Radioactive waste isolation in salt: geochemistry of brine in rock salt in temperature gradients and gamma-radiation fields - a selective annotated bibliography

Energy Technology Data Exchange (ETDEWEB)

Hull, A.B.; Williams, L.B.

1985-07-01

Evaluation of the extensive research concerning brine geochemistry and transport is critically important to successful exploitation of a salt formation for isolating high-level radioactive waste. This annotated bibliography has been compiled from documents considered to provide classic background material on the interactions between brine and rock salt, as well as the most important results from more recent research. Each summary elucidates the information or data most pertinent to situations encountered in siting, constructing, and operating a mined repository in salt for high-level radioactive waste. The research topics covered include the basic geology, depositional environment, mineralogy, and structure of evaporite and domal salts, as well as fluid inclusions, brine chemistry, thermal and gamma-radiation effects, radionuclide migration, and thermodynamic properties of salts and brines. 4 figs., 6 tabs.

The HAW project. Demonstrative disposal of high-level radioactive wastes in the Asse salt mine

International Nuclear Information System (INIS)

Rothfuchs, T.; Duijves, K.

1988-04-01

Since 1968 the GSF has been carrying out research and development programs for the final disposal of high-level radioactive waste (HAW) in salt formations. The heat producing waste has been simulated so far by means of electrical heaters and also cobalt-60-sources. In order to improve the final concept for HAW disposal in salt formations the complete technical system of an underground repository is to be tested in a one-to-one scale test facility. To satisfy the test objectives thirty high radioactive canisters containing the radionuclides Cs-137 and Sr-90 will be emplaced in six boreholes located in two test galleries at the 800 m-level in the Asse salt mine. The duration of testing will be approximately five years. For the handling of the radioactive canisters and their emplacement into the boreholes a system consisting of transportation casks, transportation vehicle, disposal machine, and borehole slider will be developed and tested. The actual scientific investigation program is based on the estimation and observation of the interaction between the radioactive canisters and the rock salt. This program includes measurement of thermally and radiolytically induced water and gas release from the rock salt and the radiolytical decomposition of salt minerals. Also the thermally induced stress and deformation fields in the surrounding rock mass will be investigated carefully. (orig./HP)

Molten salt combustion of radioactive wastes

International Nuclear Information System (INIS)

Grantham, L.F.; McKenzie, D.E.; Richards, W.L.; Oldenkamp, R.D.

1976-01-01

The Atomics International Molten Salt Combustion Process reduces the weight and volume of combustible β-γ contaminated transuranic waste by utilizing air in a molten salt medium to combust organic materials, to trap particulates, and to react chemically with any acidic gases produced during combustion. Typically, incomplete combustion products such as hydrocarbons and carbon monoxide are below detection limits (i.e., 3 ) is directly related to the sodium chloride vapor pressure of the melt; >80% of the particulate is sodium chloride. Essentially all metal oxides (combustion ash) are retained in the melt, e.g., >99.9% of the plutonium, >99.6% of the europium, and >99.9% of the ruthenium are retained in the melt. Both bench-scale radioactive and pilot scale (50 kg/hr) nonradioactive combustion tests have been completed with essentially the same results. Design of three combustors for industrial applications are underway

Modeling internal deformation of salt structures targeted for radioactive waste disposal

International Nuclear Information System (INIS)

Chemia, Zurab

2008-01-01

This thesis uses results of systematic numerical models to argue that externally inactive salt structures, which are potential targets for radioactive waste disposal, might be internally active due to the presence of dense layers or blocks within a salt layer. The three papers that support this thesis use the Gorleben salt diapir (NW Germany), which was targeted as a future final repository for high-grade radioactive waste, as a general guideline. The first two papers present systematic studies of the parameters that control the development of a salt diapir and how it entrains a dense anhydrite layer. Results from these numerical models show that the entrainment of a dense anhydrite layer within a salt diapir depends on four parameters: sedimentation rate, viscosity of salt, perturbation width and the stratigraphic location of the dense layer. The combined effect of these four parameters, which has a direct impact on the rate of salt supply (volume/area of the salt that is supplied to the diapir with time), shape a diapir and the mode of entrainment. Salt diapirs down-built with sedimentary units of high viscosity can potentially grow with an embedded anhydrite layer and deplete their source layer (salt supply ceases). However, when salt supply decreases dramatically or ceases entirely, the entrained anhydrite layer/segments start to sink within the diapir. In inactive diapirs, sinking of the entrained anhydrite layer is inevitable and strongly depends on the rheology of the salt, which is in direct contact with the anhydrite layer. During the post-depositional stage, if the effective viscosity of salt falls below the threshold value of around 10 18 -10 19 Pa s, the mobility of anhydrite blocks might influence any repository within the diapir. However, the internal deformation of the salt diapir by the descending blocks decreases with increase in effