R and D of engineering materials in HLW repository. (authors)

Membrane preparation and process development for radioactive waste treatment

Energy Technology Data Exchange (ETDEWEB)

Lee, K. W.; Kim, G. W.; Kim, S. K. [KAERI, Daejeon (Korea, Republic of); and others

2012-01-15

The membrane manufacturing technology with hydrophilic function that can minimize fouling when it applies to the radioactive liquid waste treatment process was developed. Thermodynamic and rheological analysis for polysulfone casting solution containing polyvinylpyrrolidone was performed. On the basis of the results of preparation of the hydrophilic polymer membrane solution, the hollow fiber membrane for radioactive liquid waste treatment was manufactured and its performance analysis was carried out. As a results, it turns out the hydrophilic hollow fiber membrane has more 90 % of flux increment effect and also more 2.5 times fouling reducing effect than one prepared with only polysulfone. In addition, as investigating the separation property of radioactive liquid waste for the electrofilteration membrane process, a proper range for application of radioactive liquid wastes was established through the thorough electrofiltration analysis of various wastes containing metal salt, surfactants and oil.

Membrane preparation and process development for radioactive waste treatment

International Nuclear Information System (INIS)

Lee, K. W.; Kim, G. W.; Kim, S. K.

2012-01-01

The membrane manufacturing technology with hydrophilic function that can minimize fouling when it applies to the radioactive liquid waste treatment process was developed. Thermodynamic and rheological analysis for polysulfone casting solution containing polyvinylpyrrolidone was performed. On the basis of the results of preparation of the hydrophilic polymer membrane solution, the hollow fiber membrane for radioactive liquid waste treatment was manufactured and its performance analysis was carried out. As a results, it turns out the hydrophilic hollow fiber membrane has more 90 % of flux increment effect and also more 2.5 times fouling reducing effect than one prepared with only polysulfone. In addition, as investigating the separation property of radioactive liquid waste for the electrofilteration membrane process, a proper range for application of radioactive liquid wastes was established through the thorough electrofiltration analysis of various wastes containing metal salt, surfactants and oil

Radioactive liquid wastes processing device

International Nuclear Information System (INIS)

Sauda, Kenzo; Koshiba, Yukihiko; Yagi, Takuro; Yamazaki, Hideki.

1985-01-01

Purpose: To carry out optimum photooxidizing procession following after the fluctuation in the density of organic materials in radioactive liquid wastes to thereby realize automatic remote procession. Constitution: A reaction tank is equipped with an ultraviolet lamp and an ozone dispersing means for the oxidizing treatment of organic materials in liquid wastes under the irradiation of UV rays. There are also provided organic material density measuring devices to the inlet and outlet of the reaction tank, and a control device for controlling the UV lamp power adjusting depending on the measured density. The output of the UV lamp is most conveniently adjusted by changing the applied voltage. The liquid wastes in which the radioactivity dose is reduced to a predetermined level are returned to the reaction tank by the operation of a switching valve for reprocession. The amount of the liquid wastes at the inlet is controlled depending on the measured ozone density by the adjusting valve. In this way, the amount of organic materials to be subjected to photolysis can be kept within a certain limit. (Kamimura, M.)

Material chemistry challenges in vitrification of high level radioactive waste

International Nuclear Information System (INIS)

Kaushik, C.P.

2008-01-01

Full text: Nuclear technology with an affective environmental management plan and focused attention on safety measures is a much cleaner source of electricity generation as compared to other sources. With this perspective, India has undertaken nuclear energy program to share substantial part of future need of power. Safe containment and isolation of nuclear waste from human environment is an indispensable part of this programme. Majority of radioactivity in the entire nuclear fuel cycle is high level radioactive liquid waste (HLW), which is getting generated during reprocessing of spent nuclear fuels. A three stage strategy for management of HLW has been adopted in India. This involves (i) immobilization of waste oxides in stable and inert solid matrices, (ii) interim retrievable storage of the conditioned waste product under continuous cooling and (iii) disposal in deep geological formation. Borosilicate glass matrix has been adopted in India for immobilization of HLW. Material issue are very important during the entire process of waste immobilization. Performance of the materials used in nuclear waste management determines its safety/hazards. Material chemistry therefore has a significant bearing on immobilization science and its technological development for management of HLW. The choice of suitable waste form to deploy for nuclear waste immobilization is difficult decision and the durability of the conditioned product is not the sole criterion. In any immobilization process, where radioactive materials are involved, the process and operational conditions play an important role in final selection of a suitable glass formulation. In remotely operated vitrification process, study of chemistry of materials like glass, melter, materials of construction of other equipment under high temperature and hostile corrosive condition assume significance for safe and un-interrupted vitrification of radioactive to ensure its isolation waste from human environment. The present

Quality assurance in processing radioactive waste for land disposal

International Nuclear Information System (INIS)

1984-01-01

To provide the appropriate assurances as to the quality of processed radioactive waste it is necessary to consider the complete range of activities involved in the formation and operation of a radioactive waste processing facility. To this end, an outline has been given to the individual elements which should be addressed in quality assurance proposals to the authorising Departments. In general terms, the quality checks on product material should be aimed at demonstrating that the radioactive waste product is what was agreed at the time of process approval. In addition, at the discretion of the authorising Departments, the waste processor will be required to confirm that the product meets any specific acceptance criteria such as the capability to retain the immobilised radionuclides when in contact with water. (author)

Method of controlling radioactive waste processing systems

International Nuclear Information System (INIS)

Mikawa, Hiroji; Sato, Takao.

1981-01-01

Purpose: To minimize the pellet production amount, maximize the working life of a solidifying device and maintaining the mechanical strength of pellets to a predetermined value irrespective of the type and the cycle of occurrence of the secondary waste in the secondary waste solidifying device for radioactive waste processing systems in nuclear power plants. Method: Forecasting periods for the type, production amount and radioactivity level of the secondary wastes are determined in input/output devices connected to a control system and resulted signals are sent to computing elements. The computing elements forecast the production amount of regenerated liquid wastes after predetermined days based on the running conditions of a condensate desalter and the production amounts of filter sludges and liquid resin wastes after predetermined days based on the liquid waste processing amount or the like in a processing device respectively. Then, the mass balance between the type and the amount of the secondary wastes presently stored in a tank are calculated and the composition and concentration for the processing liquid are set so as to obtain predetermined values for the strength of pellets that can be dried to solidify, the working life of the solidifying device itself and the radioactivity level of the pellets. Thereafter, the running conditions for the solidifying device are determined so as to maximize the working life of the solidifying device. (Horiuchi, T.)

Development of cermets for high-level radioactive waste fixation

International Nuclear Information System (INIS)

Aaron, W.S.; Quinby, T.C.; Kobisk, E.H.

1979-01-01

A method is currently under development for the solidification and fixation of commercial and defense high-level radioactive wastes in the form of ceramic particles encapsulated by metal, i.e., a cermet. The chemical and physical processing techniques which have been developed and the properties of the resulting cermet bodies are described in this paper. These cermets have the advantages of high thermal conductivity and low leach rates

Remote sampling and analysis of highly radioactive samples in shielded boxes

International Nuclear Information System (INIS)

Kirpikov, D.A.; Miroshnichenko, I.V.; Pykhteev, O.Yu.

2010-01-01

equipped with radiation resistant video cameras provided for additional visual inspection. All the acquired information including visual images is sent to the workstation of an operator who controls handling of highly radioactive samples. Authorized participants of chemical control are able to supervise the sampling procedure and process the obtained results via local area network owing to inclusion of the workstation in the common data acquisition, processing and storage system. The paper describes the developed equipment and solutions and discusses the issues related to development of the local computer-aided chemical control system. (author)

Radioactive waste processing method and device

International Nuclear Information System (INIS)

Ozaki, Shigeru; Tateyama, Shinji.

1998-01-01

A powdery activated carbon is charged to radioactive liquid wastes to form a mixed slurry. The slurry is subjected to solid/liquid separation, and a high-molecular water absorbent is charged to the separated activated carbon sludge wastes to process them while stirring. The high-molecular water absorbent comprises a graft polymer of starch and acrylonitrile or a cross-linked polymer of sodium acrylate and a cross-linking agent. The high-molecular water absorbing agent is previously charged to a vessel for containing the wasted active carbon sludges. The device of the present invention comprises a filtration device for solid/liquid separation of the mixed slurry, a sludge-containing vessel, a device for charging the high-molecular water absorbent and a sludge stirring device. The device of charging the high-molecular water absorbent comprises a plurality of weighing devices for weighing the change of the weight of the charged products and a conveyor for transferring the sludge-containing vessels. With such a constitution, stable sludge can be obtained, and activated carbon sludge wastes can be burnt without crushing them. (T.M.)

Defense waste processing facility radioactive operations. Part 1 - operating experience

International Nuclear Information System (INIS)

Little, D.B.; Gee, J.T.; Barnes, W.M.

1997-01-01

The Savannah River Site's Defense Waste Processing Facility (DWPF) near Aiken, SC is the nation's first and the world's largest vitrification facility. Following a ten year construction program and a 3 year non-radioactive test program, DWPF began radioactive operations in March 1996. This paper presents the results of the first 9 months of radioactive operations. Topics include: operations of the remote processing equipment reliability, and decontamination facilities for the remote processing equipment. Key equipment discussed includes process pumps, telerobotic manipulators, infrared camera, Holledge trademark level gauges and in-cell (remote) cranes. Information is presented regarding equipment at the conclusion of the DWPF test program it also discussed, with special emphasis on agitator blades and cooling/heating coil wear. 3 refs., 4 figs

Design of Biochemical Oxidation Process Engineering Unit for Treatment of Organic Radioactive Liquid Waste

International Nuclear Information System (INIS)

Zainus Salimin; Endang Nuraeni; Mirawaty; Tarigan, Cerdas

2010-01-01

Organic radioactive liquid waste from nuclear industry consist of detergent waste from nuclear laundry, 30% TBP-kerosene solvent waste from purification or recovery of uranium from process failure of nuclear fuel fabrication, and solvent waste containing D 2 EHPA, TOPO, and kerosene from purification of phosphoric acid. The waste is dangerous and toxic matter having low pH, high COD and BOD, and also low radioactivity. Biochemical oxidation process is the effective method for detoxification of organic waste and decontamination of radionuclide by bio sorption. The result process are sludges and non radioactive supernatant. The existing treatment facilities radioactive waste in Serpong can not use for treatment of that’s organics waste. Dio chemical oxidation process engineering unit for continuous treatment of organic radioactive liquid waste on the capacity of 1.6 L/h has been designed and constructed the equipment of process unit consist of storage tank of 100 L capacity for nutrition solution, 2 storage tanks of 100 L capacity per each for liquid waste, reactor oxidation of 120 L, settling tank of 50 L capacity storage tank of 55 L capacity for sludge, storage tank of 50 capacity for supernatant. Solution on the reactor R-01 are added by bacteria, nutrition and aeration using two difference aerators until biochemical oxidation occurs. The sludge from reactor of R-01 are recirculated to the settling tank of R-02 and on the its reverse operation biological sludge will be settled, and supernatant will be overflow. (author)

Production of high intensity radioactive beams

International Nuclear Information System (INIS)

Nitschke, J.M.

1990-04-01

The production of radioactive nuclear beams world-wide is reviewed. The projectile fragmentation and the ISOL approaches are discussed in detail, and the luminosity parameter is used throughout to compare different production methods. In the ISOL approach a thin and a thick target option are distinguished. The role of storage rings in radioactive beam research is evaluated. It is concluded that radioactive beams produced by the projectile fragmentation and the ISOL methods have complementary characteristics and can serve to answer different scientific questions. The decision which kind of facility to build has to depend on the significance and breadth of these questions. Finally a facility for producing a high intensity radioactive beams near the Coulomb barrier is proposed, with an expected luminosity of ∼10 39 cm -2 s -1 , which would yield radioactive beams in excess of 10 11 s -1 . 9 refs., 3 figs., 7 tabs

Processing radioactive wastes using membrane (UF/HF/RO) systems

International Nuclear Information System (INIS)

Doyle, R.D.

1988-01-01

Over the years many technologies have been utilized to process low level radioactive waste streams generated by the nuclear industry, including: demineralization, evaporation, reverse osmosis and filtration. In the early 1980's interest was generated in membrane technologies and their application to radioactive wastes. This interest was generated based on the capabilities shown by membrane systems in non-radioactive environments and the promise that reverse osmosis systems showed in early testing with radioactive wastes. Membrane technologies have developed from the early development of reverse osmosis system to also include specifically designed membranes for ultrafiltration and hyperfiltration applications

A very high energy imaging for radioactive wastes processing; Une imagerie haute energie pour la gestion des dechets radioactifs

Energy Technology Data Exchange (ETDEWEB)

Moulin, V. [CEA Grenoble, Lab. d' Electronique et de Technologie de l' Informatique (DRT/DSIS/SSBS/LETI), 38 (France); Pettier, J.L. [CEA Cadarache, Dir. de l' Energie Nucleaire (DEN/DTN/SMTM), 13 - Saint-Paul-lez-Durance (France)

2004-07-01

The X imaging occurs at a lot of steps of the radioactive wastes processing: selection for conditioning, physical characterization with a view to radiological characterization, quality control of the product before storage, transport or disposal. Size and volume of the objects considered here necessitate to work with very high energy systems. Here is shown, through some examples, in which conditions this X imaging is carried out as well as the contribution of the obtained images. (O.M.)

Radioactive waste management

International Nuclear Information System (INIS)

Blomek, D.

1980-01-01

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

Radioactive wastes processing and disposing container

International Nuclear Information System (INIS)

Wada, Jiro; Kato, Hiroaki.

1987-01-01

Purpose: To obtain a processing and disposing container at low level radioactive wastes, excellent in corrosion and water resistance, as well as impact shock resistance for the retrieval storage over a long period of time. Constitution: The container is constituted with sands and pebbles as aggregates and glass fiber-added unsaturated polyester resins as binders. The container may entirely be formed with such material or only the entire inner surface may be formed with the material as liners. A container having excellent resistance to water, chemicals, freezing or melting, whether impact shock, etc. can be obtained, thereby enabling retrieval storage for radioactive wastes at the optimum low level. (Takahashi, M.)

Final repositories for high-level radioactive waste; Endlagerung hochradioaktiver Abfaelle

Energy Technology Data Exchange (ETDEWEB)

NONE

2015-10-15

The brochure on final repositories for high-level radioactive waste covers the following issues: What is the origin of radioactive wastes? How large are the waste amounts? What is going to happen with the wastes? What is the solution for the Waste disposal? A new site search is started. Safety requirements for the final disposal of high-level radioactive wastes. Comparison of host rocks. Who is responsible and who will pay? Final disposal of high-level radioactive wastes worldwide. Short summary: History of the search for a final repository for high-level radioactive wastes in Germany.

Boiling water reactor liquid radioactive waste processing system

International Nuclear Information System (INIS)

Anon.

1977-01-01

The standard sets forth minimum design, construction and performance requirements with due consideration for operation of the liquid radioactive waste processing system for boiling water reactor plants for routine operation including design basis fuel leakage and design basis occurrences. For the purpose of this standard, the liquid radioactive waste processing system begins at the interfaces with the reactor coolant pressure boundary, at the interface valve(s) in lines from other systems and at those sumps and floor drains provided for liquid waste with the potential of containing radioactive material. The system terminates at the point of controlled discharge to the environment, at the point of interface with the waste solidification system and at the point of recycle back to storage for reuse. The standard does not include the reactor coolant clean-up system, fuel pool clean-up system, sanitary waste system, any nonaqueous liquid system or controlled area storm drains

Biochemical process of low level radioactive liquid simulation waste containing detergent

International Nuclear Information System (INIS)

Kundari, Noor Anis; Putra, Sugili; Mukaromah, Umi

2015-01-01

Research of biochemical process of low level radioactive liquid waste containing detergent has been done. Thse organic liquid wastes are generated in nuclear facilities such as from laundry. The wastes that are cotegorized as hazard and poison materials are also radioactive. It must be treated properly by detoxification of the hazard and decontamination of the radionuclides to ensure that the disposal of the waste meets the requirement of standard quality of water. This research was intended to determine decontamination factor and separation efficiensies, its kinetics law, and to produce a supernatant that ensured the environmental quality standard. The radioactive element in the waste was thorium with activity of 5.10 −5 Ci/m 3 . The radioactive liquid waste which were generated in simulation plant contains detergents that was further processed by aerobic biochemical process using SGB 103 bacteria in a batch reactor equipped with aerators. Two different concentration of samples were processed and analyzed for 212 hours and 183 hours respectively at a room temperature. The product of this process is a liquid phase called as supernatant and solid phase material called sludge. The chemical oxygen demand (COD), biological oxygen demand (BOD), suspended solid (SS), and its alpha activity were analyzed. The results show that the decontamination factor and the separation efficiency of the lower concentration samples are higher compared to the samples with high concentration. Regarding the decontamination factor, the result for 212 hours processing of waste with detergent concentration of 1.496 g/L was 3.496 times, whereas at the detergent concentration of 0.748 g/L was 15.305 times for 183 hours processing. In case of the separation efficiency, the results for both samples were 71.396% and 93.465% respectively. The Bacterial growth kinetics equation follow Monod’s model and the decreasing of COD and BOD were first order with the rate constant of 0.01 hour −1

Biochemical process of low level radioactive liquid simulation waste containing detergent

Energy Technology Data Exchange (ETDEWEB)

Kundari, Noor Anis, E-mail: nooranis@batan.go.id; Putra, Sugili; Mukaromah, Umi [Sekolah Tinggi Teknologi Nuklir – Badan Tenaga Nuklir Nasional Jl. Babarsari P.O. BOX 6101 YKBB Yogyakarta 55281 Telp : (0274) 48085, 489716, Fax : (0274) 489715 (Indonesia)

2015-12-29

Research of biochemical process of low level radioactive liquid waste containing detergent has been done. Thse organic liquid wastes are generated in nuclear facilities such as from laundry. The wastes that are cotegorized as hazard and poison materials are also radioactive. It must be treated properly by detoxification of the hazard and decontamination of the radionuclides to ensure that the disposal of the waste meets the requirement of standard quality of water. This research was intended to determine decontamination factor and separation efficiensies, its kinetics law, and to produce a supernatant that ensured the environmental quality standard. The radioactive element in the waste was thorium with activity of 5.10{sup −5} Ci/m{sup 3}. The radioactive liquid waste which were generated in simulation plant contains detergents that was further processed by aerobic biochemical process using SGB 103 bacteria in a batch reactor equipped with aerators. Two different concentration of samples were processed and analyzed for 212 hours and 183 hours respectively at a room temperature. The product of this process is a liquid phase called as supernatant and solid phase material called sludge. The chemical oxygen demand (COD), biological oxygen demand (BOD), suspended solid (SS), and its alpha activity were analyzed. The results show that the decontamination factor and the separation efficiency of the lower concentration samples are higher compared to the samples with high concentration. Regarding the decontamination factor, the result for 212 hours processing of waste with detergent concentration of 1.496 g/L was 3.496 times, whereas at the detergent concentration of 0.748 g/L was 15.305 times for 183 hours processing. In case of the separation efficiency, the results for both samples were 71.396% and 93.465% respectively. The Bacterial growth kinetics equation follow Monod’s model and the decreasing of COD and BOD were first order with the rate constant of 0

Biochemical process of low level radioactive liquid simulation waste containing detergent

Science.gov (United States)

Kundari, Noor Anis; Putra, Sugili; Mukaromah, Umi

2015-12-01

Research of biochemical process of low level radioactive liquid waste containing detergent has been done. Thse organic liquid wastes are generated in nuclear facilities such as from laundry. The wastes that are cotegorized as hazard and poison materials are also radioactive. It must be treated properly by detoxification of the hazard and decontamination of the radionuclides to ensure that the disposal of the waste meets the requirement of standard quality of water. This research was intended to determine decontamination factor and separation efficiensies, its kinetics law, and to produce a supernatant that ensured the environmental quality standard. The radioactive element in the waste was thorium with activity of 5.10-5 Ci/m3. The radioactive liquid waste which were generated in simulation plant contains detergents that was further processed by aerobic biochemical process using SGB 103 bacteria in a batch reactor equipped with aerators. Two different concentration of samples were processed and analyzed for 212 hours and 183 hours respectively at a room temperature. The product of this process is a liquid phase called as supernatant and solid phase material called sludge. The chemical oxygen demand (COD), biological oxygen demand (BOD), suspended solid (SS), and its alpha activity were analyzed. The results show that the decontamination factor and the separation efficiency of the lower concentration samples are higher compared to the samples with high concentration. Regarding the decontamination factor, the result for 212 hours processing of waste with detergent concentration of 1.496 g/L was 3.496 times, whereas at the detergent concentration of 0.748 g/L was 15.305 times for 183 hours processing. In case of the separation efficiency, the results for both samples were 71.396% and 93.465% respectively. The Bacterial growth kinetics equation follow Monod's model and the decreasing of COD and BOD were first order with the rate constant of 0.01 hour-1.

Definition and preparation of glassy matrices by innovating processes to confine radioactive wastes and industrial toxic materials

International Nuclear Information System (INIS)

Moncouyoux, J.P.

1995-01-01

The confinement by vitrification of high-level radioactive wastes is studied in the CEA for fifteen years. These studies have lead to the preparation of glassy matrices by innovating processes. These processes can be applied to non-radioactive toxic materials treatment too. In this work are more particularly described the glassy matrix long-dated behaviour and the different vitrification processes used (by direct induction in cold crucible, by transferred arc plasma). (O.L.). 1 tab

Process for processing and conditioning radioactive effluents of low and medium activity

International Nuclear Information System (INIS)

Taponier, Jean; Pierlas, Rene.

1979-01-01

Preferably continuous process for processing radioactive effluents of low and medium activity, comprising an effluent pre-treatment: precipitation of radioactive compounds to form a stable suspension that can be concentrated. Then a mix is made of 0.6 to 2 parts of cement by weight for one part by weight of suspension, from 0.5 to 5% by weight, in relation to the cement, of asbestos fibre and, if necessary, added water for the cement to set, this suspension containing from 15 to 75% by weight of dry extract and a suspension agent. The homogeneous mix achieved is poured into a container [fr

Potential of membrane processes in management of radioactive liquid waste

International Nuclear Information System (INIS)

Kumar, Surender; Jain, Savita; Raj, Kanwar

2010-01-01

Various categories of radioactive liquid waste are generated during operations and maintenance of nuclear installations. The potential of membrane processes for the treatment of low-level radioactive liquids is discussed in this paper

Industrial-Scale Processes For Stabilizing Radioactively Contaminated Mercury Wastes

International Nuclear Information System (INIS)

Broderick, T. E.; Grondin, R.

2003-01-01

This paper describes two industrial-scaled processes now being used to treat two problematic mercury waste categories: elemental mercury contaminated with radionuclides and radioactive solid wastes containing greater than 260-ppm mercury. The stabilization processes were developed by ADA Technologies, Inc., an environmental control and process development company in Littleton, Colorado. Perma-Fix Environmental Services has licensed the liquid elemental mercury stabilization process to treat radioactive mercury from Los Alamos National Laboratory and other DOE sites. ADA and Perma-Fix also cooperated to apply the >260-ppm mercury treatment technology to a storm sewer sediment waste collected from the Y-12 complex in Oak Ridge, TN

Radioactive demonstration of the ''late wash'' Precipitate Hydrolysis Process

International Nuclear Information System (INIS)

Bibler, N.E.; Ferrara, D.M.; Ha, B.C.

1992-01-01

This report presents results of the radioactive demonstration of the DWPF Precipitate Hydrolysis Process as it would occur in the ''late wash'' flowsheet in the absence of hydroxylamine nitrate. Radioactive precipitate containing Cs-137 from the April, 1983, in-tank precipitation demonstration in Tank 48 was used for these tests

Foaming and Antifoaming and Gas Entrainment in Radioactive Waste Pretreatment and Immobilization Processes. Final Report

International Nuclear Information System (INIS)

Wasan, Darsh T.

2007-01-01

The Savannah River Site (SRS) and Hanford site are in the process of stabilizing millions of gallons of radioactive waste slurries remaining from production of nuclear materials for the Department of Energy (DOE). The Defense Waste Processing Facility (DWPF) at SRS is currently vitrifying the waste in borosilicate glass, while the facilities at the Hanford site are in the construction phase. Both processes utilize slurry-fed joule-heated melters to vitrify the waste slurries. The DWPF has experienced difficulty during operations. The cause of the operational problems has been attributed to foaming, gas entrainment and the rheological properties of the process slurries. The rheological properties of the waste slurries limit the total solids content that can be processed by the remote equipment during the pretreatment and meter feed processes. Highly viscous material can lead to air entrainment during agitation and difficulties with pump operations. Excessive foaming in waste evaporators can cause carryover of radionuclides and non-radioactive waste to the condensate system. Experimental and theoretical investigations of the surface phenomena, suspension rheology and bubble generation of interactions that lead to foaming and air entrainment problems in the DOE High Level and Low Activity Radioactive Waste separation and immobilization processes were pursued under this project. The first major task accomplished in the grant proposal involved development of a theoretical model of the phenomenon of foaming in a three-phase gas-liquid-solid slurry system. This work was presented in a recently completed Ph.D. thesis (9). The second major task involved the investigation of the inter-particle interaction and microstructure formation in a model slurry by the batch sedimentation method. Both experiments and modeling studies were carried out. The results were presented in a recently completed Ph.D. thesis. The third task involved the use of laser confocal microscopy to study

Defense Waste Processing Facility radioactive operations -- Part 2, Glass making

International Nuclear Information System (INIS)

Carter, J.T.; Rueter, K.J.; Ray, J.W.; Hodoh, O.

1996-01-01

The Savannah River Site's Defense Waste Processing Facility (DWPF) near Aiken, SC is the nation's first and world's largest vitrification facility. Following a ten year construction period and nearly 3 year non-radioactive test program, the DWPF began radioactive operations in March, 1996. The results of the first 8 months of radioactive operations are presented. Topics include facility production from waste preparation batching to canister filling

Process for reducing radioactive contamination in waste product gypsum

International Nuclear Information System (INIS)

Lange, P.H. Jr.

1979-01-01

A process is described for reducing the radioactive contamination in waste product gypsum in which waste product gypsum is reacted with a dilute sulfuric acid containing barium sulfate to form an acid slurry at an elevated temperature, the slurry is preferably cooled, the acid component is separated from the solid, and the resulting solid is separated into a fine fraction and a coarse fraction. The fine fraction predominates in barium sulfate and radioactive contamination. The coarse fraction predominates in a purified gypsum product of reduced radioactive contamination

Study of the Treatment of the Liquid Radioactive Waste Nong Son Uranium Ore Processing

International Nuclear Information System (INIS)

Nguyen Ba Tien; Trinh Giang Huong; Luu Cao Nguyen; Harvey, L.K.; Tran Van Quy

2011-01-01

Liquid waste from Nong Son uranium ore processing is treated with concentrated acid, agglomerated, leached, run through ion exchange and then treated with H 2 O 2 to precipitate yellowcake. The liquid radioactive waste has a pH of 1.86 and a high content of radioactive elements, such as: [U] 143.898 ppm and [Th] = 7.967 ppm. In addition, this waste contains many polluted chemical elements with high content, such as arsenic, mercury, aluminum, iron, zinc, magnesium, manganese and nickel. The application of the general method as one stage precipitation or precipitation in coordination with BaCl 2 is not effective. These methods generated a large amount of sludge with poor settling characteristics. The volume of final treated waste was large. This paper introduces the investigation of the treatment of this liquid radioactive waste by the method of two stage of precipitation in association with polyaluminicloride (PAC) and polymer. The impact of factors: pH, neutralizing agents, quantity of PAC and polymer to effect precipitation and improve the settling characteristics during processing was studied. The results showed that the processing of liquid radioactive waste treatment through two stages: first stage at pH = 3 and the second stage at pH = 8.0 with limited PAC and polymer (A 101) resulted in significant reduced volume of the treated waste. The discharged liquid satisfied the requirement of the National Technical Regulation on Industrial Waste Water (QCVN 24:2009). (author)

Southern routes for high-level radioactive waste: Agencies, contacts, and designations

International Nuclear Information System (INIS)

1991-05-01

The Southern Routes for High-Level Radioactive Waste: Agencies, Contacts and Designations is a compendium of sixteen southern states' routing programs for the transportation of high-level radioactive materials. The report identifies the state-designated routing agencies as defined under 49 Code of Federal Regulations (CFR) Part 171 and provides a reference to the source and scope of the agencies' rulemaking authority. Additionally, the state agency and contact designated by the state's governor to receive advance notification and shipment routing information under 10 CFR Parts 71 and 73 are also listed. This report also examines alternative route designations made by southern states and the lessons that were learned from the designation process

Radioactive gaseous waste processing device

International Nuclear Information System (INIS)

Kishi, Tadao.

1990-01-01

The present invention concerns a radioactive gaseous waste processing device used in BWR power plants. A heater is disposed to the lower portion of a dryer for dehydrating radioactive off gases. Further, a thermometer is disposed to a coolant return pipeway on the exit side of the cooling portion of the dryer and signals sent from the thermometer are inputted to an automatic temperature controller. If the load on the dryer is reduced, the value of the thermometer is lowered than a set value, then an output signal corresponding to the change is supplied from the automatic temperature controller to the heater to forcively apply loads to the dryer. Therefore, defrosting can be conducted completely without operating a refrigerator, and the refrigerator can be maintained under a constant load by applying a dummy load when the load in the dryer is reduced. (I.N.)

Disposal of radioactive waste from mining and processing of mineral sands

International Nuclear Information System (INIS)

Hartley, B.M.

1993-01-01

All mineral sands products contain the naturally radioactive elements uranium and thorium and their daughters. The activity levels in the different minerals can vary widely and in the un mined state are frequently widely dispersed and add to the natural background radiation levels. Following mining, the minerals are concentrated to a stage where radiation levels can present an occupational hazard and disposal of waste can result in radiation doses in excess of the public limit. Chemical processing can release radioactive daughters, particularly radium, leading to the possibility of dispersal and resulting in widespread exposure of the public. The activity concentration in the waste can vary widely and different disposal options appropriate to the level of activity in the waste are needed. Disposal methods can range from dilution and dispersal of the material into the mine site, for untreated mine tailings, to off site disposal in custom built and engineered waste disposal facilities, for waste with high radionuclide content. The range of options for disposal of radioactive waste from mineral sands mining and processing is examined and the principles for deciding on the appropriate disposal option are discussed. The range of activities of waste from different downstream processing paths are identified and a simplified method of identifying potential waste disposal paths is suggested. 15 refs., 4 tabs

Feasibility of large volume casting cementation process for intermediate level radioactive waste

International Nuclear Information System (INIS)

Chen Zhuying; Chen Baisong; Zeng Jishu; Yu Chengze

1988-01-01

The recent tendency of radioactive waste treatment and disposal both in China and abroad is reviewed. The feasibility of the large volume casting cementation process for treating and disposing the intermediate level radioactive waste from spent fuel reprocessing plant in shallow land is assessed on the basis of the analyses of the experimental results (such as formulation study, solidified radioactive waste properties measurement ect.). It can be concluded large volume casting cementation process is a promising, safe and economic process. It is feasible to dispose the intermediate level radioactive waste from reprocessing plant it the disposal site chosen has resonable geological and geographical conditions and some additional effective protection means are taken

Radioactive decontamination apparatus and process

International Nuclear Information System (INIS)

Jackson, O.L.

1983-01-01

Apparatus for removing radioactive contamination from metal objects is disclosed, consisting of three of three separate pieces. The first is an electro- polishing tank, pump and filter assembly, ventilation duct and filter assembly, and DC power supply. The second is a rinse tank and a pump and filter assembly therefor. The third is a divot crane. The electro-polishing tank assembly and the rinse tank assembly are each separately mounted on pallets to facilitate moving. The filter systems of the electro-polishing tank and the rinse tank are designed to remove the radioactive contamination from the fluids in those tanks. Heavy items or highly contaminated items are handled with the divot crane constructed of stainless steel. The electro- polishing tank and the rinse tank are also made of stainless steel. The ventilation system on the electro- polishing tank exhausts acid fumes resulting from the tank heaters and the electro-polishing process. Inside the electro-polishing tank are two swinging arms that carry two stainless steel probes that hang down in the electrolyte fluid. These negative DC probes and are electrically isolated from the tank and the rest of the system. Across the top center of the tank is a copper pipe, which is also electrically isolated from the tank. This is the positive side of the DC system. To decontaminate a metal object, it is suspended from the positive copper pipe, with good electrical contact, into the electrolyte fluid. The negative probes are then moved on their swinging arms to a close proximity to the object being decontaminated, without making contact

Proposed classification scheme for high-level and other radioactive wastes

International Nuclear Information System (INIS)

Kocher, D.C.; Croff, A.G.

1986-01-01

The Nuclear Waste Policy Act (NWPA) of 1982 defines high-level (radioactive) waste (HLW) as (A) the highly radioactive material resulting from the reprocessing of spent nuclear fuel...that contains fission products in sufficient concentrations; and (B) other highly radioactive material that the Commission...determines...requires permanent isolation. This paper presents a generally applicable quantitative definition of HLW that addresses the description in paragraph B. The approach also results in definitions of other wastes classes, i.e., transuranic (TRU) and low-level waste (LLW). The basic waste classification scheme that results from the quantitative definitions of highly radioactive and requires permanent isolation is depicted. The concentrations of radionuclides that correspond to these two boundaries, and that may be used to classify radioactive wastes, are given

What are Spent Nuclear Fuel and High-Level Radioactive Waste?

International Nuclear Information System (INIS)

2002-01-01

Spent nuclear fuel and high-level radioactive waste are materials from nuclear power plants and government defense programs. These materials contain highly radioactive elements, such as cesium, strontium, technetium, and neptunium. Some of these elements will remain radioactive for a few years, while others will be radioactive for millions of years. Exposure to such radioactive materials can cause human health problems. Scientists worldwide agree that the safest way to manage these materials is to dispose of them deep underground in what is called a geologic repository

Microbial processes in radioactive waste repository

International Nuclear Information System (INIS)

Gazso, L.; Farkas-Galgoczi, G.; Diosi, G.

2002-01-01

Microbial processes could potentially affect the performance of a radioactive waste disposal system and related factors that could have an influence on the mobility of radionuclides are outlined. Analytical methods, including sampling of water, rock and surface swabs from a potential disposal site, are described and the quantitative as well as qualitative experimental results obtained are given. Although the results contribute to an understanding of the impact of microbial processes on deep geological disposal of nuclear waste, there is not yet sufficient information for a model which will predict the consequences of these processes. (author)

Radioactive waste processing method

International Nuclear Information System (INIS)

Ando, Ken-ichi; Kawamura, Hideki; Takeuchi, Kunifumi.

1997-01-01

Base rock is dug in a substantially cylindrical shape, bentonite blocks in an amount for a predetermined lift are disposed on the inner side of the dug wall surfaces. Concrete blocks constituting a structure of an underground silo are disposed at the inner side. Barrier blocks are disposed to the inner side thereof, and vessels incorporated with radioactive wastes are disposed to the inner side. The bentonite disposed to the inner side of the dug wall surfaces, the concrete structure of the underground silo and the barrier members are divided in the vertical direction into a plurality of blocks, and these blocks are stacked successively from the lowermost layer together with the containing vessels of the radioactive wastes, and after stacking them to a predetermined height, a filler is filled up to the circumference of the vessels. With such a constitution, the underground silo is not fallen down or vibrated even upon occurrence of an earthquake. In addition, bending stresses are scarcely caused thereby making reinforcement of iron reinforcing materials unnecessary. Accordingly, the sealing performance is improved, and processing cost is reduced. (T.M.)

Disposal of high-level radioactive waste

International Nuclear Information System (INIS)

Glasby, G.P.

1977-01-01

Although controversy surrounding the possible introduction of nuclear power into New Zealand has raised many points including radiation hazards, reactor safety, capital costs, sources of uranium and earthquake risks on the one hand versus energy conservation and alternative sources of energy on the other, one problem remains paramount and is of global significance - the storage and dumping of the high-level radioactive wastes of the reactor core. The generation of abundant supplies of energy now in return for the storage of these long-lived highly radioactive wastes has been dubbed the so-called Faustian bargain. This article discusses the growth of the nuclear industry and its implications to high-level waste disposal particularly in the deep-sea bed. (auth.)

Processing of radioactive waste solutions in a vacuum evaporator-crystallizer

International Nuclear Information System (INIS)

Petrie, J.C.; Donovan, R.I.; Van der Cook, R.E.; Christensen, W.R.

1975-01-01

Results of the first 18 months' operation of Hanford's vacuum evaporator-crystallizer are reported. This process reduces the volume of radioactive waste solutions and simultaneously converts the waste to a less mobile salt cake. The evaporator-crystallizer is operating at better than design production rates and has reduced the volume of radioactive wastes by more than 15 million gallons. A process description, plant performance data, mechanical difficulties, and future operating plans are discussed. Also discussed is a computer model of the evaporator-crystallizer process

Automatized material and radioactivity flow control tool in decommissioning process

International Nuclear Information System (INIS)

Rehak, I.; Vasko, M.; Daniska, V.; Schultz, O.

2009-01-01

In this presentation the automatized material and radioactivity flow control tool in decommissioning process is discussed. It is concluded that: computer simulation of the decommissioning process is one of the important attributes of computer code Omega; one of the basic tools of computer optimisation of decommissioning waste processing are the tools of integral material and radioactivity flow; all the calculated parameters of materials are stored in each point of calculation process and they can be viewed; computer code Omega represents opened modular system, which can be improved; improvement of the module of optimisation of decommissioning waste processing will be performed in the frame of improvement of material procedures and scenarios.

Liquid radioactive waste processing system for pressurized water reactor plants

International Nuclear Information System (INIS)

Anon.

1976-01-01

This Standard sets forth design, construction, and performance requirements, with due consideration for operation, of the Liquid Radioactive Waste Processing System for pressurized water reactor plants for design basis inputs. For the purpose of this Standard, the Liquid Radioactive Waste Processing System begins at the interfaces with the reactor coolant pressure boundary and the interface valve(s) in lines from other systems, or at those sumps and floor drains provided for liquid waste with the potential of containing radioactive material; and it terminates at the point of controlled discharge to the environment, at the point of interface with the waste solidification system, and at the point of recycle back to storage for reuse

Application of membrane technologies for liquid radioactive waste processing

International Nuclear Information System (INIS)

2004-01-01

Membrane separation processes have made impressive progress since the first synthesis of membranes almost 40 years ago. This progress was driven by strong technological needs and commercial expectations. As a result the range of successful applications of membranes and membrane processes is continuously broadening. In addition, increasing application of membrane processes and technologies lies in the increasing variations of the nature and characteristics of commercial membranes and membrane apparatus. The objective of the report is to review the information on application of membrane technologies in the processing of liquid radioactive waste. The report covers the various types of membranes, equipment design, range of applications, operational experience and the performance characteristics of different membrane processes. The report aims to provide Member States with basic information on the applicability and limitations of membrane separation technologies for processing liquid radioactive waste streams

A process for treating radioactive water-reactive wastes

International Nuclear Information System (INIS)

Dziewinski, J.; Lussiez, G.; Munger, D.

1995-01-01

Los Alamos National Laboratory and other locations in the complex of experimental and production facilities operated by the United States Department of Energy (DOE) have generated an appreciable quantity of hazardous and radioactive wastes. The Resource Conservation and Recovery Act (RCRA) enacted by the United States Congress in 1976 and subsequently amended in 1984, 1986, and 1988 requires that every hazardous waste must be rendered nonhazardous before disposal. Many of the wastes generated by the DOE complex are both hazardous and radioactive. These wastes, called mixed wastes, require applying appropriate regulations for radioactive waste disposal and the regulations under RCRA. Mixed wastes must be treated to remove the hazardous waste component before they are disposed as radioactive waste. This paper discusses the development of a treatment process for mixed wastes that exhibit the reactive hazardous characteristic. Specifically, these wastes react readily and violently with water. Wastes such as lithium hydride (LiH), sodium metal, and potassium metal are the primary wastes in this category

Solidification of high-level radioactive wastes. Final report

International Nuclear Information System (INIS)

1979-06-01

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

Radioactive waste processing device

International Nuclear Information System (INIS)

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

1992-01-01

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

Strategy for research on radioactive waste processing and conditioning in France

International Nuclear Information System (INIS)

Cavedon, J.M.; Tallec, M.

2001-01-01

Research on radioactive medium level waste processing and conditioning aims at offering processing routes for waste forms and materials of potential value that are not yet provided easy handling by existing industrial processes. These studies are mandatory under the Dec 31, 1991 law and are coordinated by CEA. The strategy relies on the completion and rationalization of the existing processing routes, within acceptable technical and economic limits. Waste processing techniques aim at reducing the volume and the chemical diversity of medium activity waste, and are based on incineration-vitrification. Conditioning techniques call for high performance matrices and standardized containers, the latter keeping an ability to contain bulk waste. (author)

Processing of miscellaneous radioactive effluents by continous flocculation decantation

International Nuclear Information System (INIS)

Lundy, D.; Matton, P.; Petteau, J.L.; Roofthooft, R.

1985-01-01

In the nuclear power plant of Chooz an installation for flocculation and chemical precipitation has been built to treat miscellaneous radioactive effluents continuously. It is an industrial prototype of 5 m 3 /h resulting of several years of research, first on lab scale in a discontinous system and finally in a continuous pilot plant of small size (500 l/h). The process is based on the adsorption of radioactivity on a floc of copper-ferrocyanide precipitated by ferric chloride. The water is then filtered. After a series of preliminary tests and modifications, it has been possible to develop a technique which satisfies the specified decontamination conditions and to reduce the discharges of radioactivity to the Meuse to only 5 - 10% of the authorized limits. The process aims principally at the treatment of laundry waste, but other effluents such as drains from the rocks, pool water and used decontamination solutions (of the primary pumps) have been treated. A technico-economic evaluation of the process in comparison with evaporation is clearly in favour of the flocculation. 31 figs, 40 tables, 12 refs

Radioactive waste disposal process geological structure for the waste disposal

International Nuclear Information System (INIS)

Courtois, G.; Jaouen, C.

1983-01-01

The process described here consists to carry out the two phases of storage operation (intermediate and definitive) of radioactive wastes (especially the vitrified ones) in a geological dispositif (horizontal shafts) at an adequate deepness but suitable for a natural convection ventilation with fresh air from the land surface and moved only with the calorific heat released by the burried radioactive wastes when the radioactive decay has reached the adequate level, the shafts are totally and definitely occluded [fr

Efficient handling of high-level radioactive cell waste in a vitrification facility analytical laboratory

International Nuclear Information System (INIS)

Roberts, D.W.; Collins, K.J.

1998-01-01

The Savannah River Site''s (SRS) Defense Waste Processing Facility (DWPF) near Aiken, South Carolina, is the world''s largest and the United State''s first high level waste vitrification facility. For the past 1.5 years, DWPF has been vitrifying high level radioactive liquid waste left over from the Cold War. The vitrification process involves the stabilization of high level radioactive liquid waste into borosilicate glass. The glass is contained in stainless steel canisters. DWPF has filled more than 200 canisters 3.05 meters (10 feet) long and 0.61 meters (2 foot) diameter. Since operations began at DWPF in March of 1996, high level radioactive solid waste continues to be generated due to operating the facility''s analytical laboratory. The waste is referred to as cell waste and is routinely removed from the analytical laboratories. Through facility design, engineering controls, and administrative controls, DWPF has established efficient methods of handling the high level waste generated in its laboratory facility. These methods have resulted in the prevention of undue radiation exposure, wasted man-hours, expenses due to waste disposal, and the spread of contamination. This level of efficiency was not reached overnight, but it involved the collaboration of Radiological Control Operations and Laboratory personnel working together to devise methods that best benefited the facility. This paper discusses the methods that have been incorporated at DWPF for the handling of cell waste. The objective of this paper is to provide insight to good radiological and safety practices that were incorporated to handle high level radioactive waste in a laboratory setting

Process for disposal of aqueous solutions containing radioactive isotopes

Science.gov (United States)

Colombo, Peter; Neilson, Jr., Robert M.; Becker, Walter W.

1979-01-01

A process for disposing of radioactive aqueous waste solutions whereby the waste solution is utilized as the water of hydration to hydrate densified powdered portland cement in a leakproof container; said waste solution being dispersed without mechanical inter-mixing in situ in said bulk cement, thereafter the hydrated cement body is impregnated with a mixture of a monomer and polymerization catalyst to form polymer throughout the cement body. The entire process being carried out while maintaining the temperature of the components during the process at a temperature below 99.degree. C. The container containing the solid polymer-impregnated body is thereafter stored at a radioactive waste storage dump such as an underground storage dump.

Process for disposal of aqueous solutions containing radioactive isotopes

International Nuclear Information System (INIS)

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

1979-01-01

A process for disposing of radioactive aqueous waste solutions whereby the waste solution is utilized as the water of hydration to hydrate densified powdered portland cement in a leakproof container; said waste solution being dispersed without mechanical inter-mixing in situ in said bulk cement, thereafter the hydrated cement body is impregnated with a mixture of a monomer and polymerization catalyst to form polymer throughout the cement body. The entire process being carried out while maintaining the temperature of the components during the process at a temperature below 99 0 C. The container containing the solid polymer-impregnated body is thereafter stored at a radioactive waste storage dump such as an underground storage dump

Basic approach to the disposal of low level radioactive waste generated from nuclear reactors containing comparatively high radioactivity

International Nuclear Information System (INIS)

Moriyama, Yoshinori

1998-01-01

Low level radioactive wastes (LLW) generated from nuclear reactors are classified into three categories: LLW containing comparatively high radioactivity; low level radioactive waste; very low level radioactive waste. Spent control rods, part of ion exchange resin and parts of core internals are examples of LLW containing comparatively high radioactivity. The Advisory Committee of Atomic Energy Commission published the report 'Basic Approach to the Disposal of LLW from Nuclear Reactors Containing Comparatively High Radioactivity' in October 1998. The main points of the proposed concept of disposal are as follows: dispose of underground deep enough not be disturb common land use (e.g. 50 to 100 m deep); dispose of underground where radionuclides migrate very slowly; dispose of with artificial engineered barrier which has the same function as the concrete pit; control human activities such as land use of disposal site for a few hundreds years. (author)

High-level radioactive wastes

International Nuclear Information System (INIS)

Grissom, M.C.

1982-10-01

This bibliography contains 812 citations on high-level radioactive wastes included in the Department of Energy's Energy Data Base from January 1981 through July 1982. These citations are to research reports, journal articles, books, patents, theses, and conference papers from worldwide sources. Five indexes are provided: Corporate Author, Personal Author, Subject, Contract Number, and Report Number

Steam stripping of polycyclic aromatics from simulated high-level radioactive waste

International Nuclear Information System (INIS)

Lambert, D.P.; Shah, H.B.; Young, S.R.; Edwards, R.E.; Carter, J.T.

1992-01-01

The Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS) will be the United States' first facility to process High Level radioactive Waste (HLW) into a borosilicate glass matrix. The removal of aromatic precipitates by hydrolysis, evaporation, liquid-liquid extraction and decantation will be a key step in the processing of the HLW. This step, titled the Precipitate Hydrolysis Process, has been demonstrated by the Savannah River Technology Center with the Precipitate Hydrolysis Experimental Facility (PHEF). The mission of the PHEF is to demonstrate processing of simulated high level radioactive waste which contains tetraphenylborate precipitates and nitrite. Aqueous washing or nitrite destruction is used to reduce nitrite. Formic acid with a copper catalyst is used to hydrolyze tetraphenylborate (TPB). The primary offgases are benzene, carbon dioxide, nitrous oxide, and nitric oxide. Hydrolysis of TPB in the presence of nitrite results in the production of polycyclic aromatics and aromatic amines (referred as high boiling organics) such as biphenyl, diphenylamine, terphenyls etc. The decanter separates the organic (benzene) and aqueous phase, but the high boiling organic separation is difficult. This paper focuses on the evaluation of the operating strategies, including steam stripping, to maximize the removal of the high boiling organics from the aqueous stream. Two areas were investigated, (1) a stream stripping comparison of the late wash flowsheet to the HAN flowsheet and (2) the extraction performance of the original decanter to the new decanter. The focus of both studies was to minimize the high boiling organic content of the Precipitate Hydrolysis Aqueous (PHA) product in order to minimize downstream impacts caused by organic deposition

The Michigan high-level radioactive waste program: Final technical progress report

International Nuclear Information System (INIS)

1987-01-01

This report comprises the state of Michigan's final technical report on the location of a proposed high-level radioactive waste disposal site. Included are a list of Michigan's efforts to review the DOE proposal and a detailed report on the application of geographic information systems analysis techniques to the review process

Radioactive environmental impact assessment for a production project of titanium dioxide by chlorination process

International Nuclear Information System (INIS)

Qiu Guohua

2010-01-01

Based on the analysis of shifting direction of radionuclide in production process and the environmental investigation and monitoring, the radioactive environmental impact from a production project of titanium dioxide by chlorination process has been analyzed and assessed. The result of radioactive environmental investigation shows that values of assessment factors are in the range of environmental radioactive background. The radioactive environmental sensitive spot has been delineated. The results of radioactive environmental prediction show that the additional doses to workers and residents are 0.59 mSv/a and 9.28 × 10-4 mSv/a respectively which are less than the annual dose limits of administration. The radioactive environmental impact of the production project of the titanium dioxide by chlorination process will meet the needs of national regulations and standards if radiation protection and environmental protection measures are implemented and radioactive environmental monitoring are strengthened. (author)

Processing method for discharged radioactive laundry water waste

International Nuclear Information System (INIS)

Izumida, Tatsuo; Kitsukawa, Ryozo; Tsuchiya, Hiroyuki; Kiuchi, Yoshimasa; Hattori, Yasuo.

1995-01-01

In order to process discharged radioactive laundry water wastes safely and decrease radioactive wastes, bubbling of a surface active agent in a detergent which causes a problem upon its condensation is suppressed, so that the liquid condensate are continuously and easily dried into a powder. A nonionic surface active agent is used against the bubbling of the surface active agent. In addition, the bubbling in an the evaporation can is reduced, and the powderization is facilitated by adding an appropriate inorganic builder. (T.M.)

Method of processing liquid wastes containing radioactive materials

International Nuclear Information System (INIS)

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

1983-01-01

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

Training manual for process operation and management of radioactive waste treatment facility

Energy Technology Data Exchange (ETDEWEB)

Shon, J. S.; Kim, K. J.; Ahn, S. J. [and others

2004-12-01

Radioactive Waste Treatment Facility (RWTF) has been operating for safe and effective treatment of radioactive wastes generated in the Korea Atomic Energy Research Institute (KAERI). In RWTF, there are evaporation, bituminization and solar evaporation processes for liquid waste, solid waste treatment process and laundry process. As other radioactive waste treatment facilities in foreign countries, the emergency situation such as fire and overflow of liquid waste can be taken place during the operation and result in the spread of contamination of radioactivity. So, easy and definite operating procedure is necessary for the safe operation of the facility. This manual can be available as easy and concise training materials for new employees and workers dispatched from service agency. Especially, in case of emergency urgently occurred during operation, everyone working in the facility can quickly stop the facility following this procedure.

Training manual for process operation and management of radioactive waste treatment facility

International Nuclear Information System (INIS)

Shon, J. S.; Kim, K. J.; Ahn, S. J.

2004-12-01

Radioactive Waste Treatment Facility (RWTF) has been operating for safe and effective treatment of radioactive wastes generated in the Korea Atomic Energy Research Institute (KAERI). In RWTF, there are evaporation, bituminization and solar evaporation processes for liquid waste, solid waste treatment process and laundry process. As other radioactive waste treatment facilities in foreign countries, the emergency situation such as fire and overflow of liquid waste can be taken place during the operation and result in the spread of contamination of radioactivity. So, easy and definite operating procedure is necessary for the safe operation of the facility. This manual can be available as easy and concise training materials for new employees and workers dispatched from service agency. Especially, in case of emergency urgently occurred during operation, everyone working in the facility can quickly stop the facility following this procedure

The Hybrid Treatment Process for mixed radioactive and hazardous waste treatment

International Nuclear Information System (INIS)

Ross, W.A.; Kindle, C.H.

1992-06-01

This paper describes a new process for treating mixed hazardous and radioactive waste, commonly called mixed waste. The process is called the Hybrid Treatment Process (HTP), so named because it is built on the 20 years of experience with vitrification of wastes in melters, and the 12 years of experience with treatment of wastes by the in situ vitrification (ISV) process. It also uses techniques from several additional technologies. Mixed wastes are being generated by both the US Department of Energy (DOE) and by commercial sources. The wastes are those that contain both a hazardous waste regulated under the US Environmental Protection Agency's (EPA) Resource, Conservation, and Recovery Act (RCRA) regulations and a radioactive waste with source, special nuclear, or byproduct materials. The dual regulation of the wastes increases the complexity of the treatment, handling, and storage of the waste. The DOE is the largest holder and generator of mixed waste. Its mixed wastes are classified as either high-level, transuranic (TRU), or low-level waste (LLW). High-level mixed wastes will be treated in vitrification plants. Transuranic wastes may be disposed of without treatment by obtaining a no-migration variance from the EPA. Lowlevel wastes, however, will require treatment, but treatment systems with sufficient capacity are not yet available to DOE. Various facilities are being proposed for the treatment of low-level waste. The concept described in this paper represents one option for establishing that treatment capacity

Separation of aromatic precipitates from simulated high level radioactive waste by hydrolysis, evaporation and liquid-liquid extraction

International Nuclear Information System (INIS)

Young, S.R.; Shah, H.B.; Carter, J.T.

1991-01-01

The Defense Waste Processing Facility (DWPF) at the SRS will be the United States' first facility to process High Level radioactive Waste (HLW) into a borosilicate glass matrix. The removal of aromatic precipitates by hydrolysis, evaporation and liquid-liquid extraction will be a key step in the processing of the HLW. This step, titled the Precipitate Hydrolysis Process, has been demonstrated by the Savannah River Laboratory with the Precipitate Hydrolysis Experimental Facility (PHEF). The mission of the PHEF is to demonstrate processing of simulated high level radioactive waste which contains tetraphenylborate precipitates and nitrite. Reduction of nitrite by hydroxylamine nitrate and hydrolysis of the tetraphenylborate by formic acid is discussed. Gaseous production, which is primarily benzene, nitrous oxide and carbon dioxide, has been quantified. Production of high-boiling organic compounds and the accumulation of these organic compounds within the process are addressed

Decontamination process applied to radioactive solid wastes from nuclear power plants

International Nuclear Information System (INIS)

Franco, Milton B.; Kastner, Geraldo F.; Monteiro, Roberto Pellacani G.

2009-01-01

The process of decontamination is an important step in the economic operation of nuclear facilities. A large number of protective clothing, metallic parts and equipment get contaminated during the handling of radioactive materials in laboratory, plants and reactors. Safe and economic operation of these nuclear facilities will have a bearing on the extent to which these materials are reclaimed by the process of decontamination. The most common radioactive contaminants are fission products, corrosion products, uranium and thorium. The principles involved in decontamination are the same as those for an industrial cleaning process. However, the main difference is in the degree of cleaning required and at times special techniques have to be employed for removing even trace quantities of radioactive materials. This paper relate decontaminations experiences using acids and acids mixtures (HCl, HF, HNO 3 , KMnO 4 , C 2 H 2 O 4 , HBF 4 ) in several kinds of radioactive solid wastes from nuclear power plants. The result solutions were monitored by nuclear analytical techniques, in order to contribute for radiochemical characterization of these wastes. (author)

Radioactive Demonstration of Caustic Recovery from Low-Level Alkaline Nuclear Waste by an Electrochemical Separation Process

International Nuclear Information System (INIS)

Hobbs, D.T.

1998-04-01

Bench-scale radioactive tests successfully demonstrated an electrochemical process for the recovery of sodium hydroxide (caustic) from Decontaminated Salt Solution produced from the In-Tank Precipitation and Effluent Treatment Processes at the Savannah River Site (SRS). This testing evaluated two membranes: an organic-based membrane, Nafion Type 350, manufactured by E. I. duPont de Nemours ampersand Company, Inc. (DuPont) and an inorganic-based membrane, NAS D, being developed by Ceramatec. Both membranes successfully separated caustic from radioactive SRS waste.Key findings of the testing indicate the following attributes and disadvantages of each membrane. The commercially-available Nafion membrane proved highly conductive. Thus, the electrochemical cell can operate at high current density minimizing the number of cells at the desired volumetric processing rate. Testing indicated cesium transported across the Nafion membrane into the caustic product. Therefore, the caustic product will contain low-levels of radioactive cesium due to the presence of 134,137 Cs in the waste feed. To meet customer requirements, a post treatment stage may prove necessary to remove radioactive cesium resulting in increased overall process costs and decreased cost savings. In contrast to the Nafion membrane, the NAS D membrane demonstrated the production of caustic with much lower levels of gamma radioactivity ( 137 Cs activity was < 51 dpm/g). Therefore, the caustic product could possibly release for onsite/offsite use without further treatment. The NAS D membrane remains in the development stage and does not exist as a commercial product. Operating costs and long-term membrane durability remain unknown.Caustic recovery has been successfully demonstrated in a bench-scale, 2-compartment electrochemical reactor operated for brief periods of time with simulated and radioactive waste solutions and two different types of membranes. The next phase of testing should be directed at (1

Peculiarities of the radioactive particles transport phenomena in the facilitated diffusion processes

International Nuclear Information System (INIS)

Gavryushenko, D.A.; Sisojev, V.M.; Cherevko, K.V.; Vlasenko, T.S.

2017-01-01

The work is devoted to the up to date problem that is the description of the radioactive particle diffusion processes. One of the aims of the present study is to estimate the effects caused by the irradiation of the liquid systems on the ongoing transport processes. That can allow predicting the behavior of the liquid systems in the presence of the radioactive sources. The main objective of the present work is studying the radioactive particles diffusion phenomena with the possible facilitated diffusion processes being considered. The phenomena are studied based on the fundamental relations of the nonequilibrium statistical thermodynamics. The diffusive flows are evaluated with the special attention given to the accounting of the entropy effects due to the appearance of the new radioactive particles in the system. The developed approach is used to estimate the diffusive flow of the radioactive particles for the case of the plane-parallel pore with the semi-transparent walls. The choice of a model can be justified as it might be used to describe the production of the radioactive contaminated water when the radionuclide face the diffusion process after being washed from the radioactive wastes and the rests of the nuclear fuel. Within the suggested model it is shown that the diffusion coefficient depends on the structural properties of the liquid systems that might be changed under the influence of the irradiation. The obtained equations for calculating the diffusive flows show the definite stabilizing effect in respect to the concentration difference in between the boundaries of the plane-parallel pore. It leads to the decreased changes of the diffusive flow when the concentrations of the radioactive particles at the boundaries are changed in comparison with those observed for the constant diffusion coefficient. The observed behavior for the ideal solution model is explained by the entropy effects. The qualitative analysis of the possible influence of the changes in

Process for selectively concentrating the radioactivity of thorium containing magnesium slag

International Nuclear Information System (INIS)

Wilson, D.A.; Christiansen, S.H.; Simon, J.; Morin, D.W.

1993-01-01

In a process for separating magnesium from a magnesium slag using water and carbon dioxide, the improvement described comprises: (a) forming an aqueous magnesium slurry from the magnesium slag, which slag contains radioactive thorium and its daughters, and water; (b) solubilizing magnesium from the magnesium slurry by reacting the aqueous magnesium slurry with carbon dioxide wherein the carbon dioxide is at a pressure from greater than ambient to about 1,000 psig (about 7,000 kPa); (c) selectively concentrating by filtering the radioactive thorium and its daughters such that the radioactive thorium and its daughters are separated from the solubilized magnesium filtrate; and (d) reducing volume and/or weight of radioactive solids for disposal as radioactive waste

Radioactive Waste Conditioning, Immobilisation, And Encapsulation Processes And Technologies: Overview And Advances (Chapter 7)

Energy Technology Data Exchange (ETDEWEB)

Jantzen, Carol M. [Savannah River National Lab., Aiken SC (United States); Lee, William E. [Imperial College, London (United Kingdom). Dept. of Materials; Ojovan, Michael I. [Univ. of Sheffield (United Kingdom). Dept. of Materials Science and Engineering

2012-10-19

The main immobilization technologies that are available commercially and have been demonstrated to be viable are cementation, bituminization, and vitrification. Vitrification is currently the most widely used technology for the treatment of high level radioactive wastes (HLW) throughout the world. Most of the nations that have generated HLW are immobilizing in either alkali borosilicate glass or alkali aluminophosphate glass. The exact compositions of nuclear waste glasses are tailored for easy preparation and melting, avoidance of glass-in-glass phase separation, avoidance of uncontrolled crystallization, and acceptable chemical durability, e.g., leach resistance. Glass has also been used to stabilize a variety of low level wastes (LLW) and mixed (radioactive and hazardous) low level wastes (MLLW) from other sources such as fuel rod cladding/decladding processes, chemical separations, radioactive sources, radioactive mill tailings, contaminated soils, medical research applications, and other commercial processes. The sources of radioactive waste generation are captured in other chapters in this book regarding the individual practices in various countries (legacy wastes, currently generated wastes, and future waste generation). Future waste generation is primarily driven by interest in sources of clean energy and this has led to an increased interest in advanced nuclear power production. The development of advanced wasteforms is a necessary component of the new nuclear power plant (NPP) flowsheets. Therefore, advanced nuclear wasteforms are being designed for robust disposal strategies. A brief summary is given of existing and advanced wasteforms: glass, glass-ceramics, glass composite materials (GCM’s), and crystalline ceramic (mineral) wasteforms that chemically incorporate radionuclides and hazardous species atomically in their structure. Cementitious, geopolymer, bitumen, and other encapsulant wasteforms and composites that atomically bond and encapsulate

Foaming and Antifoaming in Radioactive Waste Pretreatment and Immobilization Processes

International Nuclear Information System (INIS)

Wasan, Darsh T.; Nikolov, Alex D.; Lamber, D.P.; Calloway, T. Bond; Stone, M.E.

2005-01-01

Savannah River National Laboratory (SRNL) has reported severe foaminess in the bench scale evaporation of the Hanford River Protection - Waste Treatment Plant (RPP-WPT) envelope C waste. Excessive foaming in waste evaporators can cause carryover of radionuclides and non-radioactive waste to the condensate system. The antifoams used at Hanford and tested by SRNL are believed to degrade and become inactive in high pH solutions. Hanford wastes have been known to foam during evaporation causing excessive down time and processing delays

Spent fuel and high-level radioactive waste storage

International Nuclear Information System (INIS)

Trigerman, S.

1988-06-01

The subject of spent fuel and high-level radioactive waste storage, is bibliographically reviewed. The review shows that in the majority of the countries, spent fuels and high-level radioactive wastes are planned to be stored for tens of years. Sites for final disposal of high-level radioactive wastes have not yet been found. A first final disposal facility is expected to come into operation in the United States of America by the year 2010. Other final disposal facilities are expected to come into operation in Germany, Sweden, Switzerland and Japan by the year 2020. Meanwhile , stress is placed upon the 'dry storage' method which is carried out successfully in a number of countries (Britain and France). In the United States of America spent fuels are stored in water pools while the 'dry storage' method is still being investigated. (Author)

Effects of food processing on the radioactivity of vegetables and cereals

International Nuclear Information System (INIS)

Roussel-Debet, S.; Real, J.

1995-01-01

The decrease of radioactivity for 134 Cs, 85 Sr, 106 Ru, 57 Co, 110m Ag during food processing was measured. Washing and bleaching vegetables directly contaminated by a single aerosol deposit led to a radioactivity decrease (depending on the vegetable, the radionuclide and the time of deposition) up to 90% for caesium. The efficiency of these processes was less important when the vegetables were contaminated by root transfer, except for peeling of root vegetables. Canning of vegetables indirectly contaminated by caesium, strontium, cobalt and ruthenium, after paring and bleaching, was rather interesting with residual radioactivity varying from 30 to 50% for green beans and from 5 to 20% for carrots. Measurements of stable caesium and strontium in cereals before and after industrial processing showed an important activity decrease; retention factors varied from 0.1 to 0.2 for wheat milling and 0.1 to 0.4 for rice processing. (authors). 6 refs., 9 tabs., 1 figs

USING STATISTICAL PROCESS CONTROL TO MONITOR RADIOACTIVE WASTE CHARACTERIZATION AT A RADIOACTIVE FACILITY

International Nuclear Information System (INIS)

WESTCOTT, J.L.

2006-01-01

Two facilities for storing spent nuclear fuel underwater at the Hanford site in southeastern Washington State being removed from service, decommissioned, and prepared for eventual demolition. The fuel-storage facilities consist of two separate basins called K East (KE) and K West (KW) that are large subsurface concrete pools filled with water, with a containment structure over each. The basins presently contain sludge, debris, and equipment that have accumulated over the years. The spent fuel has been removed from the basins. The process for removing the remaining sludge, equipment, and structure has been initiated for the basins. Ongoing removal operations generate solid waste that is being treated as required, and then disposed. The waste, equipment and building structures must be characterized to properly manage, ship, treat (if necessary), and dispose as radioactive waste. As the work progresses, it is expected that radiological conditions in each basin may change as radioactive materials are being moved within and between the basins. It is imperative that these changing conditions be monitored so that radioactive characterization of waste is adjusted as necessary

USING STATISTICAL PROCESS CONTROL TO MONITOR RADIOACTIVE WASTE CHARACTERIZATION AT A RADIOACTIVE FACILITY

International Nuclear Information System (INIS)

WESTCOTT, J.L.; JOCHEN; PREVETTE

2007-01-01

Two facilities for storing spent nuclear fuel underwater at the Hanford site in southeastern Washington State are being removed from service, decommissioned, and prepared for eventual demolition. The fuel-storage facilities consist of two separate basins called K East (KE) and K West (KW) that are large subsurface concrete pools filled with water, with a containment structure over each. The basins presently contain sludge, debris, and equipment that have accumulated over the years. The spent fuel has been removed from the basins. The process for removing the remaining sludge, equipment, and structure has been initiated for the basins. Ongoing removal operations generate solid waste that is being treated as required, and then disposed. The waste, equipment and building structures must be characterized to properly manage, ship, treat (if necessary), and dispose as radioactive waste. As the work progresses, it is expected that radiological conditions in each basin may change as radioactive materials are being moved within and between the basins. It is imperative that these changing conditions be monitored so that radioactive characterization of waste is adjusted as necessary

Forward osmosis - a novel membrane process for concentration of low level radioactive wastes

International Nuclear Information System (INIS)

Ghosh, A.K.; Bindal, R.C.; Tewari, P.K.

2013-01-01

Forward osmosis (FO) is an emerging membrane process in which osmotic pressure differential across a semi-permeable membrane between the solution to be concentrated (feed) and a concentrated solution of high osmotic pressure (draw solution) than the feed is used to effect separation of water from dissolved solutes. With time, feed stream gets concentrated with dilution of draw solution and this technology recently being used as more energy efficient alternative to reverse osmosis (RO) in some of the application areas, particularly for the concentration of low volume high value products. The use of pressure driven membrane processes like reverse osmosis (RO) and ultrafiltration (UF) are already demonstrated in the treatment of radioactive laundry, laboratory effluents and some other applications in nuclear industry. The application of FO membrane process to concentrate simulated inactive ammonium-diuranate (ADU) filtered effluent solution (by mixing uranyl nitrate and ammonium nitrate) using indigenously developed cellulose acetate (CA) and thin-film composite polyamide (TFCP) membranes has been published recently from our laboratory. In this presentation, we briefly discuss our views on possibility of using FO membrane process with proper selection of membrane for concentration of low level radioactive wastes generated in various steps of nuclear fuel cycle in most effective way. (author)

High-level radioactive waste isolation by incorporation in silicate rock

International Nuclear Information System (INIS)

Schwartz, L.L.; Cohen, J.J.; Lewis, A.E.; Braun, R.L.

1978-01-01

A number of technical possibilities for isolating high-level radioactive materials have been theoretically investigated at various times and places. Isolating such wastes deep underground to ensure long term removal from the biosphere is one such possibility. The present concept involves as a first step creating the necessary void space at considerable depth, say 2 to 5 km, in a very-low-permeability silicate medium such as shale. Waste in dry, calcined or vitrified form is then lowered into the void space, and the access hole or shaft sealed. Energy released by the radioactive decay raises the temperature to a point where the surrounding rock begins to melt. The waste is then dissolved in it. The extent of this melt region grows until the heat generated is balanced by conduction away from the molten zone. Resolidification then begins, and ends when the radioactive decay has progressed to the point that the temperature falls below the melting point of the rock-waste solution. Calculations are presented showing the growth and resolidification process. A nuclear explosion is one way of creating the void space. (author)

Removal of high-level radioactive substances contained with water from the Fukushima No. 1 Nuclear Power Stations. Some technical problems in waste treatment

International Nuclear Information System (INIS)

Amano, Osamu; Mimura, Hitoshi; Sato, Nobuaki; Kirishima, Akira; Hattori, Toshio

2011-01-01

The Japanese government and plant operator Tokyo Electric Power Co. announced to process the highly radioactive water amounting to about 250,000 cubic meters by the end of fiscal year 2011. Radiation-contaminated water will be moved to the waste facility to remove oil and radioactive cesium using zeolite. The process using Prussian Blue is expected for the effectiveness. Other radioactive substances will be removed through precipitation using special chemicals and radioactivity in the water will be reduced to 10 -6 of its original level. The water will be then be returned to the reactors and used to cool them after going through a desalination process. The facility can process about 1,200 tons of contaminated water a day. TEPCO will store radioactive materials and other waste from the cleansing process at the Fukushima plant. They need to decide how the waste will finally be disposed of and to figure out what to do with the highly radioactive waste produced in the above process. Kurion Inc., Areva SA, and some domestic firms provide equipment and technology, but all the Japanese facilities and institutions should join to settle the problems. (S. Ohno)

The conceptual design of waste repository for radioactive waste from medical, industrial and research facilities containing comparatively high radioactivity

International Nuclear Information System (INIS)

Yamamoto, Masayuki; Hashimoto, Naro

2002-02-01

Advisory Committee on Nuclear Fuel Cycle Backend Policy reported the basic approach to the RI and Institute etc. wastes on March 2002. According to it, radioactive waste form medical, industrial and research facilities should be classified by their radioactivity properties and physical and chemical properties, and should be disposed in the appropriate types of repository with that classification. For the radioactive waste containing comparatively high radioactivity generated from reactors, NSC has established the Concentration limit for disposal. NSC is now discussing about the limit for the radioactive waste from medical, industrial and research facilities containing comparatively high radioactivity. Japan Nuclear Cycle Development Institute (JNC) preliminary studied about the repository for radioactive waste from medical, industrial and research facilities and discussed about the problems for design on H12. This study was started to consider those problems, and to develop the conceptual design of the repository for radioactive waste from medical, industrial and research facilities. Safety assessment for that repository is also performed. The result of this study showed that radioactive waste from medical, industrial and research facilities of high activity should be disposed in the repository that has higher performance of barrier system comparing with the vault type near surface facility. If the conditions of the natural barrier and the engineering barrier are clearer, optimization of the design will be possible. (author)

Argentine project for the final disposal of high-level radioactive wastes

International Nuclear Information System (INIS)

Palacios, E.; Ciallella, N.R.; Petraitis, E.J.

1989-01-01

From 1980 Argentina is carrying out a research program on the final disposal of high level radioactive wastes. The quantity of wastes produced will be significant in next century. However, it was decided to start with the studies well in advance in order to demonstrate that the high level wastes could be disposed in a safety way. The option of the direct disposal of irradiated fuel elements was discarded, not only by the energetic value of the plutonium, but also for ecological reasons. In fact, the presence of a total inventory of actinides in the non-processed fuel would imply a more important radiological impact than that caused if the plutonium is recycled to produce energy. The decision to solve the technological aspects connected with the elimination of high-level radioactive wastes well in advance, was made to avoid transfering the problem to future generations. This decision is based not only on technical evaluations but also on ethic premises. (Author)

Proposed classification scheme for high-level and other radioactive wastes

International Nuclear Information System (INIS)

Kocher, D.C.; Croff, A.G.

1986-01-01

The Nuclear Waste Policy Act (NWPA) of 1982 defines high-level radioactive waste (HLW) as: (A) the highly radioactive material resulting from the reprocessing of spent nuclear fuel....that contains fission products in sufficient concentrations; and (B) other highly radioactive material that the Commission....determines....requires permanent isolation. This paper presents a generally applicable quantitative definition of HLW that addresses the description in paragraph (B). The approach also results in definitions of other waste classes, i.e., transuranic (TRU) and low-level waste (LLW). A basic waste classification scheme results from the quantitative definitions

Method of processing radioactive wastes

International Nuclear Information System (INIS)

Takahashi, Toshihiko; Maruko, Morihisa; Takamura, Yoshiyuki.

1981-01-01

Purpose: To effectively separate radioactive claddings from the slurry of wasted ion exchange resins containing radioactive claddings. Method: Wasted ion exchange resins having radioactive claddings (fine particles of iron oxides or hydroxide adhered with radioactive cobalt) are introduced into a clad separation tank. Sulfuric acid or sodium hydroxide is introduced to the separation tank to adjust the pH value to 3 - 6. Then, sodium lauryl sulfate is added for capturing claddings and airs are blown from an air supply nozzle to generate air bubbles. The claddings are detached from the ion exchange resins and adhered to the air bubbles. The air bubbles adhered with the claddings float up to the surface of the liquid wastes and then forced out of the separation tank. (Ikeda, J.)

Processes for the administration give the radio-active lightning rod in Brazil

International Nuclear Information System (INIS)

Torres de Miranda, L.E.; Vicente, R.

1998-01-01

Radioactive lightning rods were manufactured in Brazil until 1989, when the national nuclear authority lifted the licenses to using radioactive sources in this consumer product. Since then, the installed devices are being replaced by Franklin type rods and being collected for treatment as radioactive waste. In this report we present options for the treatment of this radioactive waste. The conclusions may serve as input to the process of decision on the overall management strategic to be adopted by the authorities

Solid radioactive waste processing system for light water cooled reactor plants

International Nuclear Information System (INIS)

Anon.

1979-01-01

Design, construction and performance requirements are given for the operation of the solid radioactive waste processing system for light water-cooled reactor plants. All radioactive or contaminated materials, including spent air and liquid filter elements, spent bead resins, filter sludge, spent powdered resins, evaporator and reverse osmosis concentrates, and dry radioactive wastes are to be processed in appropriate portions of the system. Sections of the standard cover: overall system requirements; equipment requirement; controls and instrumentation; physical arrangement; system capacity and redundancy; operation and maintenance; and system construction and testing. Provisions contained in this standard are to take precedence over ANS-51.1-1973(N18.2-1973) and its revision, ANS-51.8-1975(N18.2a-1975), Sections 2.2 and 2.3. The product resulting from the solid radioactive waste processing system must meet criteria imposed by standards and regulations for transportation and burial (Title 10, Code of Federal Regulations, Part 71, Title 49, Code of Federal Regulations, Parts 100 to 199). As a special feature, all statements in this standard which are related to nuclear safety are set off in boxes

Electrochemical probing of high-level radioactive waste tanks containing washed sludge and precipitates

International Nuclear Information System (INIS)

Bickford, D.F.; Congdon, J.W.; Oblath, S.B.

1987-01-01

At the U.S. Department of Energy's Savannah River Plant, corrosion of carbon steel storage tanks containing alkaline, high-level radioactive waste is controlled by specification of limits on waste composition and temperature. Processes for the preparation of waste for final disposal will result in waste with low corrosion inhibitor concentrations and, in some cases, high aromatic organic concentrations, neither of which are characteristic of previous operations. Laboratory tests, conducted to determine minimum corrosion inhibitor levels indicated pitting of carbon steel near the waterline for proposed storage conditions. In situ electrochemical measurements of full-scale radioactive process demonstrations have been conducted to assess the validity of laboratory tests. Probes included pH, Eh (potential relative to a standard hydrogen electrode), tank potential, and alloy coupons. In situ results are compared to those of the laboratory tests, with particular regard given to simulated solution composition

Radioactive gas waste processing device

International Nuclear Information System (INIS)

Soma, Koichi.

1996-01-01

The present invention concerns a radioactive gas waste processing device which extracts exhaust gases from a turbine condensator in a BWR type reactor and releases them after decaying radioactivity thereof during temporary storage. The turbine condensator is connected with an extracting ejector, a preheater, a recombiner for converting hydrogen gas into steams, an off gas condensator for removing water content, a flow rate control valve, a dehumidifier, a hold up device for removing radiation contaminated materials, a vacuum pump for sucking radiation decayed-off gases, a circulation water tank for final purification and an exhaustion cylinder by way of connection pipelines in this order. An exhaust gas circulation pipeline is disposed to circulate exhaust gases from an exhaust gas exit pipeline of the recycling water tank to an exhaust gas exit pipeline of the exhaust gas condensator, and a pressure control valve is disposed to the exhaust gas circulation pipeline. This enable to perform a system test for the dehumidification device under a test condition approximate to the load of the dehumidification device under actual operation state, and stabilize both of system flow rate and pressure. (T.M.)

Hydrogeological investigation for sitting disposal repository for high level radioactive waste

International Nuclear Information System (INIS)

Guo Yonghai; Liu Shufen; Lv Chuanhe

2005-01-01

Based on the research experiences of our country and some developed countries in the world, the purpose, process and methods, as well as the function of hydrogeological investigation for sitting disposal repository for high radioactive waste are discussed. Meanwhile, the topic related to the acquisition of hydrogeological parameters is described as well, aiming at providing reference for the future study. (authors)

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.)

Processing method of radioactive cleaning drain

International Nuclear Information System (INIS)

Otsuka, Shigemitsu; Murakami, Tadashi; Kitao, Hideo

1998-01-01

Upon processing of radioactive cleaning drains, contained Co-60 is removed by a selectively adsorbing adsorbent. In addition, after removing suspended materials by a filtering device, Co-60 as predominant nuclides in the drain is selectively adsorbed. The concentration of objective Co-60 is in the order of 0.1 Bq/cc, and non-radioactive metal ions such as Na + ions are present in the order of ppm in addition to Co-60. A granular adsorbent for selectively adsorbing Co-60 is oxine-added activated carbon, and has a grain size of from 20 to 48 mesh. The granular adsorbent is used while being filled in an adsorbing tower. Since a relatively simple device comprising the filtering device and the adsorbing tower in combination is provided, the reduction of the construction cost can be expected. In addition, since no filtering aid is used in the filtering device, the amount of secondary wastes is small. (N.H.)

Discarding processing method for radioactive waste

International Nuclear Information System (INIS)

Komura, Shiro; Kato, Hiroaki; Hatakeyama, Takao; Oura, Masato.

1992-01-01

At first, in a discrimination step, extremely low level radioactive wastes are discriminated to metals and concretes and further, the metal wastes are discriminated to those having hollow portions and those not having hollow portions, and the concrete wastes are discriminated to those having block-like shape and those having other shapes respectively. Next, in a processing step, the metal wastes having hollow portions are applied with cutting, devoluming or packing treatment and block-like concrete wastes are applied with surface solidification treatment, and concrete wastes having other shapes are applied with crushing treatment respectively. Then, the extremely low level radioactive wastes contained in a container used exclusively for transportation are taken out, in a movable burying facility with diffusion inhibiter kept at a negative pressure as required, in a field for burying operation, and buried in a state that they are isolated from the outside. Accordingly, they can be buried safely and efficiently. (T.M.)

Measurements of radioactive dust in high altitude air

International Nuclear Information System (INIS)

Kobayashi, Mika; Kohara, Eri; Muronoi, Naohiro; Masuda, Yousuke; Midou, Tomotaka; Ishida, Yukiko; Shimizu, Toshihiko; Saga, Minoru; Endo, Hiromu

2012-01-01

The radioactivity in samples of airborne dust was measured. The samples had been collected at high altitude by the Japan Air Self-Defense Force. The data were obtained for the gross beta activity, gamma nuclide determination and radiochemical analysis. It was shown that there was no appreciable difference between the activity levels obtained in this time and in the year before. Seasonal variations were not very pronounced. It was found that the radioactivity at high altitude had been stable at a low level. Radioactive gases (gaseous radioiodine and xenon gas) were not detected. This report does not include the result on radionuclide measurements that Technical Research and Development Institute executed for examining the nuclear emergency situation at Fukushima Daiichi and Daini nuclear power plants after Tohoku Region Pacific Ocean Earthquake on March 11, 2011. (author)

Storage process of large solid radioactive wastes

International Nuclear Information System (INIS)

Morin, Bruno; Thiery, Daniel.

1976-01-01

Process for the storage of large size solid radioactive waste, consisting of contaminated objects such as cartridge filters, metal swarf, tools, etc, whereby such waste is incorporated in a thermohardening resin at room temperature, after prior addition of at least one inert charge to the resin. Cross-linking of the resin is then brought about [fr

Chernobyl radioactivity and high altitude air-particulate monitoring at Islamabad

International Nuclear Information System (INIS)

Bhatti, M.S.; Ihsanullah; Shafiq, M.; Perveen, N.; Orfi, S.D.

1987-11-01

High altitude sampling of air particulates for radioactivity monitoring was conducted at Islamabad after the CHERNOBYL accident. Smears from aeroplanes flying at varying altitudes were collected and analysed for fresh fission products mainly gamma emitters e.g. Ru-103 and Cs-137 etc. The maximum radioactivity observed was of the order of 15Bq/sample for Ru-103 and 9Bq/sample for Cs-137 respectively. The study was purely qualitative in nature indicated the presence of fresh fission radioactivity at high altitudes over Islamabad. For quantitative measurements at high altitudes sophisticated instrumentation/procedure needs to be adopted. (author)

High level radioactive wastes: Considerations on final disposal

International Nuclear Information System (INIS)

Ciallella, Norberto R.

2000-01-01

When at the beginnings of the decade of the 80 the National Commission on Atomic Energy (CNEA) in Argentina decided to study the destination of the high level radioactive wastes, was began many investigations, analysis and multidisciplinary evaluations that be origin to a study of characteristics never before carried out in Argentina. For the first time in the country was faced the study of an environmental eventual problem, several decades before that the problem was presented. The elimination of the high level radioactive wastes in the technological aspects was taken in advance, avoiding to transfer the problems to the future generations. The decision was based, not only in technical evaluations but also in ethical premises, since it was considered that the future generations may enjoy the benefits of the nuclear energy and not should be solve the problem. The CNEA in Argentina in 1980 decided to begin a feasibility study and preliminary engineering project for the construction of the final disposal of high level radioactive wastes

Method of processing radioactive nuclide-containing liquids

International Nuclear Information System (INIS)

Hirai, Masahide; Tomoshige, Shozo; Kondo, Kozo; Suzuki, Kazunori; Todo, Fukuzo; Yamanaka, Akihiro.

1985-01-01

Purpose: To solidify radioactive nuclides in to a much compact state and facilitate the storage. Method: Liquid wastes such as drain liquids generated from a nuclear power plant at a low density of 1 x 10 -6 - 10 -4 μCi/ml are previously brought into contact with a chelate type ion exchange resin such as of phenolic resin or ion exchange resin to adsorb the radioactive nuclides on the resin and the nuclides are eluted with sulfuric acid or the like to obtain liquid concentrates. The liquid concentrates are electrolyzed in an ordinary electrolytic facility using platinum or the like as the anode, Al or the like as the cathode, under the presence of 1 - 20 g/l of non-radioactive heavy metals such as Co and Ni in the liquid and while adjusting pH to 2 - 8. The electrolysis liquid residue is returned again to the electrolysis tank as it is or in the form of precipitates coagulated with a polymeric floculant. The supernatant liquid upon floculating treatment is processed with the chelate type ion exchange resin into hazardless liquid. (Sekiya, K.)

The high level and long lived radioactive wastes

International Nuclear Information System (INIS)

2005-01-01

This report presents the main conclusions of 15 years of researches managed by the CEA. This report is the preliminary version of the 2005 final report. It presents the main conclusions of the actions on the axis 1 and 3 of the law of the 30 December 1991. The synthesis report on the axis 1 concerns results obtained on the long lived radionuclides separation and transmutation in high level and long lived radioactive wastes. the synthesis report on the axis 3 presents results obtained by the processes of conditioning and of ground and underground long term storage. (A.L.B.)

Cement solidification method for miscellaneous radioactive solid, processing device and processing tool therefor

International Nuclear Information System (INIS)

Mihara, Shigeru; Suzuki, Kazunori; Hasegawa, Akira.

1994-01-01

A basket made of a metal net and a lid with a spacer constituting a processing tool for processing miscellaneous radioactive solid wastes is formed as a mesh which scarcely passes the miscellaneous solids but pass mortars. The size of the mesh is usually from about 10 to 30mm. Since this mesh allows fine solids approximate to powders such as burning ashes and heat insulation materials, they fall to the bottom of a dram can, to cause corrosion. Then, the corners of the bottom and the bottom of the dram can are coated with cement. The miscellaneous solid wastes are contained, and the lid of a metal net having a spacer at the upper portion thereof is set, a provisional lid is put on, and it is evacuated, and mortars are injected. Since there is a possibility that light and fine radioactive powders are exposed on the surface of the mortars coagulated and hardened by curing, conditioning for further adding mortars is applied for securing the mortars in order to prevent scattering of the radioactive powders. With such procedures, a satisfactory safe solidified products can be formed. (T.M.)

Glasses used for the high level radioactive wastes storage

International Nuclear Information System (INIS)

Sombret, C.

1983-06-01

High level radioactive wastes generated by the reprocessing of spent fuels is an important concern in the conditioning of radioactive wastes. This paper deals with the status of the knowledge about glasses used for the treatment of these liquids [fr

Evaluation procedure for radioactive waste treatment processes

International Nuclear Information System (INIS)

Whitty, W.J.

1979-11-01

An aspect of the Los Alamos Scientific Laboratory's nuclear waste management R and D programs has been to develop an evaluation procedure for radioactive waste treatment processes. This report describes the process evaluation method. Process worth is expressed as a numerical index called the Figure-of-Merit (FOM), which is computed using a hierarchial, linear, additive, scoring model with constant criteria weights and nonlinear value functions. A numerical example is used to demonstrate the procedure and to point out some of its strengths and weaknesses. Potential modifications and extensions are discussed, and an extensive reference list is included

Radioactivity partitioning of oil sludge undergoing incineration process

International Nuclear Information System (INIS)

Muhamat Omar; Suhaimi Hamzah; Muhd Noor Muhd Yunus

1997-01-01

Oil sludge waste is a controlled item under the Atomic Energy Act (Act 304) 1984 of which the radioactivity content shall be subjected to analysis. Apart from that the treatment method also shall be approved by Atomic Energy Licensing Board (AELB). Thus, an analysis of the oil sludge for MSE fluidized incinerator was conducted to comply with above requirements using various techniques. Further screening analysis of fly ash as well as bed material were done to study the effect of incinerating the sludge. This paper highlights the analysis techniques and discusses the results with respect to the radioactivity level and the fate of radionuclides subjected to the processing of the waste

Predisposal management of high level radioactive waste. Safety guide

International Nuclear Information System (INIS)

2005-01-01

Radioactive waste is generated in the generation of electricity in nuclear power plants and in the use of radioactive material in industry, research and medicine. The importance of the safe management of radioactive waste for the protection of human health and the environment has long been recognized. The principles and requirements that govern the safety of the management of radioactive waste are presented in 'The Principles of Radioactive Waste Management', 'Legal and Governmental Infrastructure for Nuclear, Radiation, Radioactive Waste and Transport Safety' and 'Predisposal Management of Radioactive Waste, Including Decommissioning'. The objective of this Safety Guide is to provide regulatory bodies and the operators that generate and manage radioactive waste with recommendations on how to meet the principles and requirements established in Refs for the predisposal management of HLW. This Safety Guide applies to the predisposal management of HLW. For liquid HLW arising from the reprocessing of spent fuel the recommendations of this Safety Guide apply from when liquid waste from the first extraction process is collected for storage and subsequent processing. Recommendations and guidance on the storage of spent fuel, whether or not declared as waste, subsequent to its removal from the storage facility of a reactor are provided in Refs. For spent fuel declared as waste this Safety Guide applies to all activities subsequent to its removal from the storage facility of a reactor and prior to its disposal. Requirements pertaining to the transport of spent fuel, whether or not declared as waste, and of all forms of HLW are established. This Safety Guide provides recommendations on the safety aspects of managing HLW, including the planning, design, construction, commissioning, operation and decommissioning of equipment or facilities for the predisposal management of HLW. It addresses the following elements: (a) The characterization and processing (i.e. pretreatment

Optimal operation planning of radioactive waste processing system by fuzzy theory

International Nuclear Information System (INIS)

Yang, Jin Yeong; Lee, Kun Jai

2000-01-01

This study is concerned with the applications of linear goal programming and fuzzy theory to the analysis of management and operational problems in the radioactive processing system (RWPS). The developed model is validated and verified using actual data obtained from the RWPS at Kyoto University in Japan. The solution by goal programming and fuzzy theory would show the optimal operation point which is to maximize the total treatable radioactive waste volume and minimize the released radioactivity of liquid waste even under the restricted resources. (orig.)

Radioactive waste processing field

International Nuclear Information System (INIS)

Ito, Minoru.

1993-01-01

Storing space for radioactive wastes (storage tunnels) are formed underground of the sea bottom along coast. A plurality of boreholes through which sea water flows are pored vertically in a direction intersecting underground streams of brine in the ground between the tunnels and seaside. Sea water introduction pipes are joined to the upper side walls of the boreholes. The sea water introduction pipes have introduction ports protruded under the sea level of the coastal sea area region. Since sea water flows from the introduction ports to the boreholes passing through the sea water introduction pipes, sea water is always filled in the boreholes. Therefore, brine is sufficiently supplied toward the land by sea water from the boreholes, the underground stream of brine is negligibly small. This can prevent radioactive contamination due to flow of the underground water when radioactive wastes are buried in the underground near coast. (I.N.)

Method of processing radioactive wastes

International Nuclear Information System (INIS)

Matagi, Yoshihiko; Takahara, Akira; Ootsuka, Katsuyuki.

1984-01-01

Purpose: To avoid the reduction in the atmospheric insulation by preventing the generation of CO 2 , H 2 O, etc. upon irradiation of microwave heat. Method: Radioactive wastes are charged into a hopper, supplied on a conveyor, fed each by a predetermined amount to a microwave furnace and heated by microwaves applied from a microwave guide. Simultaneously, inert gases are supplied from a supply line. The Radioactive wastes to be treated are shielded by the inert gases to prevent the combustion of decomposed gases produced from the wastes upon irradiation of microwave heat to thereby prevent the generation of CO 2 , H 2 , etc., as well as the generated decomposed gases are diluted with the inert gases to decrease the dissociation of the decomposed gases to prevent the reduction in the atmospheric insulation. Since the spent inert gases can be recovered for reuse, the amount of gaseous wastes released to the atmosphere can be decreased and the working life of the high performance air filters can be extended. (Sekiya, K.)

A proposed classification system for high-level and other radioactive wastes

International Nuclear Information System (INIS)

Kocher, D.C.; Croff, A.G.

1987-06-01

This report presents a proposal for quantitative and generally applicable risk-based definitions of high-level and other radioactive wastes. On the basis of historical descriptions and definitions of high-level waste (HLW), in which HLW has been defined in terms of its source as waste from reprocessing of spent nuclear fuel, we propose a more general definition based on the concept that HLW has two distinct attributes: HLW is (1) highly radioactive and (2) requires permanent isolation. This concept leads to a two-dimensional waste classification system in which one axis, related to ''requires permanent isolation,'' is associated with long-term risks from waste disposal and the other axis, related to ''highly radioactive,'' is associated with shorter-term risks due to high levels of decay heat and external radiation. We define wastes that require permanent isolation as wastes with concentrations of radionuclides exceeding the Class-C limits that are generally acceptable for near-surface land disposal, as specified in the US Nuclear Regulatory Commission's rulemaking 10 CFR Part 61 and its supporting documentation. HLW then is waste requiring permanent isolation that also is highly radioactive, and we define ''highly radioactive'' as a decay heat (power density) in the waste greater than 50 W/m 3 or an external radiation dose rate at a distance of 1 m from the waste greater than 100 rem/h (1 Sv/h), whichever is the more restrictive. This proposal also results in a definition of Transuranic (TRU) Waste and Equivalent as waste that requires permanent isolation but is not highly radioactive and a definition of low-level waste (LLW) as waste that does not require permanent isolation without regard to whether or not it is highly radioactive

Charging and coagulation of water aerosols with negligible addition of high-radioactive droplets

International Nuclear Information System (INIS)

Vasil'eva, N.L.; Sedova, G.L.; Chernyj, L.T.

1994-01-01

The mechanics of electrocoagulation of water aerosols with negligible admixture of high-radioactive droplets is considered. A corresponding mathematical model has been worked out which describes the processes of ionization, electrification and coagulation of radioactive aerosols. Numerical studies are carried out for a series of typical aerosols on the time dependence of ion concentrations, charge and pure droplet concentrations, as well as the charge and radius of radioactive droplets. It is shown that coagulation can give rise to the growth of droplet radius from 5-10 μm up to 30-40 μm for a 10 4 s period f time, and therefore it can play a considerable role in the development of aerosols with droplet radius up to 20 μm when gravitational coagulation is insignificant

Released radioactivity reducing facility

International Nuclear Information System (INIS)

Tanaka, Takeaki.

1992-01-01

Upon occurrence of a reactor accident, penetration portions of a reactor container, as a main leakage source from a reactor container, are surrounded by a plurality of gas-tight chambers, the outside of which is surrounded by highly gas-tightly buildings. Branched pipelines of an emergency gas processing system are introduced to each of the gas-tight chambers and they are joined and in communication with an emergency gas processing device. With such a constitution, radioactive materials are prevented from leaking directly from the buildings. Further, pipeline openings of the emergency gas processing facility are disposed in the plurality highly gas-tight penetration chambers. If the radioactive materials are leaked from the reactor to elevate the pressure in the penetration chambers, the radioactive materials are introduced to a filter device in the emergency gas processing facility by way of the branched pipelines, filtered and then released to the atmosphere. Accordingly, the reliability and safety of the system can be improved. (T.M.)

High-level radioactive waste glass and storage canister design

International Nuclear Information System (INIS)

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

1979-01-01

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

Gaseous radioactive waste processing system

International Nuclear Information System (INIS)

Onizawa, Hideo.

1976-01-01

Object: To prevent explosion of hydrogen gas within gaseous radioactive waste by removing the hydrogen gas by means of a hydrogen absorber. Structure: A coolant extracted from a reactor cooling system is sprayed by nozzle into a gaseous phase (hydrogen) portion within a tank, thus causing slipping of radioactive rare gas. The gaseous radioactive waste rich in hydrogen, which is purged in the tank, is forced by a waste gas compressor into a hydrogen occlusion device. The hydrogen occlusion device is filled with hydrogen occluding agents such as Mg, Mg-Ni alloy, V-Nb alloy, La-Ni alloy and so forth, and hydrogen in the waste gas is removed through reaction to produce hydrogen metal. The gaseous radioactive waste, which is deprived of hydrogen and reduced in volume, is stored in an attenuation tank. The hydrogen stored in the hydrogen absorber is released and used again as purge gas. (Horiuchi, T.)

Improvement of the IRIS Process for Incineration of Various Radioactive Waste Compositions

International Nuclear Information System (INIS)

Lemort, F.; Charvillat, J. P.

2003-01-01

Incineration represents a promising weight and volume reduction technique for alpha-contaminated organic waste. Following several years of laboratory research initiated in 1983 on a nonradioactive prototype unit at the CEA's Rhone Valley (Marcoule) Research Center, an innovative process, IRIS, has been developed to meet the need for processing nuclear glove box waste containing large amounts of chlorine. In March 1999, the first highly chlorinated alpha-contaminated waste was incinerated in the industrial facility based on the IRIS process at the CEA's Valduc Center. The nonradioactive prototype at Marcoule and the radioactive facility at Valduc demonstrated that the process is highly effective with a continuously fed rotating tubular kiln and with a very effective control of corrosion by pyrolytic decomposition of the waste initially at 550 C. The ash quality meets specification requirements (< 1% carbon, < 1% chlorine) and the volume and weight reduction factors are sufficient (around 30). The offgas treatment system exhibits very high operating efficiency complying with gaseous emission standards

Confidence improvement of disosal safety bydevelopement of a safety case for high-level radioactive waste disposal

Energy Technology Data Exchange (ETDEWEB)

Baik, Min Hoon; Ko, Nak Youl; Jeong, Jong Tae; Kim, Kyung Su [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-12-15

Many countries have developed a safety case suitable to their own countries in order to improve the confidence of disposal safety in deep geological disposal of high-level radioactive waste as well as to develop a disposal program and obtain its license. This study introduces and summarizes the meaning, necessity, and development process of the safety case for radioactive waste disposal. The disposal safety is also discussed in various aspects of the safety case. In addition, the status of safety case development in the foreign countries is briefly introduced for Switzerland, Japan, the United States of America, Sweden, and Finland. The strategy for the safety case development that is being developed by KAERI is also briefly introduced. Based on the safety case, we analyze the efforts necessary to improve confidence in disposal safety for high-level radioactive waste. Considering domestic situations, we propose and discuss some implementing methods for the improvement of disposal safety, such as construction of a reliable information database, understanding of processes related to safety, reduction of uncertainties in safety assessment, communication with stakeholders, and ensuring justice and transparency. This study will contribute to the understanding of the safety case for deep geological disposal and to improving confidence in disposal safety through the development of the safety case in Korea for the disposal of high-level radioactive waste.

Method of processing radioactive solid wastes

International Nuclear Information System (INIS)

Ootaka, Hisashi; Aizu, Tadashi.

1980-01-01

Purpose: To improve the volume-reducing effect for the radioactive solids wastes by freezing and then pulverizing them. Method: Miscellaneous radioactive solid wastes produced from a nuclear power plant and packed in vinyl resin bags are filled in a drum can and nitrogen gas at low temperature (lower than 0 0 C) from a cylinder previously prepared by filling liquid nitrogen (at 15kg/cm 2 , -196 0 C) to freeze the radioactive solid wastes. Thereafter, a hydraulic press is inserted into the drum can to compress and pulverize the thus freezed miscellaneous radioactive solid wastes into powder. The powder thus formed does not expand even after removing the hydraulic press from the drum can, whereby the volume reduction of the radioactive solid wastes can be carried out effectively. (Horiuchi, T.)

Method of volume-reducing processing for radioactive wastes

International Nuclear Information System (INIS)

Sato, Koei; Yamauchi, Noriyuki; Hirayama, Toshihiko.

1985-01-01

Purpose: To process the processing products of radioactive liquid wastes and burnable solid wastes produced from nuclear facilities into stable solidification products by heat melting. Method: At first, glass fiber wastes of contaminated air filters are charged in a melting furnace. Then, waste products obtained through drying, sintering, incineration, etc. are mixed with a proper amount of glass fibers and charged into the melting furnace. Both of the charged components are heated to a temperature at which the glass fibers are melted. The burnable materials are burnt out to provide a highly volume-reduced products. When the products are further heated to a temperature at which metals or metal oxides of a higher melting point than the glass fiber, the glass fibers and the metals or metal oxides are fused to each other to be combined in a molecular structure into more stabilized products. The products are excellent in strength, stability, durability and leaching resistance at ambient temperature. (Kamimura, M.)

Predicting Induced Radioactivity at High Energy Accelerators

Energy Technology Data Exchange (ETDEWEB)

Fasso, Alberto

1999-08-27

Radioactive nuclides are produced at high-energy electron accelerators by different kinds of particle interactions with accelerator components and shielding structures. Radioactivity can also be induced in air, cooling fluids, soil and groundwater. The physical reactions involved include spallations due to the hadronic component of electromagnetic showers, photonuclear reactions by intermediate energy photons and low-energy neutron capture. Although the amount of induced radioactivity is less important than that of proton accelerators by about two orders of magnitude, reliable methods to predict induced radioactivity distributions are essential in order to assess the environmental impact of a facility and to plan its decommissioning. Conventional techniques used so far are reviewed, and a new integrated approach is presented, based on an extension of methods used at proton accelerators and on the unique capability of the FLUKA Monte Carlo code to handle the whole joint electromagnetic and hadronic cascade, scoring residual nuclei produced by all relevant particles. The radiation aspects related to the operation of superconducting RF cavities are also addressed.

Radioactive waste gas processing systems

International Nuclear Information System (INIS)

Kita, Kaoru; Minemoto, Masaki; Takezawa, Kazuaki.

1981-01-01

Purpose: To effectively separate and remove only hydrogen from hydrogen gas-containing radioactive waste gases produced from nuclear power plants without using large scaled facilities. Constitution: From hydrogen gas-enriched waste gases which contain radioactive rare gases (Kr, Xe) sent from the volume control tank of a chemical volume control system, only the hydrogen is separated in a hydrogen separator using palladium alloy membrane and rare gases are concentrated, volume-decreased and then stored. In this case, an activated carbon adsorption device is connected at its inlet to the radioactive gas outlet of the hydrogen separator and opened at its outlet to external atmosphere. In this system, while only the hydrogen gas permeates through the palladium alloy membrane, other gases are introduced, without permeation, into the activated carbon adsorption device. Then, the radioactive rare gases are decayed by the adsorption on the activated carbon and then released to the external atmosphere. (Furukawa, Y.)

Natural diatomite process for removal of radioactivity from liquid waste

Energy Technology Data Exchange (ETDEWEB)

Osmanlioglu, Ahmet Erdal [Radioactive Waste Management Unit (RWMU), Turkish Atomic Energy Authority, Cekmece Nuclear Research and Training Center, Altinsehir Yolu 5 km. Halkali, 34303K Cekmece, Istanbul (Turkey)]. E-mail: Erdal.Osmanlioglu@taek.gov.tr

2007-01-15

Diatomite has a number of unique physical properties and has found diversified industrial utilization. The filtration characteristics are particularly significant in the purification of liquids. The purpose of this study was to test natural diatomaceous earth (diatomite) as an alternative material that could be used for removal of radioactivity from liquid waste. A pilot-scale column-type device was designed. Natural diatomite samples were ground, sieved and prepared to use as sorption media. In this study, real waste liquid was used as radioactive liquid having special conditions. The liquid waste contained three radionuclides (Cs-137, Cs-134 and Co-60). Following the treatment by diatomite, the radioactivity of liquid waste was reduced from the initial 2.60 Bq/ml to less than 0.40 Bq/ml. The results of this study show that most of the radioactivity was removed from the solution by processing with diatomite.

Natural diatomite process for removal of radioactivity from liquid waste

International Nuclear Information System (INIS)

Osmanlioglu, Ahmet Erdal

2007-01-01

Diatomite has a number of unique physical properties and has found diversified industrial utilization. The filtration characteristics are particularly significant in the purification of liquids. The purpose of this study was to test natural diatomaceous earth (diatomite) as an alternative material that could be used for removal of radioactivity from liquid waste. A pilot-scale column-type device was designed. Natural diatomite samples were ground, sieved and prepared to use as sorption media. In this study, real waste liquid was used as radioactive liquid having special conditions. The liquid waste contained three radionuclides (Cs-137, Cs-134 and Co-60). Following the treatment by diatomite, the radioactivity of liquid waste was reduced from the initial 2.60 Bq/ml to less than 0.40 Bq/ml. The results of this study show that most of the radioactivity was removed from the solution by processing with diatomite

Natural diatomite process for removal of radioactivity from liquid waste.

Science.gov (United States)

Osmanlioglu, Ahmet Erdal

2007-01-01

Diatomite has a number of unique physical properties and has found diversified industrial utilization. The filtration characteristics are particularly significant in the purification of liquids. The purpose of this study was to test natural diatomaceous earth (diatomite) as an alternative material that could be used for removal of radioactivity from liquid waste. A pilot-scale column-type device was designed. Natural diatomite samples were ground, sieved and prepared to use as sorption media. In this study, real waste liquid was used as radioactive liquid having special conditions. The liquid waste contained three radionuclides (Cs-137, Cs-134 and Co-60). Following the treatment by diatomite, the radioactivity of liquid waste was reduced from the initial 2.60 Bq/ml to less than 0.40 Bq/ml. The results of this study show that most of the radioactivity was removed from the solution by processing with diatomite.

Chemical precipitation processes for the treatment of aqueous radioactive waste

International Nuclear Information System (INIS)

1992-01-01

Chemical precipitation by coagulation-flocculation and sedimentation has been commonly used for many years to treat liquid (aqueous) radioactive waste. This method allows the volume of waste to be substantially reduced for further treatment or conditioning and the bulk of the waste to de discharged. Chemical precipitation is usually applied in combination with other methods as part of a comprehensive waste management scheme. As with any other technology, chemical precipitation is constantly being improved to reduce cost to increase the effectiveness and safety on the entire waste management system. The purpose of this report is to review and update the information provided in Technical Reports Series No. 89, Chemical Treatment of Radioactive Wastes, published in 1968. In this report the chemical methods currently in use for the treatment of low and intermediate level aqueous radioactive wastes are described and illustrated. Comparisons are given of the advantages and limitations of the processes, and it is noted that good decontamination and volume reduction are not the only criteria according to which a particular process should be selected. Emphasis has been placed on the need to carefully characterize each waste stream, to examine fully the effect of segregation and the importance of looking at the entire operation and not just the treatment process when planning a liquid waste treatment facility. This general approach includes local requirements and possibilities, discharge authorization, management of the concentrates, ICRP recommendations and economics. It appears that chemical precipitation process and solid-liquid separation techniques will continue to be widely used in liquid radioactive waste treatment. Current research and development is showing that combining different processes in one treatment plant can provide higher decontamination factors and smaller secondary waste arisings. Some of these processes are already being incorporated into new and

Radioactive wastes processing device

International Nuclear Information System (INIS)

Takamura, Yoshiyuki; Fukujoji, Seiya.

1986-01-01

Purpose: To exactly recognize the deposition state of mists into conduits thereby effectively conduct cleaning. Constitution: A drier for performing drying treatment of liquid wastes, a steam decontaminating tower for decontaminating the steams generated from the drier and a condenser for condensating the decontaminating steams are connected with each other by means of conduits to constitute a radioactive wastes processing apparatus. A plurality of pressure detectors are disposed to the conduits, the pressure loss within the conduits is determined based on the detector output and the clogged state in the conduits due to the deposition of mists is detected by the magnitude of the pressure loss. If the clogging exceeds a certain level, cleaning water is supplied to clean-up the conduits thereby keep the operation to continue always under sound conditions. (Sekiya, K.)

Measurement of gross beta radioactivity in high-level liquid waste

International Nuclear Information System (INIS)

Lu Feng; Lin Cansheng; Zhang Xianzi; Chen Guoan; Zhang Chonghai

1992-01-01

Using beta plastic scintillation counter of low level background, gross beta radioactivity of twelve samples for high-level liquid waste is determined directly. Beta efficiency curves of plastic scintillation counter for four mass thickness are calibrated in advance. Determining gross beta radioactivity, gross efficiency of the scintillation counter for various energy beta ray is calculated via weighted mean method with the ratio of radioactivity for each nuclide. The ratio of radioactivity for nuclides which have gamma disintegration is determined in terms of the radioactivity measured by gamma spectrometer. The ratio of the radioactivity for 90 Sr which has purity beta disintegration is calculated in terms of half life time approximation. The ratio of the radioactivity for 147 Pm which also has purity disintegration is calculated by means of apparent cooling-time approximation. The uncertainty of results for the present work is about +-15%

Experience base for Radioactive Waste Thermal Processing Systems: A preliminary survey

International Nuclear Information System (INIS)

Mayberry, J.; Geimer, R.; Gillins, R.; Steverson, E.M.; Dalton, D.; Anderson, G.L.

1992-04-01

In the process of considering thermal technologies for potential treatment of the Idaho National Engineering Laboratory mixed transuranic contaminated wastes, a preliminary survey of the experience base available from Radioactive Waste Thermal Processing Systems is reported. A list of known commercial radioactive waste facilities in the United States and some international thermal treatment facilities are provided. Survey focus is upon the US Department of Energy thermal treatment facilities. A brief facility description and a preliminary summary of facility status, and problems experienced is provided for a selected subset of the DOE facilities

Design and fabrication of a glovebox for the Plasma Hearth Process radioactive bench-scale system

International Nuclear Information System (INIS)

Wahlquist, D.R.

1996-01-01

This paper presents some of the design considerations and fabrication techniques for building a glovebox for the Plasma Hearth Process (PHP) radioactive bench-scale system. The PHP radioactive bench-scale system uses a plasma torch to process a variety of radioactive materials into a final vitrified waste form. The processed waste will contain plutonium and trace amounts of other radioactive materials. The glovebox used in this system is located directly below the plasma chamber and is called the Hearth Handling Enclosure (HHE). The HHE is designed to maintain a confinement boundary between the processed waste and the operator. Operations that take place inside the HHE include raising and lowering the hearth using a hydraulic lift table, transporting the hearth within the HHE using an overhead monorail and hoist system, sampling and disassembly of the processed waste and hearth, weighing the hearth, rebuilding a hearth, and sampling HEPA filters. The PHP radioactive bench-scale system is located at the TREAT facility at Argonne National Laboratory-West in Idaho Falls, Idaho

Processing of combustible radioactive waste using incineration techniques

International Nuclear Information System (INIS)

Maestas, E.

1981-01-01

Among the OECD Nuclear Energy Agency Member countries numerous incineration concepts are being studied as potential methods for conditioning alpha-bearing and other types of combustible radioactive waste. The common objective of these different processes is volume reduction and the transformation of the waste to a more acceptable waste form. Because the combustion processes reduce the mass and volume of waste to a form which is generally more inert than the feed material, the resulting waste can be more uniformly compatible with safe handling, packaging, storage and/or disposal techniques. The number of different types of combustion process designed and operating specifically for alpha-bearing wastes is somewhat small compared with those for non-alpha radioactive wastes; however, research and development is under way in a number of countries to develop and improve alpha incinerators. This paper provides an overview of most alpha-incineration concepts in operation or under development in OECD/NEA Member countries. The special features of each concept are briefly discussed. A table containing characteristic data of incinerators is presented so that a comparison of the major programmes can be made. The table includes the incinerator name and location, process type, capacity throughput, operational status and application. (author)

The IAEA's high level radioactive waste management programme

International Nuclear Information System (INIS)

Saire, D.E.

1994-01-01

This paper presents the different activities that are performed under the International Atomic Energy Agency's (IAEA) high level radioactive waste management programme. The Agency's programme is composed of five main activities (information exchange, international safety standards, R ampersand D activities, advisory services and special projects) which are described in the paper. Special emphasis is placed on the RADioactive WAste Safety Standards (RADWASS) programme which was implemented in 1991 to document international consensus that exists on the safe management of radioactive waste. The paper also raises the question about the need for regional repositories to serve certain countries that do not have the resources or infrastructure to construct a national repository

Novel reprocessing methods with nuclide separation for volume reduction of high level radioactive waste

International Nuclear Information System (INIS)

Suzuki, Tatsuya

2015-01-01

We have proposed the reprocessing system with nuclide separation processes based on the chromatographic technique in the hydrochloric acid solution system. Our proposing system consists of the dissolution process, the reprocessing process, the MA separation process, and nuclide separation processes. In our proposing processes, the pyridine resin is used as a main separation media. We expect that our proposing will contribute to that volume reduction of high level radioactive waste by combining the transmutation techniques, usage of valuable elements, and so on. (author)

Some legal aspects on high level radioactive waste disposal in Japan

International Nuclear Information System (INIS)

Tanabe, Tomoyuki

1997-01-01

In Japan, it is considered to be an urgent problem to prepare the system for the research and execution of high level radioactive waste disposal. Under what regulation scheme the disposal should be done has not been sufficiently examined. In this research, the examination was carried out on the legal aspects of high level radioactive waste disposal as follows. First, the current legislation on the disposal in Japan was analyzed, and it was made clear that high level radioactive waste disposal has not been stipulated clearly. Next, on the legal choices which are conceivable on the way the legislation for high level radioactive waste disposal should be, from the aspects of applying the law on regulating nuclear reactors and others, applying the law on nuclear power damage reparation, and industrialization by changing the government ordinances, those were arranged in six choices, and the examination was carried out for each choice from the viewpoints of the relation with the base stipulation for waste-burying business, the speciality of high level radioactive waste disposal as compared with other actions of nuclear power business, the coordination with existing nuclear power of nuclear power business, the coordination with existing nuclear power law system and the formation of national consensus. In this research, it was shown that the execution of high level radioactive waste disposal as the business based on the separate legislation is the realistic choice. (K.I.)

Radioactive waste processing apparatus

Science.gov (United States)

Nelson, R.E.; Ziegler, A.A.; Serino, D.F.; Basnar, P.J.

1985-08-30

Apparatus for use in processing radioactive waste materials for shipment and storage in solid form in a container is disclosed. The container includes a top, and an opening in the top which is smaller than the outer circumference of the container. The apparatus includes an enclosure into which the container is placed, solution feed apparatus for adding a solution containing radioactive waste materials into the container through the container opening, and at least one rotatable blade for blending the solution with a fixing agent such as cement or the like as the solution is added into the container. The blade is constructed so that it can pass through the opening in the top of the container. The rotational axis of the blade is displaced from the center of the blade so that after the blade passes through the opening, the blade and container can be adjusted so that one edge of the blade is adjacent the cylindrical wall of the container, to insure thorough mixing. When the blade is inside the container, a substantially sealed chamber is formed to contain vapors created by the chemical action of the waste solution and fixant, and vapors emanating through the opening in the container. The chamber may be formed by placing a removable extension over the top of the container. The extension communicates with the apparatus so that such vapors are contained within the container, extension and solution feed apparatus. A portion of the chamber includes coolant which condenses the vapors. The resulting condensate is returned to the container by the force of gravity.

ENTRIA 2014. Memorandum on the disposal of high-level radioactive residuals

International Nuclear Information System (INIS)

Roehlig, Klaus-Juergen; Walther, Clemens; Bach, Friedrich-Wilhelm

2014-01-01

The memorandum on the disposal of high-level radioactive residuals covers the following issues: description of the problem: a ''wicked problem'', risks and NIMBY, the site selection law, international boundary conditions; disposal strategy and types of facilities: safety and reversibility, long-term surface storage, deep storage; risk and safety; procedural justice and the site selection process; social innovations and the requirement of long-term institutions; conclusion - central stress fields.

High-level waste processing at the Savannah River Site: An update

International Nuclear Information System (INIS)

Marra, J.E.; Bennett, W.M.; Elder, H.H.; Lee, E.D.; Marra, S.L.; Rutland, P.L.

1997-01-01

The Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS) in Aiken, SC mg began immobilizing high-level radioactive waste in borosilicate glass in 1996. Currently, the radioactive glass is being produced as a ''sludge-only'' composition by combining washed high-level waste sludge with glass frit. The glass is poured in stainless steel canisters which will eventually be disposed of in a permanent, geological repository. To date, DWPF has produced about 100 canisters of vitrified waste. Future processing operations will, be based on a ''coupled'' feed of washed high-level waste sludge, precipitated cesium, and glass frit. This paper provides an update of the processing activities completed to date, operational/flowsheet problems encountered, and programs underway to increase production rates

Short-lived radioactive nuclides in meteorites and early solar system processes

International Nuclear Information System (INIS)

Chaussidon, M.; Gounelle, M.

2007-01-01

Now extinct, short-lived radioactive nuclides, such as 7 Be (T 1/2 = 53 days), 10 Be (T 1/2 = 1.5 Ma), 26 Al (T 1/2 = 0.74 Ma), 36 Cl (T 1/2 = 0.3 Ma), 41 Ca (T 1/2 = 0.1 Ma), 53 Mn (T 1/2 = 3.7 Ma) and 60 Fe (T 1/2 = 1.5 Ma), were present in the proto-solar nebula when the various components of meteorites formed. The presence of these radioactive isotopes requires a 'last-minute' origin, either nucleosynthesis in a massive star dying close in space and time to the nascent solar system or production by local irradiation of part of the proto-solar disk by high-energy solar cosmic rays. In this review, we list: (i) the different observations indicating the existence of multiple origins for short-lived radioactive nuclides, namely 7 Be, 10 Be and 36 Cl for irradiation scenario and 60 Fe for injection scenario; (ii) the constraints that exist on their distribution (homogeneous or heterogeneous) in the accretion disk; (iii) the constraints they brought on the timescales of nebular processes (from Ca-Al-rich inclusions to chondrules) and of the accretion and differentiation of planetesimals. (authors)

Numerical analysis of thermal process in the near field around vertical disposal of high-level radioactive waste

Directory of Open Access Journals (Sweden)

H.G. Zhao

2014-02-01

Full Text Available For deep geological disposal of high-level radioactive waste (HLW in granite, the temperature on the HLW canisters is commonly designed to be lower than 100 °C. This criterion dictates the dimension of the repository. Based on the concept of HLW disposal in vertical boreholes, thermal process in the near field (host rock and buffer surrounding HLW canisters has been simulated by using different methods. The results are drawn as follows: (a the initial heat power of HLW canisters is the most important and sensitive parameter for evolution of temperature field; (b the thermal properties and variations of the host rock, the engineered buffer, and possible gaps between canister and buffer and host rock are the additional key factors governing the heat transformation; (c the gaps width and the filling by water or air determine the temperature offsets between them.

Use of radioactive sodium-22 to study the processes of soil salinization and desalinization

International Nuclear Information System (INIS)

Alzubaidi, A.H.

1979-01-01

This study deals with the salinization of four undisturbed soil columns of silt loam soil, collected with special plexiglass columns. The salinization was effected by adding a certain volume of salt solution consisting of a mixture of NaCl, CaCl 2 and MgCl 2 and containing 0.5 mCi of sodium-22. The salt solution was added to the surface of the first two columns and then the soil columns were leached with distilled water, while for the other two columns, the salt solution was added from the bottom of the columns using a syphon technique. The first two columns represent a model for the desalinization process of saline soils, while the latter two columns represent a model for the salinization process under the effect of high groundwater table. The downward and upward movements of sodium through the soil columns were recorded by measuring sodium radioactivity periodically, using a special scanner which continuously and automatically detected the radioactivity of sodium with the help of a gamma spectrometer. The final distribution curves for sodium movement throughout these soil columns versus time were obtained by computer. The data obtained indicate that radioactive sodium can be used with success to study the movement of salts in soil. The results also bring a new and better understanding of the nature of the salt movement during the processes of salinization and desalinization, the most important soil processes in the arid and semi-arid regions. (author)

Entrapment process of radioactive gaseous wastes

International Nuclear Information System (INIS)

Gagneraud, Francis; Gagneraud, Michel.

1981-01-01

Process for collecting chemically inert gaseous radioactive waste in melted substances, whereby the gaseous waste is injected under pressure in a molten substance to its saturation point followed by fast cooling. This substance is constituted of glass, ceramics, metallurgical drosses and slag masses in fusion. Its cooling is carried out by quenching by means of running water or a gas fluid, or by casting into vessels with great thermal inertia such as cast iron or similar, before recovery and confinement in receptacles for storage [fr

Management of radioactive waste at INR-technical support for processing of radioactive waste from nuclear facilities

International Nuclear Information System (INIS)

Bujoreanu, D.; Popescu, I.V.; Bujoreanu, L.

2009-01-01

The Institute for nuclear research (INR) subsidiary of the Romanian authority for nuclear activities has its own radwaste treatment plant (STDR). STDR is supposed to treat and condition radioactive waste from the nuclear fuel facility, the TRIGA reactor, post irradiation examination laboratories and other research laboratories of NRI. The main steps of waste processing are: pretreatment (collection, characterization, segregation, decontamination)., treatment (waste volume reduction, radionuclide removal, compositional change), conditioning (immobilization and containerization), interim storage of the packages in compliance with safety requirements for the protection of human health and environmental protection, transport of the packages containing radioactive waste, disposal.

Inverse osmotic process for radioactive laundry waste

Energy Technology Data Exchange (ETDEWEB)

Ebara, K; Takahashi, S; Sugimoto, Y; Yusa, H; Hyakutake, H

1977-01-07

Purpose: To effectively recover the processing amount reduced in a continuous treatment. Method: Laundry waste containing radioactive substances discharged from a nuclear power plant is processed in an inverse osmotic process while adding starch digesting enzymes such as amylase and takadiastase, as well as soft spherical bodies such as sponge balls of a particle diameter capable of flowing in the flow of the liquid wastes along the inverse osmotic membrane pipe and having such a softness and roundness as not to damage the inverse osmotic membrane. This process can remove the floating materials such as thread dusts or hairs deposited on the membrane surface by the action of the soft elastic balls and remove paste or the like through decomposition by the digesting enzymes. Consequently, effective recovery can be attained for the reduced processing amount.

Radioactive waste management - a safe solution

International Nuclear Information System (INIS)

1993-01-01

This booklet sets out current United Kingdom government policy regarding radioactive waste management and is aimed at reassuring members of the public concerned about the safety of radioactive wastes. The various disposal or, processing or storage options for low, intermediate and high-level radioactive wastes are explained and sites described, and the work of the Nuclear Industry Radioactive Waste Executive (NIREX) is outlined. (UK)

Method of processing solidification product of radioactive waste

International Nuclear Information System (INIS)

Daime, Fumiyoshi.

1988-01-01

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

Thermo-Hydro Mechanical Characteristics and Processes in the Clay Barrier of a High Level Radioactive Waste Repository. State of the Art Report

International Nuclear Information System (INIS)

Villar, M. V.

2004-01-01

This document is a summary of the available information on the thermo-hydro-mechanical properties of the bentonite barrier of a high-level radioactive waste repository and of the processes taking place in it during the successive repository operation phases. Mainly the thermal properties, the volume change processes (swelling and consolidation), the permeability and the water retention capacity are analysed. A review is made of the existing experimental knowledge on the modification of the these properties by the effect of temperature, water salinity, humidity and density of the bentonite, and their foreseen evolution as a consequence of the processes expected in the repository. The compiled evolution refers mostly to the FEBEX (Spain), the MX-80 (USA) and the FoCa (France) bentonite, considered as reference barrier materials in several European disposal concepts. (Author) 102 refs

Thermo-Hydro Mechanical Characteristics and Processes in the Clay Barrier of a High Level Radioactive Waste Repository. State of the Art Report

Energy Technology Data Exchange (ETDEWEB)

Villar, M. V.

2004-07-01

This document is a summary of the available information on the thermo-hydro-mechanical properties of the bentonite barrier of a high-level radioactive waste repository and of the processes taking place in it during the successive repository operation phases. Mainly the thermal properties, the volume change processes (swelling and consolidation), the permeability and the water retention capacity are analysed. A review is made of the existing experimental knowledge on the modification of the these properties by the effect of temperature, water salinity, humidity and density of the bentonite, and their foreseen evolution as a consequence of the processes expected in the repository. The compiled evolution refers mostly to the FEBEX (Spain), the MX-80 (US) and the FoCa (France) bentonite, considered as reference barrier materials in several European disposal concepts. (Author) 102 refs.

Experimental study on intermediate level radioactive waste processing

International Nuclear Information System (INIS)

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

1977-01-01

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

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

PROCESS OF DECONTAMINATING MATERIAL CONTAMINATED WITH RADIOACTIVITY

Science.gov (United States)

Overholt, D.C.; Peterson, M.D.; Acken, M.F.

1958-09-16

A process is described for decontaminating metallic objects, such as stainless steel equipment, which consists in contacting such objects with nltric acid in a concentration of 35 to 60% to remove the major portion of the contamination; and thereafter contacting the partially decontaminated object with a second solution containing up to 20% of alkali metal hydroxide and up to 20% sodium tartrate to remove the remaining radioactive contaminats.

Michigan high-level radioactive waste program. Technical progress report for 1985

International Nuclear Information System (INIS)

1986-01-01

In 1985, five crystalline rock formations located in Michigan's Upper Peninsula were under consideration in the regional phase of the Department of Energy's (DOE) search for the site of the nation's second high-level radioactive waste repository. The Michigan Department of Public Health has been designated by the Governor as lead state agency in matters related to high-level radioactive waste (HLRW). Mr. Lee E. Jager, Chief of the Department's Bureau of Environmental and Occupational Health, has been designated as the state contact person in this matter, and the Bureau's Division of Radiological Health, Office of Radioactive Waste Management (ORWM), has been designated to provide staff support. Recognizing that adequate state involvement in the various aspects of the Federal high-level radioactive waste (HLRW) programs would require a range of expertise beyond the scope of any single state agency, Governor Blanchard established the High-Level Radioactive Waste Task Force in 1983. In support of the Task Force efforts concerning the implementation of its change, the Department negotiated and concluded an agreement with the DOE, under which federal funds are provided to support state HLRW activities. This report outlines state activities for the calendar year 1985, funded under that agreement

Risk communication system for high level radioactive waste disposal

International Nuclear Information System (INIS)

Kugo, Akihide; Uda, Akinobu; Shimoda, Hirosi; Yoshikawa, Hidekazu; Ito, Kyoko; Wakabayashi, Yasunaga

2005-01-01

In order to gain a better understanding and acceptance of the task of implementing high level radioactive waste disposal, a study on new communication system about social risk information has been initiated by noticing the rapid expansion of Internet in the society. First, text mining method was introduced to identify the core public interest, examining public comments on the technical report of high level radioactive waste disposal. Then we designed the dialog-mode contents based on the theory of norm activation by Schwartz. Finally, the discussion board was mounted on the web site. By constructing such web communication system which includes knowledge base contents, introspective contents, and interactive discussion board, we conducted the experiment for verifying the principles such as that the basic technical knowledge and trust, and social ethics are indispensable in this process to close the perception gap between nuclear specialists and the general public. The participants of the experiment increased their interest in the topics with which they were not familiar and actively posted their opinions on the BBS. The dialog-mode contents were significantly more effective than the knowledge-based contents in promoting introspection that brought people into a greater awareness of problems such as social dilemma. (author)

Heat transfer enhanced microwave process for stabilization of liquid radioactive waste slurry. Final report

International Nuclear Information System (INIS)

White, T.L.

1995-01-01

The objectve of this CRADA is to combine a polymer process for encapsulation of liquid radioactive waste slurry developed by Monolith Technology, Inc. (MTI), with an in-drum microwave process for drying radioactive wastes developed by Oak Ridge National Laboratory (ORNL), for the purpose of achieving a fast, cost-effectve commercial process for solidification of liquid radioactive waste slurry. Tests performed so far show a four-fold increase in process throughput due to the direct microwave heating of the polymer/slurry mixture, compared to conventional edge-heating of the mixer. We measured a steady-state throughput of 33 ml/min for 1.4 kW of absorbed microwave power. The final waste form is a solid monolith with no free liquids and no free particulates

Method of processing radioactive liquid wastes

Energy Technology Data Exchange (ETDEWEB)

Sugimoto, Y; Kikuchi, M; Funabashi, K; Yusa, H; Horiuchi, S

1978-12-21

Purpose: To decrease the volume of radioactive liquid wastes essentially consisting of sodium hydroxide and boric acid. Method: The concentration ratio of sodium hydroxide to boric acid by weight in radioactive liquid wastes essentially consisting of sodium hydroxide and boric acid is adjusted in the range of 0.28 - 0.4 by means of a pH detector and a sodium concentration detector. Thereafter, the radioactive liquid wastes are dried into powder and then discharged.

Radioactive waste treatment and handling in France

International Nuclear Information System (INIS)

Sivintsev, Yu.V.

1984-01-01

Classification of radioactive wastes customary in France and the program of radiation protection in handling them are discussed. Various methods of radioactive waste processing and burial are considered. The French classification of radioactive wastes differs from one used in the other countries. Wastes are classified under three categories: A, B and C. A - low- and intermediate-level radioactive wastes with short-lived radionuclides (half-life - less than 30 years, negligible or heat release, small amount of long-lived radionuclides, especially such as plutonium, americium and neptunium); B - low- and intermediate-level radioactive wastes with long-lived radionuclides (considerable amounts of long-lived radionuclides including α-emitters, low and moderate-level activity of β- and γ-emitters, low and moderate heat release); C - high-level radioactive wastes with long-lived radionuclides (high-level activity of β- and γ-emitters, high heat release, considerable amount of long-lived radionuclides). Volumetric estimations of wastes of various categories and predictions of their growth are given. It is noted that the concept of closed fuel cycle with radiochemical processing of spent fuel is customary in France

ONDRAF/NIRAS and high-level radioactive waste management in Belgium

International Nuclear Information System (INIS)

Decamps, F.

1993-01-01

The National Agency for Radioactive Waste and Enriched Fissile Materials, ONDRAF/NIRAS, is a public body with legal personality in charge of managing all radioactive waste on Belgian territory, regardless of its origin and source. It is also entrusted with tasks related to the management of enriched fissile materials, plutonium containing materials and used or unused nuclear fuel, and with certain aspects of the dismantling of closed down nuclear facilities. High-level radioactive waste management comprises essentially and for the time being the storage of high-level liquid waste produced by the former EUROCHEMIC reprocessing plant and of high-level and very high-level heat producing waste resulting from the reprocessing in France of Belgian spent fuel, as well as research and development (R and D) with regard to geological disposal in clay of this waste type

Fate and transport processes controlling the migration of hazardous and radioactive materials from the Area 5 Radioactive Waste Management Site (RWMS)

International Nuclear Information System (INIS)

Estrella, R.

1994-10-01

Desert vadose zones have been considered as suitable environments for the safe and long-term isolation of hazardous wastes. Low precipitation, high evapotranspiration and thick unsaturated alluvial deposits commonly found in deserts make them attractive as waste disposal sites. The fate and transport of any contaminant in the subsurface is ultimately determined by the operating retention and transformation processes in the system and the end result of the interactions among them. Retention (sorption) and transformation are the two major processes that affect the amount of a contaminant present and available for transport. Retention processes do not affect the total amount of a contaminant in the soil system, but rather decrease or eliminate the amount available for transport at a given point in time. Sorption reactions retard the contaminant migration. Permanent binding of solute by the sorbent is also possible. These processes and their interactions are controlled by the nature of the hazardous waste, the properties of the porous media and the geochemical and environmental conditions (temperature, moisture and vegetation). The present study summarizes the available data and investigates the fate and transport processes that govern the migration of contaminants from the Radioactive Waste Management Site (RWMS) in Area 5 of the Nevada Test Site (NTS). While the site is currently used only for low-level radioactive waste disposal, past practices have included burial of material now considered hazardous. Fundamentals of chemical and biological transformation processes are discussed subsequently, followed by a discussion of relevant results

Fate and transport processes controlling the migration of hazardous and radioactive materials from the Area 5 Radioactive Waste Management Site (RWMS)

Energy Technology Data Exchange (ETDEWEB)

Estrella, R.

1994-10-01

Desert vadose zones have been considered as suitable environments for the safe and long-term isolation of hazardous wastes. Low precipitation, high evapotranspiration and thick unsaturated alluvial deposits commonly found in deserts make them attractive as waste disposal sites. The fate and transport of any contaminant in the subsurface is ultimately determined by the operating retention and transformation processes in the system and the end result of the interactions among them. Retention (sorption) and transformation are the two major processes that affect the amount of a contaminant present and available for transport. Retention processes do not affect the total amount of a contaminant in the soil system, but rather decrease or eliminate the amount available for transport at a given point in time. Sorption reactions retard the contaminant migration. Permanent binding of solute by the sorbent is also possible. These processes and their interactions are controlled by the nature of the hazardous waste, the properties of the porous media and the geochemical and environmental conditions (temperature, moisture and vegetation). The present study summarizes the available data and investigates the fate and transport processes that govern the migration of contaminants from the Radioactive Waste Management Site (RWMS) in Area 5 of the Nevada Test Site (NTS). While the site is currently used only for low-level radioactive waste disposal, past practices have included burial of material now considered hazardous. Fundamentals of chemical and biological transformation processes are discussed subsequently, followed by a discussion of relevant results.

Study on the development of safety regulations for geological disposal of high-level radioactive waste

International Nuclear Information System (INIS)

Wei Fangxin

2012-01-01

The development of regulations under Regulations on Safety Management of Radioactive Waste has become necessary as the issuance of it. The regulations related to geological disposal of high-level radioactive waste can promote the progress of research and development on geological disposal of high-level radioactive waste in China. This paper has present suggestions on development of regulations on geological disposal of high-level radioactive waste by analyzing development of safety regulations on geological disposal of high-level radioactive waste in foreign countries and problems occurred in China and discussed important issues related to the development of safety regulations on geological disposal of high-level radioactive waste. (author)

Radioactive waste disposal assessment - overview of biosphere processes and models

International Nuclear Information System (INIS)

Coughtrey, P.J.

1992-09-01

This report provides an overview of biosphere processes and models in the general context of the radiological assessment of radioactive waste disposal as a basis for HMIP's response to biosphere aspects of Nirex's submissions for disposal of radioactive wastes in a purpose-built repository at Sellafield, Cumbria. The overview takes into account published information from the UK as available from Nirex's safety and assessment research programme and HMIP's disposal assessment programme, as well as that available from studies in the UK and elsewhere. (Author)

Astrophysical r- and rp-processes, and radioactive nuclear beams

International Nuclear Information System (INIS)

Boyd, Richard N.

1998-01-01

The modern description of the r-process follows naturally from α-rich freezeout, thought to occur in the hot neutrino wind just beyond the nascent neutron star in a type II supernova. Initially, all pre-existing nuclei are reduced to α-particles and neutrons. As the environment cools, nuclei up to about mass 90 to 100 u are synthesized, in nuclear statistical equilibrium, in about 1 s. In the next few seconds, the remaining neutrons are captured to form the r-process progenitors, which then decay to the r-process nuclides. The rp-process occurs in a high-temperature H-rich environment. It is one of the processes that synthesize the p-process nuclei, the most neutron-poor nuclei in the periodic table. It is thought to occur during the explosion of a C-O white dwarf in a type Ia supernova or in a binary system during accretion onto a white dwarf or a neutron star. It appears to be capable of forming the p-nuclei up to about mass 90 u. Both processes pass through nuclei that are far from stability. Thus, their description requires the masses, half-lives, decay modes, and structure of these nuclei. The next generation of radioactive beam facilities promises to allow the study of many such nuclei. (author)

Inverse osmotic process for radioactive laundry waste

International Nuclear Information System (INIS)

Ebara, Katsuya; Takahashi, Sankichi; Sugimoto, Yoshikazu; Yusa, Hideo; Hyakutake, Hiroshi.

1977-01-01

Purpose: To effectively recover the processing amount reduced in a continuous treatment. Method: Laundry waste containing radioactive substances discharged from a nuclear power plant is processed in an inverse osmotic process while adding starch digesting enzymes such as amylase and takadiastase, as well as soft spherical bodies such as sponge balls of a particle diameter capable of flowing in the flow of the liquid wastes along the inverse osmotic membrane pipe and having such a softness and roundness as not to damage the inverse osmotic membrane. This process can remove the floating materials such as thread dusts or hairs deposited on the membrane surface by the action of the soft elastic balls and remove paste or the like through decomposition by the digesting enzymes. Consequently, effective recovery can be attained for the reduced processing amount. (Furukawa, Y.)

THE USE OF POLYMERS IN RADIOACTIVE WASTE PROCESSING SYSTEMS

Energy Technology Data Exchange (ETDEWEB)

Skidmore, E.; Fondeur, F.

2013-04-15

The Savannah River Site (SRS), one of the largest U.S. Department of Energy (DOE) sites, has operated since the early 1950s. The early mission of the site was to produce critical nuclear materials for national defense. Many facilities have been constructed at the SRS over the years to process, stabilize and/or store radioactive waste and related materials. The primary materials of construction used in such facilities are inorganic (metals, concrete), but polymeric materials are inevitably used in various applications. The effects of aging, radiation, chemicals, heat and other environmental variables must therefore be understood to maximize service life of polymeric components. In particular, the potential for dose rate effects and synergistic effects on polymeric materials in multivariable environments can complicate compatibility reviews and life predictions. The selection and performance of polymeric materials in radioactive waste processing systems at the SRS are discussed.

Corrosion of Alloy 690 process pot by sulfate containing high level radioactive waste at feed stage

International Nuclear Information System (INIS)

Sengupta, P.; Soudamini, N.; Kaushik, C.P.; Jagannath; Mishra, R.K.; Kale, G.B.; Raj, K.; Das, D.; Sharma, B.P.

2008-01-01

Prolonged exposure of Alloy 690 process pot to sulfate containing high level radioactive waste leads to (a) depletion of Cr from the alloy, (b) intergranular attack and (c) building up of Cr 2 O 3 -Ni 2 O 3 -Fe 2 O 3 mixed oxide surface layer containing Na and Cs sulfate precipitates. Time dependence of material loss from Alloy 690 is found to follow a linear relationship of the type Δw (material loss) = -7.05 + 0.05t. Corrosion rate calculated for 2400 h exposure is 3.66 mpy. Cr and Ni leach rates obtained for the same sample are 1.61 g m -2 d -1 and 2.52 g m -2 d -1 , respectively. Ni leach rates followed a linear time dependence relationship of the type dNL Ni /dt (leach rate) = -0.09 + 0.027t, whereas Cr leach rates obeyed a non-linear relationship of the type dNL Cr /dt (leach rate) = 0.241 + 0.027t - 1.33 x 10 -4 t 1/2

Method of processing radioactive liquid wastes by using zeolites

Energy Technology Data Exchange (ETDEWEB)

Kanno, T; Mimura, H

1975-09-18

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

Radioactive waste processing method for a nuclear power plant

Energy Technology Data Exchange (ETDEWEB)

Sugimoto, Y; Kuriyama, O

1976-06-04

Object is to subject radioactive liquid waste in a nuclear power plant to reverse permeation process after which it is vaporized and concentrated thereby decreasing the quantity of foam to be used to achieve effective concentration of the liquid waste. Liquid waste containing a radioactive material produced from a nuclear power plant is first applied with pressure in excess of osmotic pressure by a reverse permeation device and is separated into clean water and concentrated liquid by semi-permeable membrane. Next, the thus reverse-permeated and concentrated waste is fed to an evaporator which control foaming by the foam and then further reconcentrated for purification of the liquid waste.

Low-level radioactive waste from rare metals processing facilities

International Nuclear Information System (INIS)

Eng, J.; Hendricks, D.W.; Feldman, J.; Giardina, P.A.

1980-01-01

This paper reviews the situations at the existing Teledyne Wah Chang Co., Inc. located at Albany, Oregon, and the former Carborundum Corp./Amax Specialty Metals, Inc., facilities located at Parkersburg, West Virginia, and Akron, New York, in order to show the extent of the radioactivity problem at rare metals processing facilities and the need to identify for radiological review other rare metal and rare earth processing sites

Feed Basis for Processing Relatively Low Radioactivity Waste Tanks

International Nuclear Information System (INIS)

Pike, J.A.

2002-01-01

This paper presents the characterization of potential feed for processing relatively low radioactive waste tanks. The feed characterization is based on waste characterization data extracted from the waste characterization system. This data is compared to salt cake sample results from Tanks 37, 38 and 41

Remote detection of radioactive material using high-power pulsed electromagnetic radiation.

Science.gov (United States)

Kim, Dongsung; Yu, Dongho; Sawant, Ashwini; Choe, Mun Seok; Lee, Ingeun; Kim, Sung Gug; Choi, EunMi

2017-05-09

Remote detection of radioactive materials is impossible when the measurement location is far from the radioactive source such that the leakage of high-energy photons or electrons from the source cannot be measured. Current technologies are less effective in this respect because they only allow the detection at distances to which the high-energy photons or electrons can reach the detector. Here we demonstrate an experimental method for remote detection of radioactive materials by inducing plasma breakdown with the high-power pulsed electromagnetic waves. Measurements of the plasma formation time and its dispersion lead to enhanced detection sensitivity compared to the theoretically predicted one based only on the plasma on and off phenomena. We show that lower power of the incident electromagnetic wave is sufficient for plasma breakdown in atmospheric-pressure air and the elimination of the statistical distribution is possible in the presence of radioactive material.

Some historical background to the IAEA Definition and Recommendations concerning high-level radioactive wastes or other high-level radioactive matter unsuitable for dumping at sea

International Nuclear Information System (INIS)

Nishiwaki, Y.

1981-01-01

The need for internationally acceptable standards and regulations for preventing pollution of the sea by radioactive materials was recognized by the United Nations Conference on the Law of the Sea, which adopted the Convention on the High Seas in April 1958. Article 25 of the Convention provides that ''every State shall take measures to prevent pollution of the seas from the dumping of radioactive wastes, taking into account any standards and regulations which may be formulated by the competent international organizations.'' The Conference also adopted a resolution recommending that the IAEA pursue studies and take action to assist States in controlling the discharge of radioactive materials into the sea. When the Convention on the Prevention of Marine Pollution by Dumping of Wastes and Other Matter was adopted at the Intergovernmental Conference in London, 1972, the IAEA was given specific responsibilities to define criteria and standards for dealing with the questions of sea disposal of radioactive wastes. The IAEA Definition and Recommendations concerning ''high-level radioactive wastes or other high-level radioactive matter unsuitable for dumping at sea'' identify material, the radioactive content of which is at such a level that the Parties to the Convention would wish to prevent any participating State from issuing a special permit even after a detailed appraisal of the safety of the proposed operation, and even for the sector of the marine environment furthest removed from man, i.e. the deep sea with depth greater than 4000 m. Some historical background to these problems is discussed and some of the Japanese findings of the deep sea survey in the Pacific are introduced for comparison with the North Atlantic data which formed a basis of the IAEA Definition and Recommendations for the London Dumping Convention

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

High-level radioactive waste disposal type and theoretical analyses

International Nuclear Information System (INIS)

Lu Yingfa; Wu Yanchun; Luo Xianqi; Cui Yujun

2006-01-01

Study of high-level radioactive waste disposal is necessary for the nuclear electrical development; the determination of nuclear waste depository type is one of importance safety. Based on the high-level radioactive disposal type, the relative research subjects are proposed, then the fundamental research characteristics of nuclear waste disposition, for instance: mechanical and hydraulic properties of rock mass, saturated and unsaturated seepage, chemical behaviors, behavior of special soil, and gas behavior, etc. are introduced, the relative coupling equations are suggested, and a one dimensional result is proposed. (authors)

Status of Closure Welding Technology of Canister for Transportation and Storage of High Level Radioactive Material and Waste

International Nuclear Information System (INIS)

Lee, H. J.; Bang, K. S.; Seo, K. S.; Seo, C. S.

2010-10-01

Closure seal welding is one of the key technologies in fabricating and handling the canister which is used for transportation and storage of high radioactive material and waste. Simple industrial fabrication processes are used before filling the radioactive waste into the canister. But, automatic and remote processes should be used after filling the radioactive material because the thickness of canister is not sufficient to shield the high radiation from filled material or waste. In order to simplify the welding process the closure structure of canister and the sealing method are investigated and developed properly. Two types of radioactive materials such as vitrified waste and compacted solid waste are produced in nuclear industry. Because the filling method of two types of waste is different, the shapes of closure and opening of canister and welding method is also different. The canister shape and sealing method should be standardized to standardize the handling facilities and inspection process such as leak test after closure welding. In order to improve the productivity of disposal and compatibility of the canister, the structure and shape of canister should be standardized considering the type of waste. Two kind of welding process such as arc welding and resistance welding are reported and used in the field. In the arc welding process GTAW and PAW are considered proper processes for closure welding. The closure seal welding process can be selected by considering material of canister, thickness of body, productivity, and applicable codes and rules. Because the storage time of nuclear waste in canister is very long, at least 20 years, the long-time corrosion at the weld should be estimated including mechanical integrity. Recently, the mitigation of residual stress around weld region, which causes stress corrosion cracking, is also interesting research issue

A proposed classification system for high-level and other radioactive wastes

International Nuclear Information System (INIS)

Kocher, D.C.; Croff, A.G.

1989-01-01

On the basis of the definition of high-level wastes (HLW) in the Nuclear Waste Policy Act of 1982 and previous descriptions of reprocessing wastes, a definition is proposed based on the concept that HLW is any waste which is highly radioactive and requires permanent isolation. This conceptual definition of HLW leads to a two-dimensional waste classification system in which one axis, related to 'highly radioactive', is associated with shorter-term risks from waste management and disposal due to high levels of decay heat and external radiation, and the other axis, related to 'requires permanent isolation', is associated with longer-term risks from waste disposal. Wastes that are highly radioactive are defined quantitatively as wastes with a decay heat (power density) greater than 50 W/m 3 or an external dose-equivalent rate greater than 100 rem/h (1 Sv/h) at a distance of 1 m from the waste, whichever is more restrictive. Wastes that require permanent isolation are defined quantitatively as wastes with concentrations of radionuclides greater than the Class-C limits that are generally acceptable for near-surface land disposal, as obtained from the Nuclear Regulatory Commission's 10 CFR Part 61 and its associated methodology. This proposal leads to similar definitions of two other waste classes: transuranic (TRU) waste and equivalent is any waste that requires permanent isolation but is not highly radioactive; and low-level waste (LLW) is any waste that does not require permanent isolation, without regard to whether or not it is highly radioactive. 31 refs.; 3 figs.; 4 tabs

An innovative approach to solid Low Level Radioactive Waste processing and disposal

International Nuclear Information System (INIS)

Pancake, D.C. Jr.; Sodaro, M.A.

1994-01-01

This paper will focus on a new system of Low Level Radioactive Waste (LLW) accumulation, processing and packaging, as-well as the implementation of a Laboratory-wide training program used to introduce new waste accumulation containers to all of the on-site radioactive waste generators, and to train them on the requirements of this innovative waste characterization and documentation program

Method of processing radioactive liquid waste

International Nuclear Information System (INIS)

Motojima, Kenji; Kawamura, Fumio.

1981-01-01

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

Radioactive waste processing device

International Nuclear Information System (INIS)

Seki, Shuji.

1992-01-01

Liquid wastes are supplied to a ceramic filter to conduct filtration. In this case, a device for adding a powdery inorganic ion exchanger is disposed to the upstream of the ceramic filter. When the powdery inorganic ion exchanger is charged to the addition device, it is precoated to the surface of the ceramic filter, to conduct separation of suspended matters and separation of ionic nuclides simultaneously. Liquid wastes returned to a collecting tank are condensed while being circulated between the ceramic filter and the tank and then contained in a condensation liquid waste tank. With such a constitution, both of radioactive nuclides accompanied by suspended matters in the radioactive liquid wastes and ionic nuclides can be captured efficiently. (T.M.)

Separation of actinides and long-lived fission products from high-level radioactive wastes (a review)

International Nuclear Information System (INIS)

Kolarik, Z.

1991-11-01

The management of high-level radioactive wastes is facilitated, if long-lived and radiotoxic actinides and fission products are separated before the final disposal. Especially important is the separation of americium, curium, plutonium, neptunium, strontium, cesium and technetium. The separated nuclides can be deposited separately from the bulk of the high-level waste, but their transmutation to short-lived nuclides is a muchmore favourable option. This report reviews the chemistry of the separation of actinides and fission products from radioactive wastes. The composition, nature and conditioning of the wastes are described. The main attention is paid to the solvent extraction chemistry of the elements and to the application of solvent extraction in unit operations of potential partitioning processes. Also reviewed is the behaviour of the elements in the ion exchange chromatography, precipitation, electrolysis from aqueous solutions and melts, and the distribution between molten salts and metals. Flowsheets of selected partitioning processes are shown and general aspects of the waste partitioning are shortly discussed. (orig.) [de

Conceptual process for conversion of high level waste to glass

International Nuclear Information System (INIS)

1975-01-01

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

On the data processing related to environmental radioactivity

International Nuclear Information System (INIS)

Nakamura, Isamu

1984-01-01

The monitoring and measurement of environmental radioactivity have been undertaken by many organizations in Japan. The Japan Chemical Analysis Center has been entrusted by the government to gather and edit these measured results. The computer processing of these data started in 1978, and it is expected that by the end of fiscal year 1984, all fallout data since 1957 and all radioactivity monitoring data since 1969 can be registered. The computer programs for processing the data such as the output of tables and figures have also been developed, and the edition of reports has been made. The replacement of the computer and the development of a new processing system capable of handing Kanji (Japanese-Chinese characters) are now scheduled. This document outlines the data system such as the quality, quantity and origin of the measured data and the frequency of report publication. Some results of the analysis of fallout nuclides, space gamma dose rate and the total β-activity in rain are presented. The effects of the nuclear explosion tests in China are very obvious in these figures. A chronological table of the explosion tests in China is also presented. The different effects of time lag at the different places of measurement are also seen. The effects of the presence of nuclear power plants were also investigated at some sites of the plants, but no discernible effect was observed. (Aoki, K.)

A process for the bituminization of radioactive wastes

International Nuclear Information System (INIS)

Bar-Ziv, E.; Manor, D.; Levin, I.

1977-06-01

A process is described for the incorporation of sodium nitrate solutions in bitumen, with a view to the bituminization of radioactive wastes. This process has involved the development of a chemical process and its technological application. A double jacket reactor equipped with an anchor-type stirrer has been built. Sodium nitrate (NaNO 3 ) solutions with sodium hydroxide (NaOH) are fed into the reactor simultaneously with a bitumen emulsion while heated oil is circulated through the double jacket. After all the water has evaporated, the mixture is ''simmered'' at 170/180 deg C for up to 5 hours in order to get a homogeneous product which will be suitable for final storage. The product contains small salt crystals (10/50 μm) and has high viscosity (200/2000 poise at 160 deg C and 12 rpm), so that the possibility of sedimentation is eliminated. In comparative experiments with molten bitumen, large salt aggregates, inhomogeneously distributed, make the product unsuitable. It has been found that NaOH is responsible for the polimerization of the bitumen, which raises its viscosity. The degree of polimerization depends on the NaOH concentration and ''simmering'' time and temperature. NaOH is added to NaNO 3 in order to raise the mixture's viscosity and thereby prevent sedimentation

Study into an organization for collecting, processing and removing of radioactive waste

International Nuclear Information System (INIS)

1983-09-01

This report presents the results of a study into a new organization for the collection, processing and removal of radioactive waste. At present these activities are carried out by the Dutch Energy Research Foundation (ECN). The new organization has to offer guarantees for a qualititatively responsible retrieval and processing of radioactive waste. It also has to be certain that the waste offered will not be send back, or even refused, if stagnation occurs in the removal. Finally the tariffs have to be not so prohibitive that they hinder a responsible handling with radioactive waste by the producers. An organization is advised which is self-employed with regard to management, directorate and materials. It is recommended to submit this organization in a limited liability company. This form of government may be supplemented optionally with a slight form of a cooperative association. (author). 10 refs.; 3 figs.; 11 tabs

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

International Nuclear Information System (INIS)

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

1981-08-01

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

Processing method for radioactive liquid waste

International Nuclear Information System (INIS)

Yasumura, Keijiro

1991-01-01

Drainages, such as water after used for washing operators' clothes and water used for washing hands and for showers have such features that the radioactive concentration is extremely low and detergent ingredients and insoluble ingredients such as waste threads, hairs and dirts are contained. At present, waste threads are removed by a strainer. Then, after measuring the radioactivity and determining that the radioactivity is less than a predetermined concentration, they are released to circumstances. However, various organic ingredients such as detergents and dirts in the liquid wastes are released as they are and it is not preferred in respect of environmental protection. Then, in the present invention, activated carbon is filled in a container orderly so that the diameter of the particles of the activated carbon is increased in the upper layer and decreased in the lower layer, and radioactive liquid wastes are passed through the container. With such a constitution. Both of soluble substances and insoluble substances can be removed efficiently without causing cloggings. (T.M.)

Hydrothermal processing of radioactive combustible waste

International Nuclear Information System (INIS)

Worl, L.A.; Buelow, S.J.; Harradine, D.; Le, L.; Padilla, D.D.; Roberts, J.H.

1998-01-01

Hydrothermal processing has been demonstrated for the treatment of radioactive combustible materials for the US Department of Energy. A hydrothermal processing system was designed, built and tested for operation in a plutonium glovebox. Presented here are results from the study of the hydrothermal oxidation of plutonium and americium contaminated organic wastes. Experiments show the destruction of the organic component to CO 2 and H 2 O, with 30 wt.% H 2 O 2 as an oxidant, at 540 C and 46.2 MPa. The majority of the actinide component forms insoluble products that are easily separated by filtration. A titanium liner in the reactor and heat exchanger provide corrosion resistance for the oxidation of chlorinated organics. The treatment of solid material is accomplished by particle size reduction and the addition of a viscosity enhancing agent to generate a homogeneous pumpable mixture

Spanish high level radioactive waste management system issues

International Nuclear Information System (INIS)

Espejo, J.M.; Beceiro, A.R.

1992-01-01

The Empresa Nacional de Residuos Radiactivos, S.A. (ENRESA) has been limited liability company to be responsible for the management of all kind of radioactive wastes in Spain. This paper provides an overview of the strategy and main lines of action stated in the third General Radioactive Waste Plan, currently in force, for the management of spent nuclear fuel and high - level wastes, as well as an outline of the main related projects, either being developed or foreseen. Aspects concerning the organizational structure, the economic and financing system and the international cooperation are also included

Measurement of cesium emissions during the vitrification of simulated high level radioactive waste

International Nuclear Information System (INIS)

Zamecnik, J.R.; Miller, D.H.; Carter, J.T.

1992-01-01

In the Defense Waste Processing Facility at the Savannah River Site, it is desired to eliminate a startup test that would involve adding small amounts of radioactive cesium-137 to simulated high-level waste. In order to eliminate this test, a reliable method for measuring non-radioactive cesium in the offgas system from the glass melter is required. From a pilot scale melter system, offgas particulate samples were taken on filter paper media and analyzed by Inductively Coupled Plasma-Mass Spectrometry (ICP-MS). The ICPMS method proved to be sufficiently sensitive to measure cesium quantities as low as 0.135 μg, with the sensitivity being limited by the background cesium present in the filter paper. Typical particulate loadings ranged from 800 μg of cesium. This sensitivity allowed determination of cesium decontamination factors for four of the five major components of the offgas system. The decontamination factors measured experimentally compared favorably with the process design basis values

Method of processing radioactive gas

International Nuclear Information System (INIS)

Saito, Masayuki.

1978-01-01

Purpose: To reduce the quantity of radioactive gas discharged at the time of starting a nuclear power plant. Method: After the stoppage of a nuclear power plant air containing a radioactive gas is extracted from a main condenser by operating an air extractor. The air is sent into a gaseous waste disposal device, and then introduced into the activated carbon adsorptive tower of a rare gas holdup device where xenon and krypton are trapped. Thereafter, the air passes through pipelines and returned to the main condenser. In this manner, the radioactive gas contained in air within the main condenser is removed during the stoppage of the operation of the nuclear power plant. After the plant has been started, when it enters the normal operation, a flow control valve is closed and another valve is opened, and a purified gas exhausted from the rare gas holdup device is discharged into the atmosphere through an exhaust cylinder. (Aizawa, K.)

Options for the disposal of high-level radioactive waste

International Nuclear Information System (INIS)

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

1977-01-01

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

THOREX processing and zeolite transfer for high-level waste stream processing blending

International Nuclear Information System (INIS)

Kelly, S. Jr.; Meess, D.C.

1997-07-01

The West Valley Demonstration Project (WVDP) completed the pretreatment of the high-level radioactive waste (HLW) prior to the start of waste vitrification. The HLW originated form the two million liters of plutonium/uranium extraction (PUREX) and thorium extraction (THOREX) wastes remaining from Nuclear Fuel Services' (NFS) commercial nuclear fuel reprocessing operations at the Western New York Nuclear Service Center (WNYNSC) from 1966 to 1972. The pretreatment process removed cesium as well as other radionuclides from the liquid wastes and captured these radioactive materials onto silica-based molecular sieves (zeolites). The decontaminated salt solutions were volume-reduced and then mixed with portland cement and other admixtures. Nineteen thousand eight hundred and seventy-seven 270-liter square drums were filled with the cement-wastes produced from the pretreatment process. These drums are being stored in a shielded facility on the site until their final disposition is determined. Over 6.4 million liters of liquid HLW were processed through the pretreatment system. PUREX supernatant was processed first, followed by two PUREX sludge wash solutions. A third wash of PUREX/THOREX sludge was then processed after the neutralized THOREX waste was mixed with the PUREX waste. Approximately 6.6 million curies of radioactive cesium-137 (Cs-137) in the HLW liquid were removed and retained on 65,300 kg of zeolites. With pretreatment complete, the zeolite material has been mobilized, size-reduced (ground), and blended with the PUREX and THOREX sludges in a single feed tank that will supply the HLW slurry to the Vitrification Facility

High-level radioactive wastes. Supplement 1

International Nuclear Information System (INIS)

McLaren, L.H.

1984-09-01

This bibliography contains information on high-level radioactive wastes included in the Department of Energy's Energy Data Base from August 1982 through December 1983. These citations are to research reports, journal articles, books, patents, theses, and conference papers from worldwide sources. Five indexes, each preceded by a brief description, are provided: Corporate Author, Personal Author, Subject, Contract Number, and Report Number. 1452 citations

Disposal of high-level radioactive wastes

Energy Technology Data Exchange (ETDEWEB)

Costello, J M [Australian Atomic Energy Commission Research Establishment, Lucas Heights

1982-03-01

The aims and options for the management and disposal of highly radioactive wastes contained in spent fuel from the generation of nuclear power are outlined. The status of developments in reprocessing, waste solidification and geologic burial in major countries is reviewed. Some generic assessments of the potential radiological impacts from geologic repositories are discussed, and a perspective is suggested on risks from radiation.

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.)

Security of highly radioactive sources in Nepal

International Nuclear Information System (INIS)

Shrestha, Kamal K.

2010-01-01

Subsequent to 9/11, concerned countries and UN agencies have taken especial interest in the security of highly radioactive sources throughout the world. The IAEA Nuclear Security Plan (2006-2009) consequently made as a result of UN Security Council Resolution 1540 is binding to all States. The Global Threat Reduction Initiative (GTRI) of the US and the Global Threat Reduction Programme (GTRP) of UK have assisted the four hospitals in Nepal having more than 1,000 Curies of radioactivity in their Cobalt-60 sources used for teletherapy. The physical upgrade of the security of the nuclear materials has also been launched in Nepal for prevention of theft with malicious intention or threats. In this presentation, the radioisotopes in Nepal that comes under different categories according to TECDOC-1355 of IAEA will be described. Problems and issues regarding the security and protection of radioactive sources at hospitals, academic and research institutions that could be prevalent in many developing counties too will be discussed by taking a case study of one of the cancer hospitals in Kathmandu valley. (author)

New facility for processing and storage of radioactive and toxic chemical waste

International Nuclear Information System (INIS)

Gallagher, F.E. III

1976-01-01

A new facility for the processing and storage of radioactive and toxic chemical waste is described. The facility is located in the science and engineering complex of the Santa Barbara campus of the University of California, near the Pacific Ocean. It is designed to provide a safe and secure processing and storage area for hazardous wastes, while meeting the high aesthetic standards and ecological requirements of campus and community regulatory boards. The ventilation system and fire prevention features will be described in detail. During the design phase, a small laboratory was added to provide an area for the radiation protection and industrial hygiene programs. Operational experience with this new facility is discussed

Deep geologic storage of high level radioactive wastes: conceptual generic designs

International Nuclear Information System (INIS)

1995-01-01

This report summarizes the studies on deep geologic storage of radioactive wastes and specially for the high-level radioactive wastes. The study is focussed to the geotechnical assessment and generic-conceptual designs. Methodology analysis, geotechnical feasibility, costs and operation are studied

NRI's research on radioactive wastes

International Nuclear Information System (INIS)

Alexa, J.; Dlouhy, Z.; Kepak, F.; Kourim, V.; Napravnik, J.; Razga, J.; Ralkova, J.; Uher, E.; Vojtech, O.

1976-01-01

A survey is given (including 41 references) of work carried out at the Nuclear Research Institute. Discussed are sorption processes (a selective sorbent for 90 Sr based on BaSO 4 , etc.), sorption on inorganic ion exchangers (heteropolyacid salts, ferrocyanides for 137 Cs capture), on organic cation exchangers (separation of lanthanides), electrocoagulation. The process is described of vitrification of highly radioactive wastes, the arrest of emissions, the deposition of radioactive wastes and surface decontamination. (M.K.)

Radioactive gas solidification apparatus

International Nuclear Information System (INIS)

Kobayashi, Yoshihiro; Seki, Eiji; Yabu, Tomohiko; Matsunaga, Hiroyuki.

1990-01-01

Handling of a solidification container from the completion for the solidifying processing to the storage of radioactive gases by a remote control equipment such as a manipulator requires a great cost and is difficult to realize. In a radioactive gas solidification device for injection and solidification in accumulated layers of sputtered metals by glow discharge, radiation shieldings are disposed surrounding the entire container, and cooling water is supplied to a cooling vessel formed between the container and the shielding materials. The shielding materials are divided into upper and lower shielding materials, so that solidification container can be taken out from the shielding materials. As a result, the solidification container after the solidification of radioactive gases can be handled with ease. Further, after-heat can be removed effectively from the ion injection electrode upon solidifying treatment upon storage, to attain a radioactive gas solidifying processing apparatus which is safe, economical and highly reliable. (N.H.)

Managing the nation's commercial high-level radioactive waste

International Nuclear Information System (INIS)

1985-03-01

This report presents the findings and conclusions of OTA's analysis of Federal policy for the management of commercial high-level radioactive waste. It represents a major update and expansion of the Analysis presented to Congress in our summary report, Managing Commercial High-Level Radioactive Waste, published in April of 1982 (NWPA). This new report is intended to contribute to the implementation of NWPA, and in particular to Congressional review of three major documents that DOE will submit to the 99th Congress: a Mission Plan for the waste management program; a monitored retrievable storage (MRS) proposal; and a report on mechanisms for financing and managing the waste program. The assessment was originally focused on the ocean disposal of nuclear waste. OTA later broadened the study to include all aspects of high-level waste disposal. The major findings of the original analysis were published in OTA's 1982 summary report

Glass as a medium for the ultimate disposal of highly radioactive waste

International Nuclear Information System (INIS)

Sombret, C.

1983-09-01

The conversion of high level radioactive liquid wastes into glass is now considered in every nuclear country. The glass composition must take into account the components of the solutions and be formulated in order to meet certain requirements, mainly those necessary for safe further disposal. The compositions of these glasses, all borosilicates, are consequently unusual. Heat due to β γ decay generates some devitrification but it has not yet been demonstrated that this is detrimental. β irradiation has minor effects on the glass structure but the effect of α emitters is not presently totally investigated. If stored energy consequenses are negligible, further experiments must be carried out to ascertain the effect of helium build up or the behaviour of the mechanical properties. Processes of industrial interest have been developped and a plant has already produced radioactive glass blocks for 5 years

Techniques of radioactive soil processing at rehabilitation of contamination territories - 59199

International Nuclear Information System (INIS)

Volkov, Victor; Chesnokov, Alexander; Danilovich, Alexey; Zverkov, Yury; Koltyshev, Sergey; Semenov, Sergey; Shisha, Anatoly

2012-01-01

Rehabilitation of nuclear- and radiation objects assumes dealing with and removal of considerable volumes of a radioactive soil. A similar situation was faced at the remediation of such sufficiently large objects, as old radioactive waste storages at the territory of 'Kurchatov Institute' and elimination of consequences of radiation accident at Podolsk plant of nonferrous metals. At rough estimates the volumes of a radioactive soil at territory of 'Kurchatov institute' were 15-20 thousand m 3 , volumes of a removed soil at carrying out of urgent measures in territory of Kirovo-Chepetsk chemical plant exceeded 20-25 thousand m 3 , volumes of a low active waste at the territory of Podolsk plant may reach 20 thousand m 3 . Such considerable volumes demand creation of technologies of their processing, an effective measuring technique of levels of their contamination and ways of considerable (in times) decrease of their volumes at the expense of decontamination or separation. Works have been aimed at the decision of these problems at carrying out of rehabilitation of territory 'Kurchatov institute'. During works technologies of radiation and water-gravitational separation of a radioactive soil have been offered and are realized in practice. A facility of water -gravitational separation of the soil was created and used within 5 years. It allowed decreasing of volumes of the low active waste in 5-6 times. In further the facility was supplied by a facility of radiation separation of the soil that has raised its efficiency. On turn there is a start-up question in experimental operation of facility of radiation separation of low active slag for Podolsk plant of nonferrous metals. The decision of these problems will allow to gain experience of creation of through technology of the processing of a radioactive soil and decrease in its volumes for using it as a design decisions for rehabilitation of other large scale radioactive-contaminated territories and industrial objects

Role of Congress in the High Level Radioactive Waste Odyssey: The Wisdom and Will of the Congress - 13096

International Nuclear Information System (INIS)

Vieth, Donald L.; Voegele, Michael D.

2013-01-01

Congress has had a dual role with regard to high level radioactive waste, being involved in both its creation and its disposal. A significant amount of time has passed between the creation of the nation's first high level radioactive waste and the present day. The pace of addressing its remediation has been highly irregular. Congress has had to consider the technical, regulatory, and political issues and all have had specific difficulties. It is a true odyssey framed by an imperative and accountability, by a sense of urgency, by an ability or inability to finish the job and by consequences. Congress had set a politically acceptable course by 1982. However, President Obama intervened in the process after he took office in January 2009. Through the efforts of his Administration, by the end of 2012, the US government has no program to dispose of high level radioactive waste and no reasonable prospect of a repository for high level radioactive waste. It is not obvious how the US government program will be reestablished or who will assume responsibility for leadership. The ultimate criteria for judging the consequences are 1) the outcome of the ongoing NRC's Nuclear Waste Confidence Rulemaking and 2) the concomitant permissibility of nuclear energy supplying electricity from operating reactors in the US. (authors)

Process for recovering xenon from radioactive gaseous wastes

International Nuclear Information System (INIS)

Kishimoto, Tsuneo.

1980-01-01

Purpose: To recover pure xenon economically and efficiently by amply removing radioactive krypton mixed in xenon without changing the rectifying capacity of an xenon rectifying system itself. Method: Xe containing radioactive Kr(Kr-85) is rectified to reduce the concentration of radioactive Kr. Thereafter, non-radioactive Kr or Ar is added to Xe and further the rectification is carried out. The raw material Xe from the Xe adsorption system of, for example, a radioactive gaseous waste disposal system is cooled to about 100 0 C by a heat-exchanger and thereafter supplied to a rectifying tower to carry out normal rectification of Xe thereby to reduce the concentration of Kr contained in Xe at the tower bottom to the rectification limit concentration. Then, non-radioactive Kr is supplied via a precooler to the tower bottom to continue the rectification, thus the Xe fractions at the tower bottom, in which the concentration of radioactive Kr is reduced, being compressed and recovered. (Kamimura, M.)

Radioactive airborne species formed in the air in high energy accelerator tunnels

International Nuclear Information System (INIS)

Kondo, K.

2005-01-01

Many radioactive airborne species have been observed in the air of high energy accelerator tunnels during machine operation. Radiation protection against these induced airborne radioactivities is one of the key issues for radiation safety, especially at high-energy and high-intense proton accelerators such as the J-PARC (Japan Proton Accelerator Research Complex, Joint project of KEK and JAERI), which is now under construction at the TOKAI site of JAERI. Information on the chemical forms and particle sizes of airborne radioactivities is essential for the estimation of internal doses. For that purpose, the study on radioactive airborne species formed in the air of beam-line tunnels at high-energy accelerators have been extensively conducted by our group. For Be-7, Na-24, S-38, Cl-38,-39, C-11, and N-13, formed by various types of nuclear reactions including nuclear spallation reactions, their aerosol and gaseous fractions are determined by a filter technique. A parallel plate diffusion battery is used for the measurement of aerosol size distributions, and the formation of radioactive aerosols is explained by the attachment of radionuclides to ambient non-radioactive aerosols which are formed through radiation induced reactions. The chemical forms of gaseous species are also determined by using a selective collection method based on a filter technique. A review is given of the physico-chemical properties of these airborne radionuclides produced in the air of accelerator beam-line tunnels.

Suggestions on R and D work of high-level radioactive waste disposal in China

International Nuclear Information System (INIS)

Xu Guoqing

2012-01-01

The difference between repository and generic underground facilities is described. Some differences and similarities of site selection between the low and medium radioactive waste disposal, nuclear power station and high-level radioactive waste repository are also discussed here. We trend to extremely emphasize the safety of high-level radioactive waste disposal because of high toxicity, long half-life and long safety disposal period of this kind of radioactive wastes; because radioactive waste in the repository is of high specific activities and buried in depth, it would be difficult to meddle with its safety. In case of repository system being destroyed, the author considers that in the stages of regional and area site selection, the first task is to investigate regional tectonic stability. Some problems about disposal options and others are also discussed in this paper. (author)

High-level radioactive wastes. Supplement 1

Energy Technology Data Exchange (ETDEWEB)

McLaren, L.H. (ed.)

1984-09-01

This bibliography contains information on high-level radioactive wastes included in the Department of Energy's Energy Data Base from August 1982 through December 1983. These citations are to research reports, journal articles, books, patents, theses, and conference papers from worldwide sources. Five indexes, each preceded by a brief description, are provided: Corporate Author, Personal Author, Subject, Contract Number, and Report Number. 1452 citations.

Hydrogen production during processing of radioactive sludge containing noble metals

International Nuclear Information System (INIS)

Ha, B.C.; Ferrara, D.M.; Bibler, N.E.

1992-01-01

Hydrogen was produced when radioactive sludge from Savannah River Site radioactive waste containing noble metals was reacted with formic acid. This will occur in a process tank in the Defense Waste Facility at SRS when waste is vitrified. Radioactive sludges from four tanks were tested in a lab-scale apparatus. Maximum hydrogen generation rates varied from 5 x10 -7 g H 2 /hr/g of sludge from the least reactive sludge (from Waste Tank 51) to 2 x10 -4 g H 2 /hr/g of sludge from the most reactive sludge (from Waste Tank 11). The time required for the hydrogen generation to reach a maximum varied from 4.1 to 25 hours. In addition to hydrogen, carbon dioxide and nitrous oxide were produced and the pH of the reaction slurry increased. In all cases, the carbon dioxide and nitrous oxide were generated before the hydrogen. The results are in agreement with large-scale studies using simulated sludges

Method of processing radioactive liquid wastes

International Nuclear Information System (INIS)

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

1984-01-01

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

Process and device for determining the spatial distribution of a radioactive substance

International Nuclear Information System (INIS)

1977-01-01

This invention describes a process for determining the spatial distribution of a radioactive substance consisting in determining the positions and energy losses associated to the interactions of the Compton effect and the photoelectric interactions that occur owing to the emission of gamma photons by the radioactive material and in deducing an information on the spatial distribution of the radioactive substance, depending on the positions and energy losses associated to the interactions of the Compton effect of these gamma photons and the positions and energy losses associated to the subsequent photoelectric interactions of these same photons. The invention also concerns a processing system for identifying, among the signals representing the positions and energy losses of the interactions of the Compton effect and the photoelectric interactions of the gamma photons emitted by a radioactive source, those signals that are in keeping with the gamma photons that have been subjected to an initial interaction of the Compton effect and a second and last photoelectric interaction. It further concerns a system for determining, among the identified signals, the positions of the sources of several gamma photons. This detector of Compton interaction can be used with conventional Auger-type imaging system (gamma camera) for detecting photoelectric interactions [fr

NPP radioactive waste processing and solidification

International Nuclear Information System (INIS)

Nikiforov, A.S.; Polyakov, A.S.; Zakharova, K.P.

1983-01-01

The problems of proce-sing NPP intermediate level- and low-level liquid radioactive wastes (LRW) are considered. Various methods are compared of LWR solidification on the base of bituminization, cement grouting and inclusion into synthetic resins. It is concluded that the considered methods ensure radioactive radionuclides effluents into open hydronetwork at the level below the sanitary, standards

Post-process intensification of photographic silver images, using radioactive compounds

International Nuclear Information System (INIS)

1979-01-01

A method of post-process intensification of silver images on a developed and fixed photographic film or plate is described, comprising the steps of (a) converting silver of the developed film or plate to a radioactive compound by contracting the film or plate with an aqueous alkaline solution of an organo-S 35 compound which reacts selectively with silver in a photographic film or plate; (b) placing the film or plate treated in step (a) in direct contact with a receiver film which is sensitive to beta radiation whereby the receiver film is exposed by radiation from the radioactive compound; and (c) developing and fixing the resulting intensified receiver film. (author)

Main approaches to solving the problems of radioactive waste processing and disposal

International Nuclear Information System (INIS)

Tarasov, V.M.; Syrkus, M.N.

1989-01-01

papers devoted to the problems of processing and disposal of radioactive wastes, formed during nuclear facility operation and after decommissioning are reviewed. Techniques for gaseous and liquid waste solidification, as well as solid waste processing by mechanical fragmentation and combustion are considered. Possibilities of radioactive waste disposal in cosmic space, their burial at the bed of seas ansd oceans, in geological storages are discussed. Special attention disposal. The conclusion is made that today there are no any uniform way for radiactive waste disposal and standard technical means for its realization. Solution of the problems considered should be of a complex character and it is carried out within international research programs

Human factors programs for high-level radioactive waste handling systems

International Nuclear Information System (INIS)

Pond, D.J.

1992-04-01

Human Factors is the discipline concerned with the acquisition of knowledge about human capabilities and limitations, and the application of such knowledge to the design of systems. This paper discusses the range of human factors issues relevant to high-level radioactive waste (HLRW) management systems and, based on examples from other organizations, presents mechanisms through which to assure application of such expertise in the safe, efficient, and effective management and disposal of high-level waste. Additionally, specific attention is directed toward consideration of who might be classified as a human factors specialist, why human factors expertise is critical to the success of the HLRW management system, and determining when human factors specialists should become involved in the design and development process

Human factors programs for high-level radioactive waste handling systems

International Nuclear Information System (INIS)

Pond, D.J.

1992-01-01

Human Factors is the discipline concerned with the acquisition of knowledge about human capabilities and limitations, and the application of such knowledge to the design of systems. This paper discusses the range of human factors issues relevant to high-level radioactive waste (HLRW) management systems and, based on examples form other organizations, presents mechanisms through which to assure application of such expertise in the safe, efficient, and effective management and disposal of high-level waste. Additionally, specific attention is directed toward consideration of who might be classified as a human factors specialist, why human factors expertise is critical to the success of the HLRW management system, and determining when human factors specialists should become involved in the design and development process

Radioactive wastes and their disposal

International Nuclear Information System (INIS)

Neumann, L.

1984-01-01

The classification of radioactive wastes is given and the achievements evaluated in the disposal of radioactive wastes from nuclear power plants. An experimental pilot unit was installed at the Jaslovske Bohunice nuclear power plant for the bituminization of liquid radioactive wastes. UJV has developed a mobile automated high-output unit for cementation. In 1985 the unit will be tested at the Jaslovske Bohunice and the Dukovany nuclear power plants. A prototype press for processing solid wastes was manufactured which is in operation at the Jaslovske Bohunice plant. A solidification process for atypical wastes from long-term storage of spent fuel elements has been developed to be used for the period of nuclear power plant decommissioning. (E.S.)

Spanish high level radioactive waste management system issues

International Nuclear Information System (INIS)

Ulibarri, A.; Veganzones, A.

1993-01-01

The Empresa Nacional de Residuous Radiactivos, S.A. (ENRESA) was set up in 1984 as a state-owned limited liability company to be responsible for the management of all kinds of radioactive wastes in Spain. This paper provides an overview of the strategy and main lines of action stated in the third General Radioactive Waste Plan, currently in force, for the management of spent nuclear fuel and high-level wastes, as well as an outline of the main related projects, either being developed or foreseen. Aspects concerning the organizational structure, the economic and financing system and the international co-operational are also included

Radioactive source management in Daya Bay NPP

International Nuclear Information System (INIS)

Mao Chun Yang

2000-01-01

'Small sources causes big accidents' had occurred worldwide many times. Radioactive source management in Nuclear Power Plant in very important for its safety record. This paper introduces the way and experience of radioactive source management in Daya Bay NPP from aspects of clarifying the responsibilities, centralizing the management of high radioactivity sources, work process management and experience feedback etc. (author)

Process for storing radioactive waste in ground

International Nuclear Information System (INIS)

Cohen, P.; Gouvenot, D.; Pagny, P.

1983-01-01

A process for storing radioactive waste in a cavity in the ground is claimed. The waste is conditioned and isolated from the ground by at least one retention barrier. A grout consisting of 1000 parts by weight of water, 40 to 400 parts by weight of cement, 80 to 1000 parts by weight of at least one clay chosen from the group including montmorillonite, illite and vermiculite, as well as 25 to 1200 parts by weight of kieselguhr and/or natural or artificial pozzuolanas is introduced into gaps in the soil areas surrounding the cavity

Radioactive Waste Management Research Program Plan for high-level waste: 1987

International Nuclear Information System (INIS)

1987-05-01

This plan will identify and resolve technical and scientific issues involved in the NRC's licensing and regulation of disposal systems intended to isolate high level hazardous radioactive wastes (HLW) from the human environment. The plan describes the program goals, discusses the research approach to be used, lays out peer review procedures, discusses the history and development of the high level radioactive waste problem and the research effort to date and describes study objectives and research programs in the areas of materials and engineering, hydrology and geochemistry, and compliance assessment and modeling. The plan also details the cooperative interactions with international waste management research programs. Proposed Earth Science Seismotectonic Research Program plan for radioactive waste facilities is appended

The disposal of radioactive waste

International Nuclear Information System (INIS)

Ormai, P.

2006-01-01

The first part shows different ways of 'producing' radioactive wastes, defines the wastes of small, medium and high activity and gives estimation on the quantity of the necessary capacities of waste disposal facilities. The modern radioactive waste disposal that is the integrated processing of the form of waste, the package, the technical facility and the embedding geological environment that guarantee the isolation together. Another factor is the lifetime of radioactive waste which means that any waste containing long lifetime waste in higher concentration than 400-4000 kBq/kg should be disposed geologically. Today the centre of debate disposal of radioactive waste is more social than technical. For this reason not only geological conditions and technical preparations, but social discussions and accepting communities are needed in selecting place of facilities. Now, the focus is on long term temporary disposal of high activity wastes, like burnt out heating elements. The final part of the paper summarizes the current Hungarian situation of disposal of radioactive wastes. (T-R.A.)

An appraisal of the hydrogeological processes involved in shallow subsurface radioactive waste management in Canadian terrain

International Nuclear Information System (INIS)

Grisak, G.E.; Jackson, R.E.

1978-01-01

The hydrogeological aspects of the problem of low-level radioactive waste management are introduced with a discussion of the Canadian nuclear power program; the nature of radioactive wastes and their rates of production; and the half-lives and health effects of ''waste'' radionuclides. As well, a general account is given of the present Canadian policy and procedures for licensing radioactive waste management sites. Following this introductory material, a detailed account is presented of the geohydrologic processes controlling the transport of radionuclides in groundwater flow systems and the attendant geochemical processes causing the retardation of the radionuclides. These geohydrologic and geochemical processes (i.e., hydrogeological processes) can be evaluated by the measurement of certain variables such as aquifer dispersivity, groundwater velocity, hydraulic conductivity, cation-exchange capacity, and total competing cations. To assess the possible importance of each variable in Canadian terrain, a comprehensive discussion of presently available (Canadian) data that have been compiled pertaining to each variable is presented. A description is then given of the hydrogeology of and the waste management experiences at radioactive waste management sites at Chalk River, Ontario; Bruce, Ontario; Whiteshell, Manitoba; and Suffield, Alberta. Along with this description there is a brief evaluation of those geohydrologic and geochemical processes that may be of importance at these sites. As a consequence of the above, site criteria outlining the nature of desirable hydrogeological environments for radioactive waste management areas are presented for those situations where the groundwater flow system acts as (a) a barrier to the migration of escaped radioactivity and (b) a joint dispersion-retardation system for liquid wastes. (author)

Non-radioactive stand-in for radioactive contamination. I. Non-radioactive tests

International Nuclear Information System (INIS)

Rohe, M.J.; Rankin, W.N.; Postles, R.L.

1985-01-01

Candidate non-radioactive materials for use as a stand-in for radioactive contamination during application of a high-pressure, hot water decontamination were identified and evaluated. A stand-in for radioactive contamination is needed to evaluate the decontaminability of replacement canyon cranes at the manufacturers location where actual radioactive contamination cannot be used. This evaluation was conducted using high-pressure, hot-water at 420 psi, 190 0 F, and 20 gal/min through a 1/8-in.-diam nozzle, the decontamination technique preferred by SRP Separations Department for this application. A non-radioactive stand-in for radioactive contamination was desired that would be removed by direct blast stream contact but would remain intact on surfaces where direct contact does not occur. This memorandum describes identification of candidate non-radioactive stand-in materials and evaluation of these materials in screening tests and tests with high-pressure, hot-water blasting. The following non-radioactive materials were tested: carpenter's line chalk; typing correction fluid; dye penetrant developer; latex paint with attapulyite added; unaltered latex paint; gold enamel; layout fluid; and black enamel. Results show that blue layout fluid and gold enamel have similar adherence that is within the range expected for actual radioactive contamination. White latex paint has less adherence than expected for actual radioactive contamination. The film was removed at a rate of 2 . Black enamel has more adherence than expected from actual radioactive contamination. In these tests ASTM No. 2B surfaces were harder to clean than either ASTM No. 1 or electropolished surfaces which had similar cleaning properties. A 90 0 blast angle was more effective than a 45 0 blast angle. In these tests there was no discernible effect of blast distance between 1 and 3 ft

Tomogram forming process and apparatus using radioactive isotopes

International Nuclear Information System (INIS)

Stoddart, H.F.

1980-01-01

This invention relates to nuclear medicine and particularly to a tomogram forming apparatus which permits, with great efficiency, the very sensitive quantitative determination and the accurate spatial localization of the radioactivity of a body section of a patient to whom a substance labelled with radioactive isotopes has been administered. This scanner is characterized in that it includes several highly focused collimators placed one after the other, according to an arrangement which surrounds a scanning field. Each collimator is mobile with respect to the adjacent one and a system enables the arrangement to be rotated about the scanning field from one scanning position to another. Another device enables the collimators to be moved so that, for each scanning position, the focus of each collimator uniformly samples at least half the scanning field [fr

Treatment of Medical Radioactive Liquid Waste Using Forward Osmosis (FO) Membrane Process

KAUST Repository

Lee, Songbok

2018-04-07

The use of forward osmosis (FO) for concentrating radioactive liquid waste from radiation therapy rooms in hospitals was systematically investigated in this study. The removal of natural and radioactive iodine using FO was first investigated with varying pHs and draw solutions (DSs) to identify the optimal conditions for FO concentration. Results showed that FO had a successful rejection rate for both natural and radioactive iodine (125I) of up to 99.3%. This high rejection rate was achieved at a high pH, mainly due to electric repulsion between iodine and membrane. Higher iodine removal by FO was also attained with a DS that exhibits a reverse salt flux (RSF) adequate to hinder iodine transport. Following this, actual radioactive medical liquid waste was collected and concentrated using FO under these optimal conditions. The radionuclides in the medical waste (131I) were removed effectively, but the water recovery rate was limited due to severe membrane fouling. To enhance the recovery rate, hydraulic washing was applied, but this had only limited success due to combined organic-inorganic fouling of the FO membrane. Finally, the effect of FO concentration on the reduction of septic tank volume was simulated as a function of recovery rate. To our knowledge, this study is the first attempt to explore the potential of FO technology for treating radioactive waste, and thus could be expanded to the dewatering of the radioactive liquid wastes from a variety of sources, such as nuclear power plants.

Operation and control of ion-exchange processes for treatment of radioactive wastes

International Nuclear Information System (INIS)

Emelity, L.A.

1967-01-01

A manual dealing with the application of ion-exchange materials to the treatment of radioactive wastes and reviewing the facilities currently using this method. This book is one of three commissioned by the IAEA on the principal methods of concentrating radioactive wastes. The content of this document is: (i) Historical review related to removal of radioactivity; (ii) Principles of ion exchange (iii) Ion-exchange materials; (iv) Limitations of ion exchangers; (v) Application of ion exchange to waste processing; (vi) Operational procedures and experiences; (vii) Cost-of-treatment by ion-exchange. The document also gives a list of producers of ion-exchange material and defines some relevant terms. 101 refs, 31 figs, 27 tabs

Operation and control of ion-exchange processes for treatment of radioactive wastes

Energy Technology Data Exchange (ETDEWEB)

Emelity, L A [Los Alamos National Lab., NM (United States)

1967-12-01

A manual dealing with the application of ion-exchange materials to the treatment of radioactive wastes and reviewing the facilities currently using this method. This book is one of three commissioned by the IAEA on the principal methods of concentrating radioactive wastes. The content of this document is: (i) Historical review related to removal of radioactivity; (ii) Principles of ion exchange (iii) Ion-exchange materials; (iv) Limitations of ion exchangers; (v) Application of ion exchange to waste processing; (vi) Operational procedures and experiences; (vii) Cost-of-treatment by ion-exchange. The document also gives a list of producers of ion-exchange material and defines some relevant terms. 101 refs, 31 figs, 27 tabs.

On risk assessment of high level radioactive waste disposal

International Nuclear Information System (INIS)

Smith, C.F.; Kastenberg, W.E.

1976-01-01

One of the major concerns with the continued growth of the nuclear power industry is the production of the high level radioactive wastes. The risks associated with the disposal of these wastes derives from the potential for release of radioactive materials into the environment. The development of a methodology for risk analysis is carried out. The methodology suggested involves the probabilistic analysis of a general accident consequence distribution. In this analysis, the frequency aspect of the distribution is treated separately from the normalized probability function. In the final stage of the analysis, the frequency and probability characteristics of the distribution are recombined to provide an estimate of the risk. The characterization of the radioactive source term is accomplished using the ORIGEN computer code. Calculations are carried out for various reactor types and fuel cycles, and the overall waste hazard for a projected 35 year nuclear power program is determined. An index of relative nuclide hazard appropriate to problems involving the management of high level radioactive wastes is developed. As an illustration of the methodology, risk analyses are made for two proposed methods for waste management: extraterrestrial disposal and interim surface storage. The results of these analyses indicate that, within the assumptions used, the risks of these management schemes are small compared with natural background radiation doses. (Auth.)

Radioactive liquid waste processing method

International Nuclear Information System (INIS)

Nishi, Takashi; Baba, Tsutomu; Fukazawa, Tetsuo; Matsuda, Masami; Chino, Koichi; Ikeda, Takashi.

1993-01-01

As an adsorbent used for removing radioactive nuclides such as cesium and strontium from radioactive liquid wastes generated from a reprocessing plant, a silicon compound having siloxane bonds constituted by silicon and oxygen and having silanol groups constituted by silicon, oxygen and hydrogen, or an inorganic material mainly comprising aluminosilicate constituted with silicon, oxygen and aluminum is used. In the adsorbent of the present invention, since silica main skeletons are partially decomposed in an aqueous alkaline solution to newly form silanol groups having a cation adsorbing property, pretreatment such as pH adjustment is not necessary. (T.M.)

Radioactive liquid waste processing system

International Nuclear Information System (INIS)

Inakuma, Masahiko; Takahara, Nobuaki; Hara, Satomi.

1996-01-01

Laundry liquid wastes and shower drains containing radioactive materials generated in a nuclear power plant are removed with radioactive materials by a fiber filtration device and an activated carbon filtration device to satisfy standers of water quality described in the environmental effect investigation report. Spent activated carbon is dehydrated together with the back-wash liquid from the fiber filtration device and the activated carbon filtration device using a Nutsche-type filtration dryer. With such procedures, the scale of the facility is minimized, space for devices, maintenance for equipments and radiation dose rate are reduced. (T.M.)

Highly water soluble nanoparticles as a draw solute in forward osmosis for the treatment of radioactive liquid waste

International Nuclear Information System (INIS)

Yang, Heeman; Choi, Hye Min; Jang, Sungchan; Seo, Bumkyoung; Lee, Kune Woo; Moon, Jei Kwon

2014-01-01

. In this study, we introduced highly water-soluble hyperbranched caroboxylated polyglycerol-coated magnetic nanoparticles (CPG-MNPs). It is known that the highly branched, globular architecture of PG significantly increase solubility compared to linear polymer and they are eco-friendly. The CPG-MNPs showed no aggregate of particles in water even after placing external magnet, and exhibited a high water flux in FO process. The CPG-MNPs are, therefore, potentially useful as a draw solute in FO processes. The operation of nuclear pressurized water reactors (PWRs) results in numerous radioactive waste streams which vary in radioactivity content. Most PWR stations have experienced leakages of boric acid into liquid radioactive waste systems. These wastes contain about 0.3∼0.8 wt% of boric acid. It is known that reverse osmosis (RO) membrane can eliminate boron at high pH and boron of 40∼90% can be removed by RO membrane in pH condition. RO uses hydraulic pressure to oppose, and exceed, the osmotic pressure of an aqueous feed solution containing boric acid. Forward osmosis (FO), a low energy technique based on membrane technologies, has recently garnered attention for its utility in wastewater treatment and desalination applications. In the FO process, water flows across a semi-permeable membrane from a solution with a low osmotic pressure (the feed solution) to a solution with a high osmotic pressure (the draw solution). The driving force in FO processes is provided by the osmotic gradient between the two solutions. Low energy costs and low degrees of membrane fouling are two of the advantages conveyed by FO processes over other processes, such as reverse osmosis processes that rely on a hydraulic pressure driving force. However, the challenges of FO still lie in the fabrication of eligible FO membranes and the readily separable draw solutes of high osmotic pressures. Superparamagnetic Fe3O4 nanoparticles can be separated from water by an external magnet field

Highly water soluble nanoparticles as a draw solute in forward osmosis for the treatment of radioactive liquid waste

Energy Technology Data Exchange (ETDEWEB)

Yang, Heeman; Choi, Hye Min; Jang, Sungchan; Seo, Bumkyoung; Lee, Kune Woo; Moon, Jei Kwon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-05-15

. In this study, we introduced highly water-soluble hyperbranched caroboxylated polyglycerol-coated magnetic nanoparticles (CPG-MNPs). It is known that the highly branched, globular architecture of PG significantly increase solubility compared to linear polymer and they are eco-friendly. The CPG-MNPs showed no aggregate of particles in water even after placing external magnet, and exhibited a high water flux in FO process. The CPG-MNPs are, therefore, potentially useful as a draw solute in FO processes. The operation of nuclear pressurized water reactors (PWRs) results in numerous radioactive waste streams which vary in radioactivity content. Most PWR stations have experienced leakages of boric acid into liquid radioactive waste systems. These wastes contain about 0.3∼0.8 wt% of boric acid. It is known that reverse osmosis (RO) membrane can eliminate boron at high pH and boron of 40∼90% can be removed by RO membrane in pH condition. RO uses hydraulic pressure to oppose, and exceed, the osmotic pressure of an aqueous feed solution containing boric acid. Forward osmosis (FO), a low energy technique based on membrane technologies, has recently garnered attention for its utility in wastewater treatment and desalination applications. In the FO process, water flows across a semi-permeable membrane from a solution with a low osmotic pressure (the feed solution) to a solution with a high osmotic pressure (the draw solution). The driving force in FO processes is provided by the osmotic gradient between the two solutions. Low energy costs and low degrees of membrane fouling are two of the advantages conveyed by FO processes over other processes, such as reverse osmosis processes that rely on a hydraulic pressure driving force. However, the challenges of FO still lie in the fabrication of eligible FO membranes and the readily separable draw solutes of high osmotic pressures. Superparamagnetic Fe3O4 nanoparticles can be separated from water by an external magnet field

An integrated approach to strategic planning in the civilian high-level radioactive waste management program

International Nuclear Information System (INIS)

Sprecher, W.M.; Katz, J.; Redmond, R.J.

1992-01-01

This paper describes the approach that the Office of Civilian Radioactive Waste Management (OCRWM) of the Department of Energy (DOE) is taking to the task of strategic planning for the civilian high-level radioactive waste management program. It highlights selected planning products and activities that have emerged over the past year. It demonstrates that this approach is an integrated one, both in the sense of being systematic on the program level but also as a component of DOE strategic planning efforts. Lastly, it indicates that OCRWM strategic planning takes place in a dynamic environment and consequently is a process that is still evolving in response to the demands placed upon it

The structures and stability of media intended for the immobilization of high level radioactive waste

International Nuclear Information System (INIS)

Tempest, P.A.

1979-05-01

High level radioactive waste contains about 40 different elements and, in time, many of these elements are transformed by radioactive decay into different-sized atoms with new chemical properties. The suitability of ordered crystal structures and unordered glass structures as media for immobilising the waste elements is compared. The structural properties of a mixture of synthetic minerals (SYNROC) are described and the various minerals' ability to accommodate ions of different radii and charge assessed. Similary the unordered structure of glass is examined and the probability of the glass remaining non-crystalline during manufacture and storage taken into account. Alternative glassification technologies in the form of the French AVM continuous process and the UK HARVEST batch processes are described and compared, and their likely effect on the structural properties of the final solid glass block considered. (author)

Process for producing zeolite adsorbent and process for treating radioactive liquid waste with the zeolite adsorbent

International Nuclear Information System (INIS)

Motojima, K.; Kawamura, F.

1984-01-01

Zeolite is contacted with an aqueous solution containing at least one of copper, nickel, cobalt, manganese and zinc salts, preferably copper and nickel salts, particularly preferably copper salt, in such a form as sulfate, nitrate, or chloride, thereby adsorbing the metal on the zeolite in its pores by ion exchange, then the zeolite is treated with a water-soluble ferrocyanide compound, for example, potassium ferrocyanide, thereby forming metal ferrocyanide on the zeolite in its pores. Then, the zeolite is subjected to ageing treatment, thereby producing a zeolite adsorbent impregnated with metal ferrocyanide in the pores of zeolite. The adsorbent can selectively recover cesium with a high percent cesium removal from a radioactive liquid waste containing at least radioactive cesium, for example, a radioactive liquid waste containing cesium and such coexisting ions as sodium, magnesium, calcium and carbonate ions at the same time at a high concentration. The zeolite adsorbent has a stable adsorbability for a prolonged time

Management of radioactive wastes

International Nuclear Information System (INIS)

Hendee, W.R.

1984-01-01

The disposal of radioactive wastes is perhaps the most controversial and least understood aspect of the use of nuclear materials in generating electrical power, the investigation of biochemical processes through tracer kinetics, and the diagnosis and treatment of disease. In the siting of nuclear power facilities, the disposal of radioactive wastes is invariably posed as the ultimate unanswerable question. In the fall of 1979, biochemical and physiologic research employing radioactive tracers was threatened with a slowdown resulting from temporary closure of sites for disposal of low-level radioactive wastes (LLW). Radioactive pharmaceuticals used extensively for diagnosis and treatment of human disease have increased dramatically in price, partly as a result of the escalating cost of disposing of radioactive wastes created during production of the labeled pharmaceuticals. These problems have resulted in identification of the disposal of LLW as the most pressing issue in the entire scheme of management of hazardous wastes. How this issue as well as the separate issue of disposal of high-level radioactive wastes (HLW) are being addressed at both national and state levels is the subject of this chapter

Electrochemical probing of high-level radioactive waste tanks containing washed sludge and precipitates

International Nuclear Information System (INIS)

Bickford, D.F.; Congdon, J.W.; Oblath, S.B.

1986-12-01

At the US Department of Energy's Savannah River Plant, corrosion of carbon steel storage tanks containing alkaline, high-level radioactive waste is controlled by specification of limits on waste composition and temperature. Processes for the preparation of waste for final disposal will result in waste with low corrosion inhibitor concentrations and, in some cases, high aromatic organic concentrations, neither of which are characteristic of previous operations. Laboratory tests, conducted to determine minimum corrosion inhibitor levels indicated pitting of carbon steel near the waterline for proposed storage conditions. In situ electrochemical measurements of full-scale radioactive process demonstrations have been conducted to assess the validity of laboratory tests. Probes included pH, Eh (potential relative to a standard hydrogen electrode), tank potential, and alloy coupons. In situ results are compared to those of the laboratory tests, with particular regard given to simulated solution composition. Transition metal hydroxide sludge contains strong passivating species for carbon steel. Washed precipitate contains organic species that lower solution pH and tend to reduce passivating films, requiring higher inhibitor concentrations than the 0.01 molar nitrite required for reactor fuel reprocessing wastes. Periodic agitation, to keep the organic phase suspended, or cathodic protection are possible alternatives to higher nitrite inhibitor concentrations

RECENT PROCESS AND EQUIPMENT IMPROVEMENTS TO INCREASE HIGH LEVEL WASTE THROUGHPUT AT THE DEFENSE WASTE PROCESSING FACILITY (DWPF)

International Nuclear Information System (INIS)

Smith, M; Allan Barnes, A; Jim Coleman, J; Robert Hopkins, R; Dan Iverson, D; Richard Odriscoll, R; David Peeler, D

2006-01-01

The Savannah River Site's (SRS) Defense Waste Processing Facility (DWPF), the world's largest operating high level waste (HLW) vitrification plant, began stabilizing about 35 million gallons of SRS liquid radioactive waste by-product in 1996. The DWPF has since filled over 2000 canisters with about 4000 pounds of radioactive glass in each canister. In the past few years there have been several process and equipment improvements at the DWPF to increase the rate at which the waste can be stabilized. These improvements have either directly increased waste processing rates or have desensitized the process and therefore minimized process upsets and thus downtime. These improvements, which include glass former optimization, increased waste loading of the glass, the melter glass pump, the melter heated bellows liner, and glass surge protection software, will be discussed in this paper

High-level radioactive waste in Canada. Background paper

International Nuclear Information System (INIS)

Fawcett, R.

1993-11-01

The disposal of radioactive waste is one of the most challenging environmental problems facing Canada today. Since the Second World War, when Canadian scientists first started to investigate nuclear reactions, there has been a steady accumulation of such waste. Research reactors built in the early postwar years produced small amounts of radioactive material but the volume grew steadily as the nuclear power reactors constructed during the 1960s and 1970s began to spawn used fuel bundles. Although this radioactive refuse has been safely stored for the short term, no permanent disposal system has yet been fully developed and implemented. Canada is not alone in this regard. A large number of countries use nuclear power reactors but none has yet put in place a method for the long-term disposal of the radioactive waste. Scientists and engineers throughout the world are investigating different possibilities; however, enormous difficulties remain. In Canada, used fuel bundles from nuclear reactors are defined as high-level waste; all other waste created at different stages in the nuclear fuel cycle is classified as low-level. Although disposal of low-level waste is an important issue, it is a more tractable problem than the disposal of high-level waste, on which this paper will concentrate. The paper discusses the nuclear fuel waste management program in Canada, where a long-term disposal plan has been under development by scientists and engineers over the past 15 years, but will not be completed for some time. Also discussed are responses to the program by parliamentary committees and aboriginal and environmental groups, and the work in the area being conducted in other countries. (author). 1 tab

High-level radioactive waste in Canada. Background paper

Energy Technology Data Exchange (ETDEWEB)

Fawcett, R [Library of Parliament, Ottawa, ON (Canada). Science and Technology Div.

1993-11-01

The disposal of radioactive waste is one of the most challenging environmental problems facing Canada today. Since the Second World War, when Canadian scientists first started to investigate nuclear reactions, there has been a steady accumulation of such waste. Research reactors built in the early postwar years produced small amounts of radioactive material but the volume grew steadily as the nuclear power reactors constructed during the 1960s and 1970s began to spawn used fuel bundles. Although this radioactive refuse has been safely stored for the short term, no permanent disposal system has yet been fully developed and implemented. Canada is not alone in this regard. A large number of countries use nuclear power reactors but none has yet put in place a method for the long-term disposal of the radioactive waste. Scientists and engineers throughout the world are investigating different possibilities; however, enormous difficulties remain. In Canada, used fuel bundles from nuclear reactors are defined as high-level waste; all other waste created at different stages in the nuclear fuel cycle is classified as low-level. Although disposal of low-level waste is an important issue, it is a more tractable problem than the disposal of high-level waste, on which this paper will concentrate. The paper discusses the nuclear fuel waste management program in Canada, where a long-term disposal plan has been under development by scientists and engineers over the past 15 years, but will not be completed for some time. Also discussed are responses to the program by parliamentary committees and aboriginal and environmental groups, and the work in the area being conducted in other countries. (author). 1 tab.

Management of spent high activity radioactive sources (SHARS)

International Nuclear Information System (INIS)

2002-09-01

The objective of this report is to provide all people involved in the handling and management of high activity sources with sufficient information about processes that are required for the safe management of spent high activity radioactive sources (SHARS). This includes examples of spent source management that are already taking place and also a description of the range of appropriate options that are available for each stage in the management process. This report also aims to identify the important issues to be addressed in order to develop a waste management strategy as part of the integrated management strategy that takes account of international experience and the guidance and principles that have been learned from that experience. This report relates specifically to SHARS, which are spent sources that have the potential, with short exposures, to produce acute health effects if handled incorrectly. In addition, they may also incur significant economic costs in any retrieval or environmental remediation operation, following loss of or damage to such a source. The report provides guidance on the technical, administrative and economic issues associated with SHARS from the moment they cease to be in use through to disposal, including temporary storage, transport, conditioning and interim storage

Dry decontamination technology development for high radioactive contaminant application - Development of residual radiation assessment methodology for high radioactive facility decontamination

Energy Technology Data Exchange (ETDEWEB)

Lee, G. J.; Hong, D. S.; Jeong, H. Y. [Korea Advanced Institute of Science and Technology, Taejon (Korea)

2000-04-01

Humidity around the concrete structure can make the moisture distribution within the concrete. Moisture content of the structure will change due to the diffusion process with time. Radioactive material on the surface of concrete will eventually diffuse into the porous concrete and contaminates the internal region of the concrete. In this study, we have assumed two different diffusion processes depending on the different moisture content. One is for slow diffusion process near the surface and the other is for fast diffusion process in deep region. We have developed the mathematical diffusion model for the two different diffusion regions and the corresponding analytic solutions for the two regions are obtained. 19 refs., 15 figs., 9 tabs. (Author)

Assessment of Radioactivity Inventory - a key parameter in the clearance for recycling process

International Nuclear Information System (INIS)

Larsson, Arne; Lundgren, Klas

2014-01-01

Decommissioning studies for nuclear power reactors are to be performed in order to assess the decommissioning costs and the waste volumes, as well as to provide data for the licensing and construction of the LILW repositories. An important part of this work is to estimate the amount of radioactivity in the different types of decommissioning waste. Studsvik has performed these assessments for all Swedish NPPs as well as other nuclear facilities in Sweden using thorough on-site sampling and robust calculations developed by Studsvik's team of senior experts. Precision has been found to be relatively high close to the reactor cores, but then declines as distance from the core increases. The decommissioning waste from a LWR can be separated into different categories such as: - Material affected by the neutron flux from the reactor core, - Process systems, - Waste handling systems, - Contaminated structures. The determined specific activities for different systems (or part of systems) are combined with data on weights and contaminated surface areas in order to assess the total activity. A key issue in the assessments has been efforts to reduce the uncertainties. Combining the unique knowledge in assessment of radioactivity inventories, the large data bank the waste processing represents and the knowledge and records from the laboratories, the activity determination codes can be validated and the waste processing analysis supported with additional data. (authors)

The Hybrid Treatment Process for treatment of mixed radioactive and hazardous wastes

International Nuclear Information System (INIS)

Ross, W.A.; Kindle, C.H.

1992-04-01

This paper describes a new process for treating mixed hazardous and radioactive waste, commonly called mixed waste. The process is called the Hybrid Treatment Process (HTP), so named because it is built on the 20 years of experience with vitrification of wastes in melters, and the 12 years of experience with treatment of wastes by the in situ vitrification (ISV) process

MethodS of radioactive waste processing and disposal in the United Kingdom

International Nuclear Information System (INIS)

Tolstykh, V.D.

1983-01-01

The results of investigations into radioactive waste processing and disposal in the United Kingdom are discussed. Methods for solidification of metal and graphite radioactive wastes and radioactive slime of the Magnox reactors are described. Specifications of different installations used for radioactive waste disposal are given. Climatic and geological conditions in the United Kingdom are such that any deep storages of wastes will be lower than the underground water level. That is why dissolution and transport by underground waters will inevitably result in radionuclide mobility. In this connection an extended program of investigations into the main three aspects of disposal problem namely radionucleide release in storages, underground water transport and radionuclide migration is realized. The program is divided in two parts. The first part deals with retrival of hydrological and geochemical data on geological formations, development of specialized methods of investigations which are necessary for identification of places for waste final disposal. The second part represents theoretical and laboratory investigations into provesses of radionuclide transport in the system of ''sttorage-geological formation''. It is concluded that vitrification on the base of borosilicate glass is the most advanced method of radioactive waste solidification

Decontamination processes for low level radioactive waste metal objects

International Nuclear Information System (INIS)

Longnecker, E.F.; Ichikawa, Sekigo; Kanamori, Osamu

1996-01-01

Disposal and safe storage of contaminated nuclear waste is a problem of international scope. Although the greatest volume of such waste is concentrated in the USA and former Soviet Union, Western Europe and Japan have contaminated nuclear waste requiring attention. Japan's radioactive nuclear waste is principally generated at nuclear power plants since it has no nuclear weapons production. However, their waste reduction, storage and disposal problems may be comparable to that of the USA on an inhabited area basis when consideration is given to population density where Japan's population, half that of the USA, lives in an area slightly smaller than that of California's. If everyone's backyard was in California, the USA might have insoluble radioactive waste reduction, storage and disposal problems. Viewing Japan's contaminated nuclear waste as a national problem requiring solutions, as well as an economic opportunity, Morikawa began research and development for decontaminating low level radioactive nuclear waste seven years ago. As engineers and manufacturers of special machinery for many years Morikawa brings special electro/mechanical/pneumatic Skills and knowledge to solving these unique problems. Genden Engineering Services and Construction Company (GESC), an affiliate of Japan Atomic Power Company, recently joined with Morikawa in this R ampersand D effort to decontaminate low level radioactive nuclear waste (LLW) and to substantially reduce the volume of such nuclear waste requiring long term storage. This paper will present equipment with both mechanical and chemical processes developed over these several years by Morikawa and most recently in cooperation with GESC

Role of Congress in the High Level Radioactive Waste Odyssey: The Wisdom and Will of the Congress - 13096

Energy Technology Data Exchange (ETDEWEB)

Vieth, Donald L. [DOE/NVOO Project Manager for Yucca Mountain, 1982 thru 1987, 1154 Cheltenham Place, Maineville, OH 45039 (United States); Voegele, Michael D. [Nye County Nuclear Waste Repository Project Office, 7404 Oak Grove Ave, Las Vegas, NV 89117 (United States)

2013-07-01

Congress has had a dual role with regard to high level radioactive waste, being involved in both its creation and its disposal. A significant amount of time has passed between the creation of the nation's first high level radioactive waste and the present day. The pace of addressing its remediation has been highly irregular. Congress has had to consider the technical, regulatory, and political issues and all have had specific difficulties. It is a true odyssey framed by an imperative and accountability, by a sense of urgency, by an ability or inability to finish the job and by consequences. Congress had set a politically acceptable course by 1982. However, President Obama intervened in the process after he took office in January 2009. Through the efforts of his Administration, by the end of 2012, the US government has no program to dispose of high level radioactive waste and no reasonable prospect of a repository for high level radioactive waste. It is not obvious how the US government program will be reestablished or who will assume responsibility for leadership. The ultimate criteria for judging the consequences are 1) the outcome of the ongoing NRC's Nuclear Waste Confidence Rulemaking and 2) the concomitant permissibility of nuclear energy supplying electricity from operating reactors in the US. (authors)

Human intrusion into geologic repositories for high-level radioactive waste: potential and prevention

Energy Technology Data Exchange (ETDEWEB)

Cameron, F X [Nuclear Regulatory Commission, Washington, DC (USA). Office of Nuclear Regulatory Research

1981-12-01

Isolation of high-level radioactive waste over long periods of time requires protection not only from natural events and processes, but also from the deliberate or inadvertent activities of future societies. This paper evaluates the likelihood of inadvertent human intrusion due to the loss of societal memory of the repository site. In addition measures to prevent inadvertent intrusion, and to guide future societies in any decision to deliberately intrude into the repository are suggested.

The Defense Waste Processing Facility: an innovative process for high-level waste immobilization

International Nuclear Information System (INIS)

Cowan, S.P.

1985-01-01

The Defense Waste Processing Facility (DWPF), under construction at the Department of Energy's Savannah River Plant (SRP), will process defense high-level radioactive waste so that it can be disposed of safely. The DWPF will immobilize the high activity fraction of the waste in borosilicate glass cast in stainless steel canisters which can be handled, stored, transported and disposed of in a geologic repository. The low-activity fraction of the waste, which represents about 90% of the high-level waste HLW volume, will be decontaminated and disposed of on the SRP site. After decontamination the canister will be welded shut by an upset resistance welding technique. In this process a slightly oversized plug is pressed into the canister opening. At the same time a large current is passed through the canister and plug. The higher resistance of the canister/plug interface causes the heat which welds the plug in place. This process provides a high quality, reliable weld by a process easily operated remotely

US and Russian innovative technologies to process low-level liquid radioactive wastes: The Murmansk initiative

International Nuclear Information System (INIS)

Dyer, R.S.; Duffey, R.B.; Penzin, R.; Sorlie, A.

1996-01-01

This paper documents the status of the technical design for the upgrade and expansion to the existing Low-level Liquid Radioactive Waste (LLLRW) treatment facility in Murmansk, the Russian Federation. This facility, owned by the Ministry of Transportation and operated by the Russian company RTP Atomflot in Murmansk, Russia, has been used by the Murmansk Shipping Company (MSCo) to process low-level liquid radioactive waste generated by the operation of its civilian icebreaker fleet. The purpose of the new design is to enable Russia to permanently cease the disposal at sea of LLLRW in the Arctic, and to treat liquid waste and high saline solutions from both the Civil and North Navy Fleet operations and decommissioning activities. Innovative treatments are to be used in the plant which are discussed in this paper

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

International Nuclear Information System (INIS)

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

1979-01-01

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

The Radioactive Waste Management course: 14 High-yield editions

International Nuclear Information System (INIS)

Alonso, A.; Gallego, E.; Marco, M. L.; Falcon, S.

2003-01-01

The doctorate course on Radioactive Waste Management was initiated in February 1988, by initiative of the Chair of Nuclear Technology, under the sponsoring of the national radioactive waste management company (ENRESA), in a fruitful collaboration between the Institute Artigas of the Technical School of Industrial Engineering and the Institute of Formation on Energy of the research centre CIEMAT. The course is also offered as a post-graduate through both institutes. After completion of fourteen consecutive editions in 2002, the course constituted a landmark in the field of nuclear education in Spain. The last edition offered, along 35 lessons published in two books, the general aspects of generation, treatment and conditioning of radioactive wastes, the basic Safety and Radiological Protection criteria, the detailed technical questions of the management of both low-intermediate.activity and the high-activity level, together with the wastes generated during decommissioning and dismantling of installations, as well as the general and institutional aspects. Experts in each field, belonging either to ENRESA, CIEMAT, the Nuclear Safety Council, the UPM and the industry, present such wide programme. A technical visit to the low and intermediate radioactive waste repository of El Cabril was also offered to the participants as part of the course, as in previous years the visit to the dismantling workers of Vandellos I NPP. More than 500 engineers and graduates in different science branches have participated in the course along 14 years, with both students and professionals belonging to ENRESA, the Nuclear Safety Council, CIEMAT and other research centers, hospitals, civil protection at different levels, service and engineering companies related with the radioactive waste management. Altogether, it is possible to say, as the title is expressed, that the course has given in these 14 years a high-production yield. (Author)

Radioactivity monitor for high-performance liquid chromatography

International Nuclear Information System (INIS)

Reeve, D.R.; Crozier, A.

1977-01-01

The coupling of a homogeneous radioactivity monitor to a liquid chromatograph involves compromises between the sensitivity of the monitor and the resolution and speed of analysis of the chromatograph. The theoretical relationships between these parameters are considered and expressions derived which make it possible to calculate suitable monitor operating conditions for most types of high-performance liquid chromatography

Method of processing radioactive liquid waste

International Nuclear Information System (INIS)

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

1988-01-01

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

Spent nuclear fuel and high level radioactive waste transportation. White paper

International Nuclear Information System (INIS)

1985-06-01

The High-Level Radioactive Waste Committee of the Western Interstate Energy Board has been involved in a year-long cooperative project with the US Department of Energy (DOE) to develop an information base on the transportation of spent nuclear fuel and high-level radioactive waste (HLW) so that western states can be constructive and informed participants in the repository program under the Nuclear Waste Policy Act (NWPA). The historical safety record of transportation of HLW and spent fuel is excellent; no release of these radioactive materials has ever occurred during transportation. Projected shipments under the NWPA will, however, greatly exceed current shipments in the US. For example, over the past five years, 119 metric tons of civilian spent fuel have been shipped in this country, while shipments to the first and second repository are each expected to peak at 3000 metric tons per year. The Committee believes that the successful development and operation of a national HLW/spent fuel transportation system can best be accomplished through an open process based on the common sense approach of taking all reasonable measures to minimize public risk and performing whatever actions are reasonably required to promote public acceptance. Therefore, the Committee recommends that the Department of Energy further the goals of the NWPA by developing a Comprehensive Transportation Plan which adopts a systematic, comprehensive, and integrated approach to resolving all spent fuel and HLW transportation issues in a timely manner. The suggested scope of such a plan is discussed in this White paper. Many of the suggested elements of such a plan are similar to those being developed by the Department of energy for inclusion in the Department's Transportation Institutional Plan

Characterization and vitrification of Hanford radioactive high level wastes

International Nuclear Information System (INIS)

Tingey, J.M.; Elliott, M.L.; Larson, D.E.; Morrey, E.V.

1991-01-01

Radioactive Neutralized Current Acid Waste (NCAW) samples from the Hanford waste tanks have been chemically, radiochemically and physically characterized. The wastes were processed according to the Hanford Waste vitrification Plant (HWVP) flowsheet, and characterized after each process step. The waste glasses were sectioned and leach tested. Chemical, radiochemical and physical properties of the waste will be presented and compared to nonradioactive simulant data and the HWVP reference composition and properties

Technical career opportunities in high-level radioactive waste management

International Nuclear Information System (INIS)

1993-01-01

Technical career opportunities in high-level radioactive waste management are briefly described in the areas of: Hydrology; geology; biological sciences; mathematics; engineering; heavy equipment operation; and skilled labor and crafts

Natural analogues to the conditions around a final repository for high-level radioactive waste

International Nuclear Information System (INIS)

Smellie, J.A.T.

1984-12-01

This report documents the proceedings resulting from a Workshop held at Lake Geneva, Wisconsin, USA, from 1-3 October, 1984. The theme of the Workshop was entitled 'Natural analogues to the conditions around a final repository for high-level radioactive waste', and was restricted to ultimate disposal in a crystalline bedrock environment. The Workshop provided an important first step in co-ordinating and focussing different national and individual interests and approaches towards natural analogue studies. One of the points highlighted at the concluding forum of the meeting was the necessity to first define the geochemical processes which are assumed to occur after disposal of the radioactive waste, and then locate suitable analogue systems which can be used to test the mechanisms of one, or a simple combination of these geochemical processes. Even accepting that the choice of which geochemical process(es) to be selected for validation will be sensitive to individual national disposal strategies, farfield radionuclide retardation mechanisms in the geosphere were considered to be a central topic of importance, and should therefore be given high priority. At this early stage in the development of natural analogue studies it was not possible to cover all the important aspects. In retrospect, the role of the models should have received more attention; bridging the gap between geoscientists and the modellers was seen as being of prime importance in future meetings of this nature. (author)

On barrier performance of high compaction bentonite in facilities of disposing high level radioactive wastes in formation

International Nuclear Information System (INIS)

Ikeda, Hidefumi; Komada, Hiroya

1989-01-01

As for the method of disposing high level radioactive wastes generated in the reprocessing of spent fuel, at present formation disposal is regarded as most promising. The most important point in this formation disposal is to prevent the leak of radioactive nuclides within the disposal facilities into bedrocks and their move to the zone of human life. As the method of formation disposal, the canisters containing high level radioactive wastes are placed in the horizontal or vertical holes for disposal dug from horizontal tunnels which are several hundreds m underground, and the tunnels and disposal holes are filled again. For this filling material, the barrier performance to prevent and retard the leak of radioactive nuclides out of the disposal facilities is expected, and the characteristics of low water permeability, the adsorption of nuclides and long term stability are required. However, due to the decay heat of wastes just after the disposal, high temperature and drying condition arises, and this must be taken in consideration. The characteristics required for filling materials and the selection of the materials, the features and classification of bentonite, the properties of high compaction bentonite, and the move of water, heat and nuclides in high compaction bentonite are reported.(Kako, I.)

The work of the High Commission for transparency and information on nuclear safety concerning the management of radioactive materials and waste

International Nuclear Information System (INIS)

Revol, H.

2011-01-01

Created by law in June 2006, the High Committee for Transparency and Information about Nuclear Security (HCTISN) is a pluralist instance dedicated to information, consultation and debates on issues related to nuclear activities and their impacts on environment and health. Currently chaired by a honorary senator, the High committee is composed of about forty persons who meet at least every three months. The subject of sustainable management of radioactive materials and waste is one of the main topical issues examined by the High committee. Since 2009, a dedicated think-tank has been set up to deal with questions of information, consultation and transparency all along the decision making process for the creation of a new storage installation for radioactive waste. In 2010, the High committee has been mandated by the Ministry in charge of nuclear safety to draw up a report on the transparency in the field of management of waste and materials produced at each stage of fuel cycle. Management of radioactive waste continues to raise discussion and rich debates within the High committee. In 2011, the High committee's intentions are to organize public information, consultations and debates about sustainable management of radioactive materials and waste. (author)

Commercial processing and disposal alternatives for very low levels of radioactive waste in the United States

International Nuclear Information System (INIS)

Benda, G.A.

2005-01-01

The United States has several options available in the commercial processing and disposal of very low levels of radioactive waste. These range from NRC licensed low level radioactive sites for Class A, B and C waste to conditional disposal or free release of very low concentrations of material. Throughout the development of disposal alternatives, the US promoted a graded disposal approach based on risk of the material hazards. The US still promotes this approach and is renewing the emphasis on risk based disposal for very low levels of radioactive waste. One state in the US, Tennessee, has had a long and successful history of disposal of very low levels of radioactive material. This paper describes that approach and the continuing commercial options for safe, long term processing and disposal. (author)

Particle size of radioactive aerosols generated during machine operation in high-energy proton accelerators

International Nuclear Information System (INIS)

Oki, Yuichi; Kanda, Yukio; Kondo, Kenjiro; Endo, Akira

2000-01-01

In high-energy accelerators, non-radioactive aerosols are abundantly generated due to high radiation doses during machine operation. Under such a condition, radioactive atoms, which are produced through various nuclear reactions in the air of accelerator tunnels, form radioactive aerosols. These aerosols might be inhaled by workers who enter the tunnel just after the beam stop. Their particle size is very important information for estimation of internal exposure doses. In this work, focusing on typical radionuclides such as 7 Be and 24 Na, their particle size distributions are studied. An aluminum chamber was placed in the EP2 beam line of the 12-GeV proton synchrotron at High Energy Accelerator Research Organization (KEK). Aerosol-free air was introduced to the chamber, and aerosols formed in the chamber were sampled during machine operation. A screen-type diffusion battery was employed in the aerosol-size analysis. Assuming that the aerosols have log-normal size distributions, their size distributions were obtained from the radioactivity concentrations at the entrance and exit of the diffusion battery. Radioactivity of the aerosols was measured with Ge detector system, and concentrations of non-radioactive aerosols were obtained using condensation particle counter (CPC). The aerosol size (radius) for 7 Be and 24 Na was found to be 0.01-0.04 μm, and was always larger than that for non-radioactive aerosols. The concentration of non-radioactive aerosols was found to be 10 6 - 10 7 particles/cm 3 . The size for radioactive aerosols was much smaller than ordinary atmospheric aerosols. Internal doses due to inhalation of the radioactive aerosols were estimated, based on the respiratory tract model of ICRP Pub. 66. (author)

Risk comparison of different treatment and disposal strategies of high level liquid radioactive waste

International Nuclear Information System (INIS)

Fang Dong

1997-01-01

The risk of different treatment and disposal strategies of high level liquid radioactive waste from spent fuel reprocessing is estimated and compared. The conclusions obtained are that risk difference from these strategies is very small and high level liquid waste can be reduced to middle and low level waste, if the decontamination factor for 99 Tc is large enough, which is the largest risk contributor in the high level radioactive waste from spent fuel reprocessing. It is also shown that the risk of high level radioactive waste could be reduced by the technical strategy of combining partitioning and transmutation

High purity radioactive beams at the bevalac

International Nuclear Information System (INIS)

Alonso, J.R.; Chatterjee, A.; Tobias, C.A.

1979-03-01

Peripheral nuclear fragmentation reactions of primary Bevalac heavy ion beams are used to produce secondary beams of radioactive nuclei. The large cross section and small deflection of the projectile fragments lead to high production and delivery efficiency for these beams. Dispersive beam transport allows good separation and purification of the desired secondary beams. 11 C and 19 Ne beams of high purity and good intensity (almost 0.2% of the primary beam current) are presently being used for biomedical experiments

High spin studies with radioactive ion beams

International Nuclear Information System (INIS)

Garrett, J.D.

1992-01-01

The variety of new research possibilities afforded by the culmination of the two frontier areas of nuclear structure: high spin and studies far from nuclear stability (utilizing intense radioactive ion beams) are discussed. Topics presented include: new regions of exotic nuclear shape (e.g. superdeformation, hyperdeformation, and reflection-asymmetric shapes); the population of and consequences of populating exotic nuclear configurations; and complete spectroscopy (i.e. the overlap of state of the art low-and high-spin studies in the same nucleus)

Materials performance in a high-level radioactive waste vitrification system

International Nuclear Information System (INIS)

Imrich, K.J.; Chandler, G.T.

1996-01-01

The Defense Waste Processing Facility (DWPF) is a Department of Energy Facility designed to vitrify highly radioactive waste. An extensive materials evaluation program has been completed on key components in the DWPF after twelve months of operation using nonradioactive simulated wastes. Results of the visual inspections of the feed preparation system indicate that the system components, which were fabricated from Hastelloy C-276, should achieve their design lives. Significant erosion was observed on agitator blades that process glass frit slurries; however, design modifications should mitigate the erosion. Visual inspections of the DWPF melter top head and off gas components, which were fabricated from Inconel 690, indicated that varying degrees of degradation occurred. Most of the components will perform satisfactorily for their two year design life. The components that suffered significant attack were the borescopes, primary film cooler brush, and feed tubes. Changes in the operation of the film cooler brush and design modifications to the feed tubes and borescopes is expected to extend their service lives to two years. A program to investigate new high temperature engineered materials and alloys with improved oxidation and high temperature corrosion resistance will be initiated

Vitrification of high-level radioactive and hazardous wastes

International Nuclear Information System (INIS)

Lutze, W.

1993-12-01

The main objective is to summarize work conducted on glasses as waste forms for high-level radioactive fission product solutions up to the late 1980's (section I and II). Section III addresses the question, whether waste forms designed for the immobilization of radioactive residues can be used for the same purpose for hazardous wastes. Of particular interest are those types of hazardous wastes, e.g., fly ashes from municipal combustion plants, easy to convert into glasses or ceramic materials. A large number of base glass compositions has been studied to vitrify waste from reprocessing but only borosilicate glasses with melting temperatures between 1100 C and 1200 C and very good hydrolytic stability is used today. (orig./HP) [de

Method of electrolytic processing for radioactive liquid waste

International Nuclear Information System (INIS)

Otsuka, Katsuyuki; Takahashi, Yoshiharu; Tamai, Hideaki.

1989-01-01

Radioactive liquid wastes containing sodium compounds are electrolized using mercury as a cathode. As a result, they are separated into sodium-containing metal amalgam and residues. Metals containing sodium are separated from amalgam, purified and re-utilized, while mercury is recycled to the electrolysis vessel. The foregoing method can provide advantageous effect such as: (1) volume of the wastes to be processed can be reduced, (2) since processing can be carried out at a relatively low temperature, low boiling elements can be handled with no evaporization, (3) useful elements can be recovered and (4) other method than glass solidification can easily be employed remarkable volume-reduction of solidification products can be expected. (K.M.)

Halftone display, particularly for a high resolution radioactivity distribution detection system

International Nuclear Information System (INIS)

Grenier, R.P.

1977-01-01

A device is described for presenting a halftone pictorial presentation composed of dot picture elements by selectively controlling the number of dot picture elements per unit area at locations on a display. In a high resolution radioactivity distribution detection system, the number of detected radioactive elements at XY locations of an array of sensing devices are fed to a computer and stored at corresponding address locations. The number of radioactive events detected at each address location is normalized into Gray scale coded signals as a function of the greatest number of radioactive events detected at any one address location. The normalized Gray scale coded signals are applied to a display for controlling the number of dot picture elements per unit area presented at corresponding XY locations on the display. The number of radioactive events detected at XY locations of the array are presented on the display as a halftone pictorial representation; the greatest number of picture dot elements per unit are being presented as a brighter image

Treatment and Storage of High-Level Radioactive Wastes. Proceedings of the Symposium on Treatment and Storage of High-Level Radioactive Wastes

International Nuclear Information System (INIS)

1963-01-01

A variety of radioactive materials having no immediate use result from the utilization of atomic energy. The manner in which these materials are handled has repercussions on reactor economy and technology, on the health and safety of persons and populations and on atomic legislation. Excellent progress has been made in developing a technology capable of safely and economically dealing with these materials so that no immediate problems exist. The highly radioactive ''wastes'' arising from the reprocessing of irradiated fuel pose long-range problems, however, and methods for the ultimate disposal of these wastes must be developed and evaluated. Such development and evaluation can be materially assisted by providing the scientists doing the work with an opportunity of exchanging ideas and information on their experience. Therefore, the IAEA, as part of its programme of promoting nuclear technology, convened in Vienna from 8-12 October 1962 the Symposium on the Treatment and Storage of High-level Radioactive Wastes. The Symposium was attended by 130 scientists from 19 countries and two international organizations. Thirty-three papers were presented and discussed in full and formed a background for a panel discussion of chairmen near the end of the Symposium. The papers and a record of the discussions are published in this single volume. It is hoped that the information thus recorded will achieve the desired purpose of assisting the peaceful development of atomic energy

Comparison of costs for solidification of high-level radioactive waste solutions: glass monoliths vs metal matrices

International Nuclear Information System (INIS)

Jardine, L.J.; Carlton, R.E.; Steindler, M.J.

1981-05-01

A comparative economic analysis was made of four solidification processes for liquid high-level radioactive waste. Two processes produced borosilicate glass monoliths and two others produced metal matrix composites of lead and borosilicate glass beads and lead and supercalcine pellets. Within the uncertainties of the cost (1979 dollars) estimates, the cost of the four processes was about the same, with the major cost component being the cost of the primary building structure. Equipment costs and operating and maintenance costs formed only a small portion of the building structure costs for all processes

Characterisation of radioactivity carrying aerosol in a mineral sand processing plant

International Nuclear Information System (INIS)

Jeffries, C.; Morawska, L.

1998-01-01

The techniques used to separate heavy mineral sand into mineral products produce a large amount of airborne particulate. Some of these particles are radioactive which is due to the thorium and, to a lesser extent, the uranium content of mineral sands. This study has investigated both the radioactive and respirable particle components (<10 μm) of the aerosol in a dry sand processing plant in Brisbane, Australia. A number of different measurement techniques have been used to characterise the aerosol in the plant. The mass, number and activity distributions have been determined by an eight stage cascade impactor and an Aerodynamic Particle Sizer (APS) with both instruments measuring 0.4 to 10 μm. Measurements of radon progeny concentrations and the extent of radon progeny attachment to micrometer sized particles has been investigated, as well as the extent of airborne thorium and uranium. The preliminary results from two sites are presented and comments are made on the relationship between total and radioactive aerosol

Processing of concentrated radioactive wastes into cement and bitumens following calcination

International Nuclear Information System (INIS)

Napravnik, J.; Sazavsky, P.; Ditl, P.; Prikryl, P.

1985-01-01

A brief characteristic is presented of the most frequently used processes of solidification of liquid radioactive wastes, viz., bituminization, cementation and their combination with calcination. The effect of individual parameters is assessed on the choice of the type of solidification process as is their importance in the actual process, in temporary storage, during transportation and under conditions of long-term storage. It has been found that a combination of the procedures could lead to a modular system of methods and equipment. This would allow to approach optimal solidification of wastes in the present period and to establish a research reserve for the development of more modern, economically advantageous and safer procedures. A rough estimate is made of the costs of the solidification of 1 m 3 of radioactive concentrate from the V-1 power plant at a production of 380 m 3 /year, this for the cementation-calcination and bituminization-calcination procedures. The said rough economic analysis only serves to identify the major operating components which have the greatest effect on the economic evaluation of the solidification procedures. (Z.M.)

Method of processing nitrate-containing radioactive liquid wastes

International Nuclear Information System (INIS)

Ogawa, Norito; Nagase, Kiyoharu; Otsuka, Katsuyuki; Ouchi, Jin.

1983-01-01

Purpose: To efficiently concentrate nitrate-containing low level radioactive liquid wastes by electrolytically dialyzing radioactive liquid wastes to decompose the nitrate salt by using an electrolytic cell comprising three chambers having ion exchange membranes and anodes made of special materials. Method: Nitrate-containing low level radioactive liquid wastes are supplied to and electrolytically dialyzed in a central chamber of an electrolytic cell comprising three chambers having cationic exchange membranes and anionic exchange membranes made of flouro-polymer as partition membranes, whereby the nitrate is decomposed to form nitric acid in the anode chamber and alkali hydroxide compound or ammonium hydroxide in the cathode chamber, as well as concentrate the radioactive substance in the central chamber. Coated metals of at least one type of platinum metal is used as the anode for the electrolytic cell. This enables efficient industrial concentration of nitrate-containing low level radioactive liquid wastes. (Yoshihara, H.)

Microbial processes in radioactive waste disposal

International Nuclear Information System (INIS)

Pedersen, Karsten

2000-04-01

Independent scientific work has unambiguously demonstrated life to be present in most deep geological formations investigated, down to depths of several kilometres. Microbial processes have consequently become an integral part of the performance safety assessment of high-level radioactive waste (HLW) repositories. This report presents the research record from the last decade of the microbiology research programme of the Swedish Nuclear Fuel and Waste Management Company (SKB) and gives current perspectives of microbial processes in HLW disposal. The goal of the microbiology programme is to understand how microbes may interact with the performance of a future HLW repository. First, for those who are not so familiar with microbes and their ways of living, the concept of 'microbe' is briefly defined. Then, the main characteristics of recognised microbial assemblage and microbial growth, activity and survival are given. The main part of the report summarises data collected during the research period of 1987-1999 and interpretations of these data. Short summaries introduce the research tasks, followed by reviews of the results and insight gained. Sulphate-reducing bacteria (SRB) produce sulphide and have commonly been observed in groundwater environments typical of Swedish HLW repositories. Consequently, the potential for sulphide corrosion of the copper canisters surrounding the HLW must be considered. The interface between the copper canister and the buffer is of special concern. Despite the fact that nowhere are the environmental constraints for life as strong as here, it has been suggested that SRB could survive and locally produce sulphide in concentrations large enough to cause damage to the canister. Experiments conducted thus far have indicated the opposite. Early studies in the research programme revealed previously unknown microbial ecosystems in igneous rock aquifers at depths exceeding 1000 m. This discovery triggered a thorough exploration of the

Microbial processes in radioactive waste disposal

Energy Technology Data Exchange (ETDEWEB)

Pedersen, Karsten [Goeteborg Univ. (Sweden). Dept. of Cell and Molecular Biology, Microbiology

2000-04-15

Independent scientific work has unambiguously demonstrated life to be present in most deep geological formations investigated, down to depths of several kilometres. Microbial processes have consequently become an integral part of the performance safety assessment of high-level radioactive waste (HLW) repositories. This report presents the research record from the last decade of the microbiology research programme of the Swedish Nuclear Fuel and Waste Management Company (SKB) and gives current perspectives of microbial processes in HLW disposal. The goal of the microbiology programme is to understand how microbes may interact with the performance of a future HLW repository. First, for those who are not so familiar with microbes and their ways of living, the concept of 'microbe' is briefly defined. Then, the main characteristics of recognised microbial assemblage and microbial growth, activity and survival are given. The main part of the report summarises data collected during the research period of 1987-1999 and interpretations of these data. Short summaries introduce the research tasks, followed by reviews of the results and insight gained. Sulphate-reducing bacteria (SRB) produce sulphide and have commonly been observed in groundwater environments typical of Swedish HLW repositories. Consequently, the potential for sulphide corrosion of the copper canisters surrounding the HLW must be considered. The interface between the copper canister and the buffer is of special concern. Despite the fact that nowhere are the environmental constraints for life as strong as here, it has been suggested that SRB could survive and locally produce sulphide in concentrations large enough to cause damage to the canister. Experiments conducted thus far have indicated the opposite. Early studies in the research programme revealed previously unknown microbial ecosystems in igneous rock aquifers at depths exceeding 1000 m. This discovery triggered a thorough exploration of the

Management of Radioactive Wastes

International Nuclear Information System (INIS)

Tchokosa, P.

2010-01-01

Management of Radioactive Wastes is to protect workers and the public from the radiological risk associated with radioactive waste for the present and future. It application of the principles to the management of waste generated in a radioisotope uses in the industry. Any material that contains or is contaminated with radionuclides at concentrations or radioactivity levels greater than ‘exempt quantities’ established by the competent regulatory authorities and for which no further use is foreseen or intended. Origin of the Radioactive Waste includes Uranium and Thorium mining and milling, nuclear fuel cycle operations, Operation of Nuclear power station, Decontamination and decommissioning of nuclear facilities and Institutional uses of isotopes. There are types of radioactive waste: Low-level Waste (LLW) and High-level Waste. The Management Options for Radioactive Waste Depends on Form, Activity, Concentration and half-lives of the radioactive waste, Storage and disposal methods will vary according to the following; the radionuclides present, and their concentration, and radio toxicity. The contamination results basically from: Contact between radioactive materials and any surface especially during handling. And it may occur in the solid, liquid or gas state. Decontamination is any process that will either reduce or completely remove the amount of radionuclides from a contaminated surface

Radioactive waste processing facility and underground processing method for radioactive wastes using the facility

International Nuclear Information System (INIS)

Hasegawa, Yasuyuki

1998-01-01

There are disposed a communication pit laterally extended in an underground base rock, an access pit extended from the ground surface to the communication pit, discarding pits laterally extended at a plurality of longitudinal positions of the communication pit and layered buffer materials for keeping a radioactive waste-sealing container at substantially the center of the discarding pit. The layered buffer material comprises fan-shaped buffer blocks divided so that the axial end faces of inner and outer layers are displaced with each other in the axial direction of the discarding pit and so that the circumferential end faces of the inner and the outer layers are circumferentially displaced with each other. Even if the base lock should move, the layered buffer material reduces the propagation of the movement to the radioactive waste-sealing vessel thereby enabling to enhance supporting strength. (N.H.)

Method of processing radioactive materials

International Nuclear Information System (INIS)

Kondo, Susumu; Moriya, Tetsuo; Ishibashi, Tadashi; Kariya, Masahiro.

1986-01-01

Purpose: To improve contamination proofness, water proofness, close bondability and stretching performance of strippable paints coated to substrates liable to be contaminated with radioactive materials. Method: Strippable paints are previously coated on substrates which may possibly be contaminated with radioactive materials. After the contamination, the coated membranes are stripped and removed. Alternatively, the strippable paints may be coated on the already contaminated substrates and, after drying, the paints are stripped and removed. The strippable paints used herein have a composition comprising a styrene-butadiene block copolymer containing from 60 to 80 wt% of styrene as a main ingredient and from 0.3 to 5 % by weight of a higher alkyl amine compound having 12 to 18 carbon atoms blended with the copolymer. (Ikeda, J.)

Solidification processing method for radioactive waste

International Nuclear Information System (INIS)

Hiraki, Akimitsu; Tanaka, Keiji; Heta, Katsutoshi.

1991-01-01

The pressure in a vessel containing radioactive wastes is previously reduced and cement mortar prepared by kneading cement, sand and kneading agent with water is poured under shaking substantially to the upper end of the vessel. After the lowering of the mortar level due to the deforming has been terminated, the pressure is increased gradually. Then, the cement mortar is further poured substantially to the upper end of the vessel again. With such a two step pouring method, spaces other than the radioactive wastes in the vessel can be filled substantially completely with the cement mortar. Accordingly, it is possible to avoid the problem in view of the strength due to the formation of gaps at the inside of the vessel, or leaching of radioactive materials due to the intrusion of water into the gaps. Further, if washing water is reutilized as water for kneading or washing after the precipitation of the solid contents, the amount of the secondary wastes generated can be reduced. (T.M.)

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

International Nuclear Information System (INIS)

Sakamoto, Motoi

1983-01-01

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

Identification of tetraphenylborate radiolysis products in a simulated feedstock for radioactive waste processing

International Nuclear Information System (INIS)

Eibling, R.E.; Bartlett, M.G.; Carlson, R.E.; Testino, S.A. Jr.; Kunkel, G.J.; Browner, R.F.; Busch, K.L.

1994-01-01

The first step towards immobilization of the soluble radioactive species in borosilicate glass is the addition of sodium tetraphenylborate (TPB) and sodium titanate to the radioactive aqueous solution. Initial studies of the TPB hydrolysis process have found that some component of the radiolysis mixture inactivates the Cu catalyst. The interaction of organic materials with the catalyst, and the subsequent interference with the hydrolysis process, would have presented problems with the use of the vitrification process. Prevention of the catalyst deactivation is obtained by washing the irradiated TPB precipitate in the Late Wash Facility prior to hydrolysis to remove the soluble radiolysis products. Identification of the organic radiolysis products, their distribution in the Late Wash Facility, and their interactions with the Cu catalyst has become an important analytical issue. To further investigate the reaction products of the TPB precipitation process, a simulated feedstock was created from compounds known to be present in the starting materials. This simulated feedstock was precipitated with sodium TPB and then exposed to Co-60 gamma radiation to simulate two years of additional storage time prior to the hydrolysis process. The irradiated product was divided into two parts, the filtered supernatant liquid and the precipitate slurry, which contains the TPB and the solid sodium titanate. Using gas chromatography/mass spectrometry, liquid secondary ion mass spectrometry, inductively coupled plasma/mass spectrometry, ion chromatography, and high performance liquid chromatography, over 50 organic and inorganic species have been identified in the aqueous portion of a simulated feedstock for TPB hydrolysis. The major organic species present are benzene, phenol, benzamide and a variety of substituted phenylphenols. The major inorganic species present are sodium, nitrite, and oxalate ions

High spin studies with radioactive ion beams

Energy Technology Data Exchange (ETDEWEB)

Garrett, J D [Oak Ridge National Lab., TN (United States)

1992-08-01

The variety of new research possibilities afforded by the culmination of the two frontier areas of nuclear structure: high spin and studies far from nuclear stability (utilizing intense radioactive ion beams) are discussed. Topics presented include: new regions of exotic nuclear shape (e.g. superdeformation, hyperdeformation, and reflection-asymmetric shapes); the population of and consequences of populating exotic nuclear configurations; and, complete spectroscopy (i.e. the overlap of state of the art low- and high-spin studies in the same nucleus). (author). 47 refs., 8 figs.

Safety of geologic disposal of high level radioactive waste

International Nuclear Information System (INIS)

Zaitsu, Tomohisa; Ishiguro, Katsuhiko; Masuda, Sumio

1992-01-01

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

Preliminary tests of an infrared process monitor for polyethylene encapsulation of radioactive waste

International Nuclear Information System (INIS)

Wright, S.L.; Jones, R.W.; McClelland, J.F.; Kalb, P.D.

1996-01-01

Polyethylene encapsulation is a process that is being investigated for the solidification of radioactive nitrate salts at Brookhaven National Laboratory and Rocky Flats Plant. In the encapsulation process, radioactive-salt waste is mixed with polyethylene pellets, heated, and extruded as a molten stream. Upon cooling, the mixture solidifies to a monolithic waste form with excellent properties for long-term waste storage. This paper describes a novel method to monitor the composition of the salt/polymer stream as it exits the extruder. The monitor is based on a technique known as transient infrared spectroscopy (TIRS). The TIRS monitor is able to capture the real-time mid-infrared spectrum of the processed waste stream as it exits the extruder. The wealth of chemical information contained in a mid-infrared spectrum makes this technique very appealing for on-line monitoring and process control. Data from the monitor can be used to guide processing, minimize waste volume, and certify the composition of the final waste form

Intergenerational ethics of high level radioactive waste

Energy Technology Data Exchange (ETDEWEB)

Takeda, Kunihiko [Nagoya Univ., Graduate School of Engineering, Nagoya, Aichi (Japan); Nasu, Akiko; Maruyama, Yoshihiro [Shibaura Inst. of Tech., Tokyo (Japan)

2003-03-01

The validity of intergenerational ethics on the geological disposal of high level radioactive waste originating from nuclear power plants was studied. The result of the study on geological disposal technology showed that the current method of disposal can be judged to be scientifically reliable for several hundred years and the radioactivity level will be less than one tenth of the tolerable amount after 1,000 years or more. This implies that the consideration of intergenerational ethics of geological disposal is meaningless. Ethics developed in western society states that the consent of people in the future is necessary if the disposal has influence on them. Moreover, the ethics depends on generally accepted ideas in western society and preconceptions based on racism and sexism. The irrationality becomes clearer by comparing the dangers of the exhaustion of natural resources and pollution from harmful substances in a recycling society. (author)

Intergenerational ethics of high level radioactive waste

International Nuclear Information System (INIS)

Takeda, Kunihiko; Nasu, Akiko; Maruyama, Yoshihiro

2003-01-01

The validity of intergenerational ethics on the geological disposal of high level radioactive waste originating from nuclear power plants was studied. The result of the study on geological disposal technology showed that the current method of disposal can be judged to be scientifically reliable for several hundred years and the radioactivity level will be less than one tenth of the tolerable amount after 1,000 years or more. This implies that the consideration of intergenerational ethics of geological disposal is meaningless. Ethics developed in western society states that the consent of people in the future is necessary if the disposal has influence on them. Moreover, the ethics depends on generally accepted ideas in western society and preconceptions based on racism and sexism. The irrationality becomes clearer by comparing the dangers of the exhaustion of natural resources and pollution from harmful substances in a recycling society. (author)

Neotectonic movement feature in preselection area for high level radioactive waste repository

International Nuclear Information System (INIS)

Huang Xianfang; Gao Yang; He Jianguo; Li Jianzhong; Gao Honglei; Xu Guoqing

2010-01-01

Neotectonic activity intensity is an important criteria for evaluating high level radioactive waste repository. The guiding ideology, methods and application of neotectonic study are elaborated in the paper. According to comparison research between the south and north part of east Tianshan area, the south part of east Tianshan is regarded as relative stable or relative weak in neotectonic movement in Neogene period and was selected as preselection area for high level radioactive waste repository. (authors)

Effects of cooking process on the changes of concentration and total amount of radioactive caesium in beef, wild plants and fruits

International Nuclear Information System (INIS)

Nabeshi, Hiromi; Tsutsumi, Tomoaki; Uekusa, Yoshinori; Matsuda, Rieko; Akiyama, Hiroshi; Teshima, Reiko; Hachisuka, Akiko

2016-01-01

In order to obtain information about effects of the cooking process on the changes of concentration and amount of radioactive materials in foods, we determined the concentration of radioactive caesium in several foods such as beef, edible wild plants, blueberries and mushrooms, before and after cooking. Our results showed that drying after soaking in liquid seasoning and the removal of astringent taste were effective in removing radioactive caesium from foods. More than 80% of radioactive caesium could be removed by these cooking methods. These results suggest that cooking processes such as boiling and soaking in liquid seasoning or water are effective to remove radioactive caesium from foods. Moreover, appropriate food additives such as baking soda were useful to promote the removal of radioactive caesium from foods. On the other hand, simple drying, jam making, grilling and tempura cooking could not remove radioactive caesium from foods. In addition, we showed that the concentration of radioactive caesium in foods was raised after simple drying, although the amount of radioactive caesium was unchanged. It would be necessary to monitor radioactive caesium concentration in processed foods because they might have undergone dehydration by cooking, which could result in concentrations exceeding regulatory levels. (author)

Site selection process for radioactive waste repository (radioactive facility) in Cuba as a fundamental safety criteria; Proceso de seleccion de emplazamiento como criterio fundamental de la seguridad para el repositorio de desechos radiactivos (instalacion radiactiva) en Cuba

Energy Technology Data Exchange (ETDEWEB)

Vital, Jose Luis Peralta; Castillo, Reinaldo Gil; Chales Suarez, Gustavo; Rodriguez Reyes, Aymee [Centro de Tecnologia Nuclear, La Habana (Cuba)

1999-11-01

The paper show the process of search carried out for the selection of the safest site in the National territory, in order to sitting the Facility (Repository) that will disposal the low and intermediate level radioactive wastes, as well as the possible Storage Facility for nuclear spent Fuel (radioactive wastes of high activity). We summarize the obtained Methodology and the Criterions of exclusion adopted for the development of the Process of site selection, as well as the current condition of the researches that will permit the obtaining of the nominative objectives. (author) 18 refs., 1 fig., 1 tab.

DCHAIN-SP 2001: High energy particle induced radioactivity calculation code

Energy Technology Data Exchange (ETDEWEB)

Kai, Tetsuya; Maekawa, Fujio; Kasugai, Yoshimi; Takada, Hiroshi; Ikeda, Yujiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kosako, Kazuaki [Sumitomo Atomic Energy Industries, Ltd., Tokyo (Japan)

2001-03-01

For the purpose of contribution to safety design calculations for induced radioactivities in the JAERI/KEK high-intensity proton accelerator project facilities, the DCHAIN-SP which calculates the high energy particle induced radioactivity has been updated to DCHAIN-SP 2001. The following three items were improved: (1) Fission yield data are included to apply the code to experimental facility design for nuclear transmutation of long-lived radioactive waste where fissionable materials are treated. (2) Activation cross section data below 20 MeV are revised. In particular, attentions are paid to cross section data of materials which have close relation to the facilities, i.e., mercury, lead and bismuth, and to tritium production cross sections which are important in terms of safety of the facilities. (3) User-interface for input/output data is sophisticated to perform calculations more efficiently than that in the previous version. Information needed for use of the code is attached in Appendices; the DCHAIN-SP 2001 manual, the procedures of installation and execution of DCHAIN-SP, and sample problems. (author)

Risk assessments for the disposal of high level radioactive wastes

International Nuclear Information System (INIS)

Smith, C.F.

1975-01-01

The risks associated with the disposal of high level wastes derive from the potential for release of radioactive materials into the environment. The assessment of these risks requires a methodology for risk analysis, an identification of the radioactive sources, and a method by which to express the relative hazard of the various radionuclides that comprise the high level waste. The development of a methodology for risk analysis is carried out after a review of previous work in the area of probabilistic risk assessment. The methodology suggested involves the probabilistic analysis of a general accident consequence distribution. In this analysis, the frequency aspect of the distribution is treated separately from the normalized probability function. At the final stage of the analysis, the frequency and probability characteristics of the distribution are recombined to provide an estimate of the risk. The characterization of the radioactive source term is accomplished using the ORIGEN computer code. Calculations are carried out for various reactor types and fuel cycles, and the overall waste hazard for a projected thirty-five year nuclear power program is determined

Transport of radioactive wastes

International Nuclear Information System (INIS)

Stuller, C.

2003-01-01

In this article author describes the system of transport and processing of radioactive wastes from nuclear power of Slovenske elektrarne, plc. It is realized the assurance of transport of liquid and solid radioactive wastes to processing links from places of their formation, or of preliminary storage and consistent transports of treated radioactive wastes fixed in cement matrix of fibre-concrete container into Rebublic storage of radioactive wastes in Mochovce

Radioactive wastes

International Nuclear Information System (INIS)

Teillac, J.

1988-01-01

This study of general interest is an evaluation of the safety of radioactive waste management and consequently the preservation of the environment for the protection of man against ionizing radiations. The following topics were developed: radiation effects on man; radioactive waste inventory; radioactive waste processing, disposal and storage; the present state and future prospects [fr

Sponsored research on radioactive waste management

International Nuclear Information System (INIS)

1983-01-01

The report is in chapters entitled: introduction (background, responsibilities, options, structure of the programme); strategy development; disposal of accumulations; disposal of radioactive waste arisings; quality assurance for waste conditioning quality assurance related to radioactive waste disposal (effectiveness of different rock types as natural barriers to the movement of radioactivity, and non-site specific factors in the design of repositories; radiological assessment; environmental studies; research and development to meet requirements specific to UKAEA wastes; long term research (processes for the solidification of highly active liquid wastes); plutonium contamination waste minimisation. (U.K.)

Method of processing cellulose filter sludge containing radioactive waste

International Nuclear Information System (INIS)

Shibata, Setsuo; Shibuya, Hidetoshi; Kusakabe, Takao; Kawakami, Hiroshi.

1991-01-01

To cellulose filter sludges deposited with radioactive wastes, 1 to 15% of cellulase based on the solid content of the filter sludges is caused to act in an aqueous medium with 4 to 8 pH at 10 to 50degC. If the pH value exceeds 8, hydrolyzing effect of cellulase is decreased, whereas a tank is corroded if the pH value is 4 or lower. If temperature is lower than 10degC, the rate of the hydrolysis reaction is too low to be practical. It is appropriate that the temperature is at the order of 40degC. If it exceeds 50degC, the cellulase itself becomes unstable. It is most effective that the amount of cellulase is about 8% and its addition by more than 15% is not effective. In this way, liquids in which most of filter sludges are hydrolyzed are processed as low level radioactive wastes. (T.M.)

The issue resolution process in the Civilian Radioactive Waste Transportation Program

International Nuclear Information System (INIS)

Holm, J.A.; Denny, S.

1987-01-01

The Civilian Radioactive Waste Management (OCRWM) Program consists of various technical and institutional program activities which engender concern from the general public and from policymakers at federal, state, and local levels. Most familiar are the concerns centered around selection of a site for a repository; however, the transportation portion of the OCRWM program also engenders similar concerns for safety, efficiency and effectiveness. The major Transportation institutional issues were detailed in the Institutional Plan, issued in 1986, and include topics such as liability, defense waste, routing, emergency response, risk identification and mitigation, cash integrity, inspection and enforcement of high-level waste shipments and use of overweight trucks as part of the modal mix. This paper will define the process being used to identify and resolve institutional issues, show how the technical and institutional issues interface and are addressed, and briefly describe four specific activities which illustrate the process of resolving institutional issues in the Transportation program

Process for solidifying high-level nuclear waste

Science.gov (United States)

Ross, Wayne A.

1978-01-01

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

Melt-processing method for radioactive solid wastes

International Nuclear Information System (INIS)

Kobayashi, Hiroaki

1998-01-01

Radioactive solid wastes are charged into a water-cooled type cold crucible induction melting furnace disposed in high frequency coils, and high frequency currents are supplied to high frequency coils which surround the melting furnace to melt the solid wastes by induction-heating. In this case, heat plasmas are jetted from above the solid wastes to the solid wastes to conduct initial heating to melt a portion of the solid wastes. Then, high frequency currents are supplied to the high frequency coils to conduct induction heating. According to this method, even when waste components of various kinds of materials are mixed, a portion of the solid wastes in the induction melting furnace can be melted by the initial heating by jetting heat plasmas irrespective of the kinds and the electroconductivity of the materials of the solid wastes. With such procedures, entire solid wastes in the furnace can be formed into a molten state uniformly and rapidly. (T.M.)

Managing commercial high-level radioactive waste: summary

International Nuclear Information System (INIS)

1982-04-01

This summary presents the findings and conclusions of OTA's analysis of Federal policy for the management of commercial high-level radioactive waste - an issue that has been debated over the last decade and that now appears to be moving toward major congressional action. After more than 20 years of commercial nuclear power, the Federal Government has yet to develop a broadly supported policy for fulfilling its legal responsibility for the final isolation of high-level radioactive waste. OTA's study concludes that until such a policy is adopted in law, there is a substantial risk that the false starts, shifts of policy, and fluctuating support that have plagued the final isolation program in the past will continue. The continued lack of final isolation facilities has raised two key problems that underlie debates about radioactive waste policy. First, some question the continued use of nuclear power until it is shown that safe final isolation for the resulting wastes can and will be accomplished, and argue that the failure to develop final isolation facilities is evidence that it may be an insoluble problem. Second, because there are no reprocessing facilities or federal waste isolation facilities to accept spent fuel, existing reactors are running out of spent fuel storage space, and by 1986 some may face a risk of shutting down for some period. Most of the 72,000 metric tons of spent fuel expected to be generated by the year 2000 will still be in temporary storage at that time. While it is possible that utilities could provide all necessary additional storage at reactor sites before existing basins are filled, some supplemental storage may be needed if there are delays in their efforts

A radioactive tracer dilution method to determine the mass of molten salt

International Nuclear Information System (INIS)

Lei Cao; Jarrell, Josh; Hardtmayer, D.E.; White, Susan; Herminghuysen, Kevin; Kauffman, Andrew; Sanders, Jeff; Li, Shelly

2017-01-01

A new technique for molten salt mass determination, termed radioactive tracer dilution, that uses 22 Na as a tracer was validated at bench scale. It has been a challenging problem to determine the mass of molten salt in irregularly shaped containers, where a highly radioactive, high-temperature molten salt was used to process nuclear spent/used fuel during electrochemical recycling (pyro-processing) or for coolant/fuel salt from molten salt reactors. A radioactive source with known activity is dissolved into the salt. After a complete mixture, a small amount of the salt is sampled and measured in terms of its mass and radioactivity. By finding the ratio of the mass to radioactivity, the unknown salt mass in the original container can be precisely determined. (author)

Processing method and device for radioactive liquid waste

International Nuclear Information System (INIS)

Matsuo, Toshiaki; Nishi, Takashi; Matsuda, Masami; Yukita, Atsushi.

1997-01-01

When only suspended particulate ingredients are contained as COD components in radioactive washing liquid wastes, the liquid wastes are heated by a first process, for example, an adsorption step to adsorb the suspended particulate ingredients to an activated carbon, and then separating and removing the suspended particulate ingredients by filtration. When both of the floating particle ingredients and soluble organic ingredients are contained, the suspended particulate ingredients are separated and removed by the first process, and then soluble organic ingredients are removed by other process, or both of the suspended particulate ingredients and the soluble organic ingredients are removed by the first process. In an existent method of adding an activated carbon and then filtering them at a normal temperature, the floating particle ingredients cover the layer of activated carbon formed on a filter paper or fabric to sometimes cause clogging. However, according to the method of the present invention, since disturbance by the floating particle ingredients does not occur, the COD components can be separated and removed sufficiently without lowering liquid waste processing speed. (T.M.)

Intrinsic vs. spurious long-range memory in high-frequency records of environmental radioactivity

International Nuclear Information System (INIS)

Donner, R.V.; Potirakis, S.M.; Barbosa, S.M.; Matos, J.A.O.; Pereira, A.J.S.C.; Neves, L.J.M.F.

2015-01-01

The presence or absence of long-range correlations in the environmental radioactivity fluctuations has recently attracted considerable interest. Among a multiplicity of practically relevant applications, identifying and disentangling the environmental factors controlling the variable concentrations of the radioactive noble gas radon is important for estimating its effect on human health and the efficiency of possible measures for reducing the corresponding exposition. In this work, we present a critical re-assessment of a multiplicity of complementary methods that have been previously applied for evaluating the presence of long-range correlations and fractal scaling in environmental radon variations with a particular focus on the specific properties of the underlying time series. As an illustrative case study, we subsequently re-analyze two high-frequency records of indoor radon concentrations from Coimbra, Portugal, each of which spans several weeks of continuous measurements at a high temporal resolution of five minutes. Our results reveal that at the study site, radon concentrations exhibit complex multi-scale dynamics with qualitatively different properties at different time-scales: (i) essentially white noise in the high-frequency part (up to time-scales of about one hour), (ii) spurious indications of a non-stationary, apparently long-range correlated process (at time scales between some hours and one day) arising from marked periodic components, and (iii) low-frequency variability indicating a true long-range dependent process. In the presence of such multi-scale variability, common estimators of long-range memory in time series are prone to fail if applied to the raw data without previous separation of time-scales with qualitatively different dynamics. (authors)

Initial demonstration of DWPF process and product control strategy using actual radioactive waste

International Nuclear Information System (INIS)

Andrews, M.K.; Bibler, N.E.; Jantzen, C.M.; Beam, D.C.

1991-01-01

The Defense Waste Processing Facility at the Savannah River Site (SRS) will vitrify high-level nuclear waste into borosilicate glass. The waste will be mixed with properly formulated glass-making frit and fed to a melter at 1150 degrees C. Process control and product quality are ensured by proper control of the melter feed composition. Algorithms have been developed to predict the processability of the melt and the durability of the final glass based on this feed composition. To test these algorithms, an actual radioactive waste contained in a shielded facility at SRS was analyzed and a frit composition formulated using a simple computer spreadsheet which contained the algorithms. This frit was then mixed with the waste and the resulting slurry fed to a research scale joule-heated melter operated remotely. Approximately 24 kg of glass were successfully prepared. This paper will describe the frit formulation, the vitrification process, and the glass durability

Studies on radioactivities of dust samples in the air at high altitudes

International Nuclear Information System (INIS)

Kohara, Eri; Muronoi, Naohiro

2015-01-01

The radioactivity concentrations of airborne dust samples were studied. The samples had been collected at high altitude by the Japan Air Self-Defense Force from April 2013 to March 2014. The obtained data were used for gross beta radioactivity analysis and gamma nuclide analysis. It is shown that cesium 137 was mainly detected at the 10 km and 3 km altitude of central area of Japan in several samples. Gaseous radioiodine was not detected in all the samples. Radioactive xenon was detected but the concentration did not show significant difference to the background level. (author)

Development of an immobilisation technique by cementation for non-radioactive simulated liquid waste, from Mo-99 production process

International Nuclear Information System (INIS)

Arva, E A; Marabini, S G; Varani, J L

2012-01-01

The Argentine Atomic Energy Commission (CNEA) is the responsible for developing a management nuclear waste disposal programme. This programme contemplates the strictly environmental safe and efficient management of the radioactive waste from different sources. Since 1985, CNEA has been producing commercially Mo-99 for medical use. In this process two types of liquid waste are produced. One of them has high alkaline (NaOH 3,5M) and aluminate contents. Since Mo-99 production started, such liquid waste was stored in specially designed containers during production, and after a decay period in smaller containers in interim storage conditions. As this waste is still a liquid, development of an immobilisation technique is required. Immobilisation of radioactive liquid waste by cementation is a frequently used technique, and will be studied in the present work using Mo-99 non-radioactive simulated liquid waste. In this second stage, a full scale (200 liters drum) cementation test using simulated non radioactive waste was carried out. Such test included: using the BEBA 201 mixing machine - the same that will be used with real waste in the future for 'tuning up' the process, construction of a specially designed temperature sensor for measuring the maximum temperature value (five different positions, four inside the drum and one outside) and the time elapsed after all components mixing. Finally, standard specimens (IRAM 1622) were made for mechanical resistance tests after cement setting at 28 days. The results show values of temperature not above 40 o C with the maximum at 12 hours before component mixing and compression strength of 14 MPa. Such values are compatible for a waste immobilisation process by cementation (author)

''Project Crystal'' for ultimate storage of highly radioactive waste

International Nuclear Information System (INIS)

Anon.

1983-01-01

NAGRA (The National Association for storage of radioactive waste) in Baden has launched in North Switzerland an extensive geological research program. The current research program, under the title of ''Project Crystal'', aims at providing the scientific knowledge which is required for the assessment of the suitability of the crystalline sub-soil of North Switzerland for the ultimate storage of highly radioactive waste. Safety and feasibility of such ultimate storage are in the forefront of preoccupations. Scientific institutes of France, Germany, USA and Canada are cooperating more particularly on boring research and laboratory analyses. Technical data are given on the USA and German installations used. (P.F.K.)

Ceramic process and plant design for high-level nuclear waste immobilization

International Nuclear Information System (INIS)

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

1983-01-01

In the last 3 years, significant advances in ceramic technology for high-level nuclear waste solidification have been made. Product quality in terms of leach-resistance, compositional uniformity, structural integrity, and thermal stability promises to be superior to borosilicate glass. This paper addresses the process effectiveness and preliminary designs for glass and ceramic immobilization plants. The reference two-step ceramic process utilizes fluid-bed calcination (FBC) and hot isostatic press (HIP) consolidation. Full-scale demonstration of these well-developed processing steps has been established at DOE and/or commercial facilities for processing radioactive materials. Based on Savannah River-type waste, our model predicts that the capital and operating cost for the solidification of high-level nuclear waste is about the same for the ceramic and glass options. However, when repository costs are included, the ceramic option potentially offers significantly better economics due to its high waste loading and volume reduction. Volume reduction impacts several figures of merit in addition to cost such as system logistics, storage, transportation, and risk. The study concludes that the ceramic product/process has many potential advantages, and rapid deployment of the technology could be realized due to full-scale demonstrations of FBC and HIP technology in radioactive environments. Based on our finding and those of others, the ceramic innovation not only offers a viable backup to the glass reference process but promises to be a viable future option for new high-level nuclear waste management opportunities

Method of processing radioactive rare gase

International Nuclear Information System (INIS)

Tagusagawa, Atsushi; Tuda, Kazuaki.

1988-01-01

Purpose: To obtain a safety processing method without using mechanical pumps or pressure-proof containers and, accordingly, with no risk for the leakage of radioactive rare gas. Method: A container filled with zeolige is inserted with a cover being opened into an autoclave. Meanwhile, krypton-containing gases are supplied to an adsorption tower filled with adsorbents, cooled, adsorbed and then heated to desorb adsorbed krypton. The krypton-containing gases are introduced due to the pressure difference to the autoclave thereby causing krypton to adsorb at ambient temperature to zeolite. Then, the inside of the autoclave is heated to desorb krypton and adsorbed moistures from zeolite and the pressure is elevated. After sending the gases under pressure to the adsorption tower, the zeolite-filled container is taken out from the autoclave, tightly closed and then transferred to a predetermined site. (Takahashi, M.)

Method for burning radioactive wastes

International Nuclear Information System (INIS)

Hattori, Akinori; Tejima, Takaya.

1987-01-01

Purpose: To completely process less combustible radioactive wastes with no excess loads on discharge gas processing systems and without causing corrosions to furnace walls. Method: Among combustible radioactive wastes, chlorine-containing less combustible wastes such as chlorine-containing rubbers and vinyl chlorides, and highly heat generating wastes not containing chloride such as polyethylene are selectively packed into packages. While on the other hand, packages of less combustible wastes are charged into a water-cooled jacket type incinerator intermittently while controlling the amount and the interval of charging so that the temperature in the furnace will be kept to lower than 850 deg C for burning treatment. Directly after the completion of the burning, the packed highly heat calorie producing wastes are charged and subjected to combustion treatment. (Yoshihara, H.)

The Optimization of Radioactive Waste Management in the Nuclear Installation Decommissioning Process

International Nuclear Information System (INIS)

Zachar, Matej; Necas, Vladimir

2008-01-01

The paper presents a basic characterization of nuclear installation decommissioning process especially in the term of radioactive materials management. A large amount of solid materials and secondary waste created after implementation of decommissioning activities have to be managed considering their physical, chemical, toxic and radiological characteristics. Radioactive materials should be, after fulfilling all the conditions defined by the authorities, released to the environment for the further use. Non-releasable materials are considered to be a radioactive waste. Their management includes various procedures starting with pre-treatment activities, continuing with storage, treatment and conditioning procedures. Finally, they are disposed in the near surface or deep geological repositories. Considering the advantages and disadvantages of all possible ways of releasing the material from nuclear installation area, optimization of the material management process should be done. Emphasis is placed on the radiological parameters of materials, availability of waste management technologies, waste repositories and on the radiological limits and conditions for materials release or waste disposal. Appropriate optimization of material flow should lead to the significant savings of money, disposal capacities or raw material resources. Using a suitable calculation code e.g. OMEGA, the evaluation of the various material management scenarios and selection of the best one, based on the multi-criterion analysis, should be done. (authors)

CIGeO geological disposal for high-level radioactive waste in France

International Nuclear Information System (INIS)

Ouzounian, Gerald; Bolia, Jelana

2014-01-01

Andra is the sole French organization responsible for the radioactive waste management in the country. Its work relies extensively on the legal basis provided by several major laws (Waste Act of 1991 and the Planning Act of 2006), which shaped the main principles of the waste management strategy and determined the corresponding implementation tools. Andra's industrial activities are essentially based around three of its national disposal facilities. Two of these operational facilities, by their design and comprehensive monitoring system, are considered worldwide as solid and proven reference solutions for the concerned types of radioactive waste. Andra is also charged with designing a future deep geological repository for intermediate-level long-lived and high-level waste and researching potential management and disposal solutions for the graphite and radium-bearing waste. The purpose of this article is to update the information to the readers about the Cigeo geological disposal project for high-level radioactive waste in France (authors)

Release process for non-real property containing residual radioactive material

International Nuclear Information System (INIS)

Ranek, N.L.; Chen, S.Y.; Kamboj, S.; Hensley, J.; Burns, D.; Fleming, R.; Warren, S.; Wallo, A.

1997-01-01

It is DOE's objective to operate its facilities and to conduct its activities so that radiation exposures to members of the public are maintained within acceptable limits and exposures to residual radioactive materials are controlled. To accomplish this, DOE has adopted Order DOE 5400.51 'Radiation Protection of the Public and the Environment', and will be promulgating IO CR Part 834 to codify and clarify the requirements of DOE 5400.5. Under both DOE 5400.5 and 10 CR Part 834, radioactively contaminated DOE property is prohibited from release unless specific actions have been completed prior to the release. This paper outlines a ten-step process that, if followed, will assist DOE Operations and contractor personnel in ensuring that the required actions established by Order DOE 5400.5 and 10 CR Part 834 have been appropriately completed prior to the release for reuse or recycle of non-real property (e.g., office furniture, computers, hand tools, machinery, vehicles and scrap metal). Following the process will assist in ensuring that radiological doses to the public from the released materials will meet applicable regulatory standards and be as low as reasonably achievable (ALARA)

The disposal of high level radioactive wastes. Proposed solutions and uses in Brazil

International Nuclear Information System (INIS)

Toledo, J.F.A.

1992-06-01

The characteristics of high level radioactive waste produced in nuclear plants similar to that used in Brazil is presented. Subsequently it is described the international experience, and the way to apply such knowledge to the Brazilian situation, defining the magnitude of the problem, applying a methodology to select sites, and choosing areas for the location of a repository. Once such areas are defined, it is presented the behaviour of rock mass, similar to those found in the brazilian territory, based on the requirements for a high radioactive waste repository site. Finally, two Projects are presented for countries with lithologies similar to that of Brazil. The first one is choosing sites for a high radioactive waste repository program, and the second is an investigation of rock mass responses program. (author)

Low-level radioactive waste processing at nuclear power plants

International Nuclear Information System (INIS)

1992-12-01

The Solid Radwaste Processing Source Book is presented as a supplement to the Liquid Radwaste Source Book released in 1990 and updated in 1991. The publication is the result of an industry-wide survey, and is intended as a resource for technical and managerial decisions involving of the processing of solid radioactive waste including ''wet'' and ''dry'' active waste as found at both Pressurized and Boiling Water Reactor sites. In addition to information on processes, vendors, volumes, and in-plant management activities, technology under consideration for future use and computer applications are listed. Together with key personnel and contact information contained in the Liquid Source Books, the collected data will be of great use when seeking specific, unbiased experience on which to base decisions related to so processing, disposal policy, or potential economic and regulatory impact

Reduction and resource recycling of high-level radioactive wastes through nuclear transmutation with PHITS code

International Nuclear Information System (INIS)

Fujita, Reiko

2017-01-01

In the ImPACT program of the Cabinet Office, programs are underway to reduce long-lived fission products (LLFP) contained in high-level radioactive waste through nuclear transmutation, or to recycle/utilize useful nuclear species. This paper outlines this program and describes recent achievements. This program consists of five projects: (1) separation/recovery technology, (2) acquisition of nuclear transmutation data, (3) nuclear reaction theory model and simulation, (4) novel nuclear reaction control and development of elemental technology, and (5) discussions on process concept. The project (1) develops a technology for dissolving vitrified solid, a technology for recovering LLFP from high-level waste liquid, and a technology for separating odd and even lasers. Project (2) acquires the new nuclear reaction data of Pd-107, Zr-93, Se-79, and Cs-135 using RIKEN's RIBF or JAEA's J-PARC. Project (3) improves new nuclear reaction theory and structural model using the nuclear reaction data measured in (2), improves/upgrades nuclear reaction simulation code PHITS, and proposes a promising nuclear transmutation pathway. Project (4) develops an accelerator that realizes the proposed transmutation route and its elemental technology. Project (5) performs the conceptual design of the process to realize (1) to (4), and constructs the scenario of reducing/utilizing high-level radioactive waste to realize this design. (A.O.)

Technical development for geological disposal of high-level radioactive wastes

International Nuclear Information System (INIS)

Asano, Hidekazu; Sugino, Hiroyuki; Kawakami, Susumu; Yamanaka, Yumiko

1997-01-01

Technical developments for geological disposal of high-level radioactive wastes materials research and design technique for engineered barriers (overpack and buffer material) were studied to evaluate more reliable disposal systems for high-level radioactive wastes. A lifetime prediction model for the maximum corrosion depth of carbon steel was developed. A preferable alloys evaluation method for crevice corrosion was established for titanium. Swelling pressure and water permeability of bentonite as a buffer material was measured, and coupled hydro-thermo-mechanical analysis code for bentonite was also studied. The CIP (cold isostatic pressing) method for monolithically formed buffer material was tested. A concept study on operation equipment for the disposal site was performed. Activities of microorganisms involved in underground performance were investigated. (author)

Process for ultimate storage of radioactive fission products

International Nuclear Information System (INIS)

Baukal, W.; Gruenthaler, K.H.; Neumann, K.

1980-01-01

In order to exclude cracking in the cooling phase during sealing of radioactive oxidic fission products in glass melts, metallic filling elements - e.g. wires, tissues - are proposed to be incorporated in the mould before the glass melt is poured in. Especially nickel alloys with corrosion proof surface layers, e.g. titanium nitride, silicon carbide, silicon nitride, aluminium oxide, suit best. These elements reduce thermal stresses and effect high thermal conductance towards the mould wall. (UWI) [de

OCRWM [Office of Civilian Radioactive Waste Management] publications catalog on high-level radioactive waste management

International Nuclear Information System (INIS)

1987-07-01

The US Department of Energy's (DOE's) Office of Civilian Radioactive Waste Management (OCRWM) is publishing this catalog to provide citations of selected technical and public information on the subject of high-level radioactive waste management. The catalog is a resource and reference tool. It will be updated and printed annually. The online catalog will be available for review through OCRWM's Product Record System (PRS) and can eventually be made available to the public. The printed catalog version is suitable for libraries and those individuals needing either a broad base of information or a particular source; the computerized catalog version provides the most current information resources available, since updates to citations will be made as they are received. The number of documents suitable for listing in this catalog is expected to grow significantly each year

Device for adsorbing exhaled radioactive gases and process

International Nuclear Information System (INIS)

Glasser, H.; Panetta, P.F.

1976-01-01

Sorption means are provided for sorbing radioactive gases, as in the exhalations of a living subject, especially for nuclear diagnostic test studies, comprising means for adsorbing the radioactive gas onto activated carbon, the carbon being contained in a plurality of independent, series-connected, chambers. The sorption means are especially adapted for the adsorption of radioactive inert gases such as xenon-133 ( 133 Xe). There can also be provided indicator means for indicating the flow-through of xenon comprising an indicator which changes color upon contact with xenon, such as dioxygenylhexafluoroantimoniate. 14 claims, 7 drawing figures

Radioactive waste repository of high ecological safety

International Nuclear Information System (INIS)

Sobolev, I.; Barinov, A.; Prozorov, L.

2000-01-01

With the purpose to construct a radioactive waste repository of high ecological safety and reliable containment, MosNPO 'Radon' specialists have developed an advanced type repository - large diameter well (LBD) one. A project is started for the development of a technology for LDW repository construction and pilot operation of the new repository for 25-30 years. The 2 LDW repositories constructed at the 'Radon' site and the developed monitoring system are described

Processing of radioactive ruthenium with aluminosilicate gels

International Nuclear Information System (INIS)

Kanno, Takuji; Ichinose, Yasuhiro; Ito, Katsuo

1979-01-01

Coprecipitation of radioactive Ru with hydroxides has been studied for the purpose of the management of the high level waste from the nuclear fuel reprocessing. Aluminosilicate gel used as coprecipitant was prepared by addition of aqueous sodium hydroxide to sodium aluminate-sodium silicate solution containing ruthenium nitrate. Ruthenium quantitatively precipitates under the conditions, aluminate > 4 x 10 -2 M, Al/Si 0 C. However, volatilization rate of Ru is suppressed by coating with mullite phase into which aluminosilicate gel transformes above 900 0 C. The amount of Ru volatilized in Ar-flow was reduced to about 10% of that in air-flow. (author)

Radioactive waste processing method and processing device therefor

International Nuclear Information System (INIS)

Matsuo, Toshiaki; Nishi, Takashi; Noge, Kenji; Matsuda, Masami; Takeshi, Kiyotaka

1998-01-01

Each predetermined amount of aggregates such as cements and sands as water-hardening solidification materials and kneading water are charged from a solidification material containing vessel, an aggregate containing vessel and a kneading water containing vessel to a kneading vessel of a paste supply device. The cements, the sands and the kneading water are kneaded by the rotation of a kneader. A produced solidification material paste is charged from the kneader to a drum through a paste transporting pump. Miscellaneous radioactive solid wastes have been filled in a drum. The solidification paste produced while supplying the cements, the sands and the kneading water into the kneader is discharged from the kneader. Since increase of viscosity of the solid material paste in the kneader is suppressed, the solidification paste can be easily flown into narrow gaps between radioactive miscellaneous solid wastes in the drum. (I.N.)

Radioactive waste solidification material

International Nuclear Information System (INIS)

Nishihara, Yukio; Wakuta, Kuniharu; Ishizaki, Kanjiro; Koyanagi, Naoaki; Sakamoto, Hiroyuki; Uchida, Ikuo.

1992-01-01

The present invention concerns a radioactive waste solidification material containing vermiculite cement used for a vacuum packing type waste processing device, which contains no residue of calcium hydroxide in cement solidification products. No residue of calcium hydroxide means, for example, that peak of Ca(OH) 2 is not recognized in an X ray diffraction device. With such procedures, since calcium sulfoaluminate clinker and Portland cement themselves exhibit water hardening property, and slugs exhibit hydration activity from the early stage, the cement exhibits quick-hardening property, has great extension of long term strength, further, has no shrinking property, less dry- shrinkage, excellent durability, less causing damages such as cracks and peeling as processing products of radioactive wastes, enabling to attain highly safe solidification product. (T.M.)

BN-350 decommissioning problems of radioactive waste management

International Nuclear Information System (INIS)

Galkin, A.; Tkachenko, V.

2002-01-01

Pursuant of modern concept on radioactive waste management applied in IAEA Member States all radioactive wastes produced during the BN-350 operation and decommissioning are subject to processing in order to be transformed to a form suitable for long-term storage and final disposal. The first two priority objectives for BN-350 reactor are as follows: cesium cleaning from sodium followed by sodium drain, and processing; processing of liquid and solid radioactive waste accumulated during BN-350 operation. Cesium cleaning from sodium and sodium processing to NaOH will be implemented under USA engineering and financial support. However the outputted product might be only subject to temporary storage under special conditions. Currently the problem is being solved on selection of technology for sodium hydroxide conversion to final product incorporated into cement-like matrix ready for disposal pursuant to existing regulatory requirements. Industrial installation is being designed for liquid radioactive waste processing followed by incorporation to cement matrix subject to further disposal. The next general objective is management of radioactive waste expected from BN-350 decommissioning procedure. Complex of engineering-radiation investigation that is being conducted at BN-350 site will provide estimation of solid and liquid radioactive waste that will be produced during the course of the BN-350 decommission. Radioactive wastes that will be produced may be shared for primary (metal structures of both reactor and reactor plant main and auxiliary systems equipment as well as construction wastes of dismantled biological protection, buildings and structures) and secondary (deactivation solutions, tools, materials, cloth, special accessory, etc.). Processing of produced radioactive wastes (including high activity waste) requires the use of special industrial facilities and construction of special buildings and structures for arrangement of facilities mentioned as well as for

Effects of natural radioactivity on food radioactivity measurement

International Nuclear Information System (INIS)

Ennyu, Atsuhito

2012-01-01

Since the accident of the Fukushima Daiichi Nuclear Power Station of Tokyo Electric Company, groups and individuals including local governments, food manufacturers, distribution circles, retail circles, and citizens are eager to measure the radioactivity of food, in order to confirm the safety of food from the concerns about radioactive contamination. The measurement of radioactivity of food is done by quantitatively determining gamma rays due to radioactive cesium that was incorporated into the biosphere cycle after having been released into the environment. As for the radioactivity measurement of food using gamma-ray spectrometry with a potassium iodide scintillation detector, which is very commonly used, this paper describes the handling method of obtained data, the principle of erroneous detection of radioactive cesium and iodine interrupted by natural radionuclides, and countermeasures for it. Major natural radioactivity sources are uranium series and thorium series. This paper explains gamma rays, which are characteristic in the decay process of uranium series and often affect the measurement of radioactive cesium in food and water. (O.A.)

Packaging of radioactive sludges at the Saclay effluent processing plant

International Nuclear Information System (INIS)

Cerre, Pierre; Mestre, Emile; Bourdrez, Jean; Leconnetable, Jean

1964-10-01

The authors describe technical and technological aspects of the packaging workshop for radioactive sludges produced by processes of co-precipitation of Saclay effluents. This facility is an achievement of studies which aimed at improving working conditions for the plant staff. This workshop implements a process of solidification of filtered sludge by mixing with a hydraulic binding agent. After some generalities on the decontamination process applied to effluents produced by the Saclay research centre, the authors present and describe the adopted process, propose a physical description of the facility: building, chemical engineering equipment (filtration, packaging, and handling). They describe facility operation: introduction of a block into the cell, block filling, output of a packaged container. They briefly discuss the first results of facility exploitation [fr

Radiant-heat spray-calcination process for the solid fixation of radioactive waste. Part 1, Non-radioactive pilot unit

Energy Technology Data Exchange (ETDEWEB)

Allemann, R.T.; Johnson, B.M. Jr.

1960-11-14

The fixation of radioactive waste in a stable solid media by means of calcination of these aqueous solutions has been the subject of considerable-effort throughout the U. S. Atomic Energy Commission and by atomic energy organizations in other countries. Several methods of doing this on a continuous or semi-continuous basis have been devised, and a fev have been demonstrated to be feasible for the handling of non-radioactive, or low-activity, simulated wastes. Notable among methods currently under development are: (a) batch-operated pot calcination of waste generated from reprocessing stainless steel clad fuel elements (Darex process) and Purex waste, (b) combination rotary kiln and ball mill calcination of aluminum nitrate (TBP-25 and Redox process), and (c) fluidized bed calcination of TBP-25 and Purex wastes. Although a considerable amount of engineering experience has been obtained on the calcination of dissolved salts in a fluidized bed, and the other methods have been the subjects of a great deal of study, none of them have been developed to-the extent which would rule out the desirability of further investigation of other possible methods of calcination.

Study on engineering economics of China high-level radioactive waste geological disposal

International Nuclear Information System (INIS)

Qu Jun; Guo Zongzhi; Yang Lirong; Hu Jiang

2012-01-01

In this paper, based on the research and analysis about the repository construction cost of the European, US and Japan, together with the concept design pattern of China's high level radioactive waste repository, the preliminary economic analysis of China is presented. Meanwhile, combining with China's nuclear power development layout and picking-up policy of spent fuel fund, the preliminary measurement concerning the capital resource of high level radioactive waste disposal is implemented, which contribute to the conclusion initiatively that the spent fuel fund could meet the need of the financial demand of disposal cost. (authors)

Development of a new process for radioactive decontamination of painted carbon steel structures by molten salt stripping

International Nuclear Information System (INIS)

Lainetti, Paulo Ernesto de O.

2009-01-01

The main practical difficulty associated to the task of the dismantling and decommissioning of the old Nuclear Fuel Cycle facilities of the IPEN has been the large amount of radioactive waste generated in the dismantling operations. The waste is mainly in the form of contaminated carbon steel structures. In the IPEN, the presence of contamination in the equipment, structures and buildings, although restricted to low and average activity levels, constituted an important concern due, on one hand, to the great volume of radioactive wastes generated during the operations. On the other hand, it should be outstanding that the capacity of stockpiling the radioactive wastes in IPEN found been exhausted. Basically, for the dismantling operations of the units, the main radionuclides of interest, from the radioprotection point of view, are U of natural isotopic composition and the thorium-232. Some attempts were done to reduce the volume of those wastes. Nevertheless, the only decontamination available methods were chemical methods such as pickling/rinsing treatments employing acid solutions (with nitric or citric acids) and alkaline solutions (sodium hydroxide). Different concentrations of such solutions were tested. The results obtained in the employed processes were not satisfactory. Ultrasonic equipment available was also employed in an attempt to increase the efficiency of decontamination. The choice of a coating removal process for radioactive material in the form of carbon steel pieces must have into account, among other factors, that it is not necessary a high quality of finishing, since the main objective is the release of the material as iron scrap. This paper describes the development of a new method for surface decontamination by immersion in molten salt baths. (author)

Methods for estimating costs of transporting spent fuel and defense high-level radioactive waste for the civilian radioactive waste management program

International Nuclear Information System (INIS)

Darrough, M.E.; Lilly, M.J.

1989-01-01

The US Department of Energy (DOE), through the Office of Civilian Radioactive Waste Management, is planning and developing a transportation program for the shipment of spent fuel and defense high-level waste from current storage locations to the site of the mined geologic repository. In addition to its responsibility for providing a safe transportation system, the DOE will assure that the transportation program will function with the other system components to create an integrated waste management system. In meeting these objectives, the DOE will use private industry to the maximum extent practicable and in a manner that is cost effective. This paper discusses various methodologies used for estimating costs for the national radioactive waste transportation system. Estimating these transportation costs is a complex effort, as the high-level radioactive waste transportation system, itself, will be complex. Spent fuel and high-level waste will be transported from more than 100 nuclear power plants and defense sites across the continental US, using multiple transport modes (truck, rail, and barge/rail) and varying sizes and types of casks. Advance notification to corridor states will be given and scheduling will need to be coordinated with utilities, carriers, state and local officials, and the DOE waste acceptance facilities. Additionally, the waste forms will vary in terms of reactor type, size, weight, age, radioactivity, and temperature

Investigation of radioactive contamination at non-radioactive drains of the Tsuruga Nuclear Power Station

International Nuclear Information System (INIS)

Koide, Hiroaki; Imanaka, Tetsuji; Ebisawa, Toru; Kawano, Shinji; Kobayashi, Keiji.

1982-05-01

In April, 1981, it was disclosed that a drainage area at the Tsuruga Nuclear Power Station was so much contaminated with radioactivites. Although Ministry of International Trade and Industry (MITI) officially provided an explanation of a process that resulted in the contamination, many problems remain unsolved on account of insufficient and limited investigations. The authors collected mud samples from contaminated manholes and examined radioactivities in them through the measurement of #betta#- and #betta#-spectra. Chemical separation of the samples was carried out in order to obtain precise concentration of radioactive cesium. Results are as follows: i) the concentration of radioactivities does not show monotonous decrease along the stream line but an anomalous peak at downstream manholes, ii) at the manhole specified No. 6 located rather downstream, 137 Cs concentration is significantly high and the composition of radioactive nuclides is quite different from that in the other manholes, and iii) additional radioactive contamination was observed in other manholes of non-radioactive drains which would not be influenced by the accident explained by MITI. Our present work has provided much more data than by MITI and made it clear that the overall data cnnot be consistent with the simple MITI explanation; a single radioactive release accident caused the disclosed contamination. It is concluded that non-radioactive water drains at the Tsuruga Nuclear Power Station had been under continual contamination. (author)

Evaluation of radionuclide concentrations in high-level radioactive wastes

International Nuclear Information System (INIS)

Fehringer, D.J.

1985-10-01

This report describes a possible approach for development of a numerical definition of the term ''high-level radioactive waste.'' Five wastes are identified which are recognized as being high-level wastes under current, non-numerical definitions. The constituents of these wastes are examined and the most hazardous component radionuclides are identified. This report suggests that other wastes with similar concentrations of these radionuclides could also be defined as high-level wastes. 15 refs., 9 figs., 4 tabs

Managing the nation's commercial high-level radioactive waste

International Nuclear Information System (INIS)

Cotton, T.

1985-01-01

With the passage of the Nuclear Waste Policy Act of 1982 (NWPA), Congress for the first time established in law a comprehensive Federal policy for commercial high-level radioactive waste management, including interim storage and permanent disposal. NWPA provides sufficient authority for developing and operating a high-level radioactive waste management system based on disposal in mined geologic repositories. Authorization for other types of waste facilities will not be required unless major problems with geologic disposal are discovered, and studies to date have identified no insurmountable technical obstacles to developing geologic repositories. The NWPA requires the Department of Energy (DOE) to submit to Congress three key documents: (1) a Mission Plan, containing both a waste management plan with a schedule for transferring waste to Federal facilities and an implementation program for choosing sites and developing technologies to carry out that plan; (2) a monitored retrievable storage (MRS) proposal, to include a site-specific design for a long-term federal storage facility, an evaluation of whether such an MRS facility is needed and feasible, and an analysis of how an MRS facility would be integrated with the repository program if authorized by Congress; and (3) a study of alternative institutional mechanisms for financing and managing the radioactive waste system, including the option of establishing an independent waste management organization outside of DOE. The Mission Plan and the report on alternative institutional mechanisms were submitted to the 99th US Congress in 1985. The MRS proposal is to be submitted in early 1986. Each of these documents is discussed following an overview of the Nuclear Waste Policy Act of 1982

Mental Models of Radioactivity and Attitudes towards Radioactive Waste

International Nuclear Information System (INIS)

Zeleznik, N.

2010-01-01

Siting of a radioactive waste repository presents a great problem in almost every country that produces such waste. The main problem is not a technical one, but socio-psychological, namely the acceptability of this kind of repository. Previous research on people's perception of the LILW repository construction, their attitudes towards radioactive waste, their willingness to accept it, indicated significant differences in answers of experts and lay persons, mainly regarding evaluation of the consequences of repository construction. Based on the findings of pilot investigations a mental model approach to the radioactivity, radioactive waste and repository was used as a method for development better risk communication strategies with local communities. The mental models were obtained by adjustment of the method developed by Morgan and co-workers where expert model of radioactivity is compared with mental model of lay people obtained through individual opened interviews. Additional information on trust, risk perception, role of main actors in the site selection process and their credibility was gained with the overall questionnaire on the representative sample of Slovenian population. Results of the survey confirm some already known findings, in addition we gained new cognitions and with analyses obtained the relationships and ratios between different factors, which are characteristics both for the general public and for the public, which is involved in the site selection process for a longer period and has been living beside a nuclear power plant for one generation. People have in general negative associations regarding the repository, the perceived risk for nuclear facilities is high, and trust in representatives of governmental institutions is low. Mental models of radioactivity, radioactive waste and the LILW repository are mostly irregular and differ from the experts' models. This is particularly valid for the models of radioactivity and the influences of

The Savannah River Site Replacement High Level Radioactive Waste Evaporator Project

International Nuclear Information System (INIS)

Presgrove, S.B.

1992-01-01

The Replacement High Level Waste Evaporator Project was conceived in 1985 to reduce the volume of the high level radioactive waste Process of the high level waste has been accomplished up to this time using Bent Tube type evaporators and therefore, that type evaporator was selected for this project. The Title I Design of the project was 70% completed in late 1990. The Department of Energy at that time hired an independent consulting firm to perform a complete review of the project. The DOE placed a STOP ORDER on purchasing the evaporator in January 1991. Essentially, no construction was to be done on this project until all findings and concerns dealing with the type and design of the evaporator are resolved. This report addresses two aspects of the DOE design review; (1) Comparing the Bent Tube Evaporator with the Forced Circulation Evaporator, (2) The design portion of the DOE Project Review - concentrated on the mechanical design properties of the evaporator. 1 ref

Studies on site characterization methodologies for high level radioactive waste disposal

International Nuclear Information System (INIS)

Wang Ju; Guo Yonghai; Chen Weiming

2008-01-01

This paper presents the final achievement of the project 'Studies of Site-specific Geological Environment for High Level Waste Disposal and Performance Assessment Methodology, Part Ⅰ: Studies on Site Characterization Methodologies for High Level Radioactive Waste Disposal', which is a 'Key Scientific and Technological Pre-Research Project for National Defense' during 2001-2005. The study area is Beishan area, Gansu Province, NW China--the most potential site for China's underground research laboratory and high level radioactive waste repository. The boreholes BS01, BS2, BS03 and BS04 drilled in fractured granite media in Beishan are used to conduct comprehensive studies on site characterization methodologies, including: bore hole drilling method, in situ measurement methods of hydrogeological parameters, underground water sampling technology, hydrogeochemical logging method, geo-stress measurement method, acoustic borehole televiewer measurement method, borehole radar measurement method, fault stability evaluation methods and rock joint evaluation method. The execution of the project has resulted in the establishment of an 'Integrated Methodological System for Site Characterization in Granite Site for High Level Radioactive Waste Repository' and the 8 key methodologies for site characterization: bore hole drilling method with minimum disturbance to rock mass, measurement method for hydrogeological parameters of fracture granite mass, in situ groundwater sampling methods from bore holes in fractured granite mass, fracture measurement methods by borehole televiewer and bore radar system, hydrogeochemical logging, low permeability measurement methods, geophysical methods for rock mass evaluation, modeling methods for rock joints. Those methods are comprehensive, advanced, innovative, practical, reliable and of high accuracy. The comprehensive utilization of those methods in granite mass will help to obtain systematic parameters of

Natural analogues for processes affecting disposal of high-level radioactive waste in the vadose zone

Science.gov (United States)

Stuckless, J. S.

2003-04-01

Natural analogues can contribute to understanding and predicting the performance of subsystems and processes affecting a mined geologic repository for high-level radioactive waste in several ways. Most importantly, analogues provide tests for various aspects of systems of a repository at dimensional scales and time spans that cannot be attained by experimental study. In addition, they provide a means for the general public to judge the predicted performance of a potential high-level nuclear waste repository in familiar terms such that the average person can assess the anticipated long-term performance and other scientific conclusions. Hydrologists working on the Yucca Mountain Project (currently the U.S. Department of Energy's Office of Repository Development) have modeled the flow of water through the vadose zone at Yucca Mountain, Nevada and particularly the interaction of vadose-zone water with mined openings. Analogues from both natural and anthropogenic examples confirm the prediction that most of the water moving through the vadose zone will move through the host rock and around tunnels. This can be seen both quantitatively where direct comparison between seepage and net infiltration has been made and qualitatively by the excellent degree of preservation of archaeologic artifacts in underground openings. The latter include Paleolithic cave paintings in southwestern Europe, murals and artifacts in Egyptian tombs, painted subterranean Buddhist temples in India and China, and painted underground churches in Cappadocia, Turkey. Natural analogues also suggest that this diversion mechanism is more effective in porous media than in fractured media. Observations from natural analogues are also consistent with the modeled decrease in the percentage of infiltration that becomes seepage with a decrease in amount of infiltration. Finally, analogues, such as tombs that have ben partially filled by mud flows, suggest that the same capillary forces that keep water in the

Midwestern High-Level Radioactive Waste Transportation Project

International Nuclear Information System (INIS)

1993-01-01

On February 17,1989, the Midwestern Office of The Council of State Governments and the US Department of Energy entered into a cooperative agreement authorizing the initiation of the Midwestern High-Level Radioactive Waste Transportation Project. The transportation project continued to receive funding from DOE through amendments to the original cooperative agreement, with December 31, 1993, marking the end of the initial 5-year period. This progress report reflects the work completed by the Midwestern Office from February 17,1989, through December 31,1993. In accordance with the scopes of work governing the period covered by this report, the Midwestern Office of The Council of State Governments has worked closely with the Midwestern High-Level Radioactive Waste Committee. Project staff have facilitated all eight of the committee's meetings and have represented the committee at meetings of DOE's Transportation Coordination Group (TCG) and Transportation External Coordination Working Group (TEC/WG). Staff have also prepared and submitted comments on DOE activities on behalf of the committee. In addition to working with the committee, project staff have prepared and distributed 20 reports, including some revised reports (see Attachment 1). Staff have also developed a library of reference materials for the benefit of committee members, state officials, and other interested parties. To publicize the library, and to make it more accessible to potential users, project staff have prepared and distributed regular notices of resource availability

Airborne anthropogenic radioactivity measurements from an international radionuclide monitoring system

International Nuclear Information System (INIS)

Mason, L.R.; Bohner, J.D.; Williams, D.L.

1998-01-01

Anthropogenic radioactivity is being measured in near-real time by an international monitoring system designed to verify the Comprehensive Nuclear Test Ban Treaty. Airborne radioactivity measurements are conducted in-situ by stations that are linked to a central data processing and analysis facility. Aerosols are separated by high-volume air sampling with high-efficiency particulate filters. Radio-xenon is separated from other gases through cryogenic methods. Gamma-spectrometry is performed by high purity germanium detectors and the raw spectral data is immediately transmitted to the central facility via Internet, satellite, or modem. These highly sensitive sensors, combined with the automated data processing at the central facility, result in a system capable of measuring environmental radioactivity on the microbecquerel scale where the data is available to scientists within minutes of the field measurement. During the past year, anthropogenic radioactivity has been measured at approximately half of the stations in the current network. Sources of these measured radionuclides include nuclear power plant emissions, Chernobyl resuspension, and isotope production facilities. The ability to thoroughly characterize site-specific radionuclides, which contribute to the radioactivity of the ambient environment, will be necessary to reduce the number of false positive events. This is especially true of anthropogenic radionuclides that could lead to ambiguous analysis. (author)

Radioactive tracer technique in process optimization: applications in the chemical industry

International Nuclear Information System (INIS)

Charlton, J.S.

1989-01-01

Process optimization is concerned with the selection of the most appropriate technological design of the process and with controlling its operation to obtain maximum benefit. The role of radioactive tracers in process optimization is discussed and the various circumstances under which such techniques may be beneficially applied are identified. Case studies are presented which illustrate how radioisotopes may be used to monitor plant performance under dynamic conditions to improve production efficiency and to investigate the cause of production limitations. In addition, the use of sealed sources to provide information complementary to the tracer study is described. (author)

Design and heat transfer calculations of burial-bunker for one-stage melting converter for vitrification of high-level radioactive waste

International Nuclear Information System (INIS)

Pioro, L.S.; P'Yanykh, K.E.; Pioro, I.L.

2001-01-01

Widespread application of radioactive materials in different branches of industry, particularly in power engineering, has created a global problem in the area of ecological-disposal of radioactive waste (RAW). In general, three methods for reprocessing and disposal of RAW with high-level radionuclides are used: reservoir storage; burial in boreholes; and vitrification (solidification into glass blocks). Analysis of the recent methods of high level RAW (HLRAW) localization has shown that the most reliable method for long-term storage is vitrification. Vitrification allows to decrease by more than one order of magnitude the volume of HLRAW which is intended for long-term storage, and also to decrease leaching rates by 3-4 orders. This method includes incorporation of waste into physicochemical conglomerates during glass processing from active nuclides and neutral charging materials. Usually, this method consists of multistage processes. One-stage vitrification methods are seldom considered. (author)

Radioactive waste management at AECL

International Nuclear Information System (INIS)

Gadsby, R.D.; Allan, C.J.

2003-01-01

AECL has maintained an active program in radioactive waste management since 1945, when the Canadian nuclear program commenced activities at the Chalk River Laboratories (CRL). Waste management activities have included operation of waste management storage and processing facilities at AECL's CRL and Whiteshell Laboratories (WL); operation of the Low Level Radioactive Waste Management Office on behalf of Natural Resources Canada to resolve historic radioactive waste problems (largely associated with radioactive ore recovery, transport and processing operations) that are the responsibility of the Federal Government; development of the concept and related technology for geological disposal of Canada's nuclear fuel waste; development of the Intrusion-Resistant Underground Structure (IRUS) disposal concept for low-level nuclear waste; development of dry storage technology for the interim storage of used fuel; and development and assessment of waste processing technology for application in CANDU nuclear power plants and at CRL and WL. Today these activities are continuing. In addition, AECL is: preparing to decommission the nuclear facilities at WL; carrying out a number of smaller decommissioning projects at CRL; putting in place projects to upgrade the low-level liquid waste processing capabilities of the CRL Waste Treatment Centre, recover and process highly active liquid wastes currently in storage, and recover, condition and improve the storage of selected fuel wastes currently stored in below-ground standpipes in the CRL waste management areas; and assessing options for additional remediation projects to improve the management of other wastes currently in storage and to address environmental contamination from past practices. (author)

Radioactive gaseous waste processing device

International Nuclear Information System (INIS)

Maruki, Shin-ichiro.

1991-01-01

Radioactive off-gases extracted from a turbine main condensator by using an air extractor flown by way of an off-gas preheater and enter to an off-gas recombiner. Hydrogen in the off-gases is combined with oxygen into steams by the effect of catalysts in the off-gas recombiner. In this case, the off-gases are heated to a high temperature by the reaction heat due to the effect of the catalysts and discharged from the exit of the off-gas recombiner. The off-gases at a high temperature are returned once to the off-gas preheater at the upstream to be used as a heat source for the off-gas preheater. With such a constitution, since the amount of heat for exchange required for heating to about 160degC can be supplied, a heated steam supply device which has been disposed to the off-gas preheater can be saved. Further, the off-gases cooled through heat exchange upon heating the off-gas preheater are flown to the off-gas condensator and the steams are returned into the condensates. Since cooled off-gases enter into a cooling water supply device, the load thereof can be reduced compared with a conventional case. (T.M.)

Laboratory characterization and vitrification of Hanford radioactive high-level waste

International Nuclear Information System (INIS)

Tingey, J.M.; Elliott, M.L.; Larson, D.E.; Morrey, E.V.

1991-05-01

Radioactive high-level wastes generated at the Department of Energy's Hanford Site are stored in underground carbon steel tanks. Two double-shell tanks contain neutralized current acid waste (NCAW) from the reprocessing of irradiated nuclear fuel in the Plutonium and Uranium Extraction (PUREX) Plant. The tanks were sampled for characterization and waste immobilization process/product development. The high-level waste generated in PUREX was denitrated with sugar to form current acid waste (CAW). The CAW was ''neutralized'' to a pH of approximately 14 by adding sodium hydroxide to reduce corrosion of the tanks. This ''neutralized'' waste is called Neutralized Current Acid Waste. Both precipitated solids and liquids are stored in the NCAW waste tanks. The NCAW contains small amounts of plutonium and most of the fission products and americium from the irradiated fuel. NCAW also contains stainless steel corrosion products, and iron and sulfate from the ferrous sulfamate reductant used in the PUREX process. The NCAW will be retrieved, pretreated, and immobilized prior to final disposal. Pretreatment consists of water washing the precipitated NCAW solids for sulfate and soluble salts removal as a waste reduction step prior to vitrification. This waste is expected to be the first waste type to be retrieved and vitrified in the Hanford Waste Vitrification Plant (HWVP). A characterization plan was developed that details the processing of the small-volume NCAW samples through retrieval, pretreatment and vitrification process steps. Physical, rheological, chemical, and radiochemical properties were measured throughout these process steps. The results of nonradioactive simulant tests were used to develop appropriate pretreatment and vitrification process steps. The processing and characterization of simulants and actual NCAW tank samples are used to evaluate the operation of these processes. 3 refs., 1 fig., 4 tabs

Evaluation of S-type fiberglass composites for use in high-level radioactive waste environments

International Nuclear Information System (INIS)

Parra, S.A.

1996-01-01

Two types of S-type fiberglass materials were evaluated for use in a high-level radioactive waste environment. The S-type fiberglass composites tested were in the form of tubes and were exposed to a simulated high-level radioactive waste environment consisting of corrosive chemicals, high gamma radiation, and elevated temperatures. The physical properties of the exposed and unexposed tube samples were compared to determine the effects of the simulated environment on the S-type fiberglass composites

Study on treating of low-level radioactive reactor wastewater by combined membrane process (UF-RO)

International Nuclear Information System (INIS)

Lu Yunyun; Cao Qiru; Chen Yunming; Huang Lijuan; Bai Xiaofeng; Li Bing; Feng Liang

2013-01-01

According to the characteristics of radionuclide exists in the low-level radioactive reactor waste water from HFETR, we use a new combined membrane process separation technology to study the efficient treating of low-lever radioactive reactor wastewater. First, the prepared the simulated wastewater contained Cs + , Sr 2+ , CO 2+ , Ni 2+ , and Fe 3+ . Then, we sequentially investigated the pressure, ion concentration, pH value and EDTA, which have effects on the desalination rate of membrane processing metal ions in wastewater. The results show that: in the condition of pH = 7, and added 0.15 mol/L EDTA, the simulated wastewater separated by UF-RO, desalination rates of Cs + , Sr 2+ , CO 2+ , Ni 2+ and Fe 3+ are all above 95%; In the subsequent trials, adding 0.15 mol/L EDTA into the radioactive residuary solution, and then treating by UF-RO-RO, the decontamination efficiency can reach 95.7%. (authors)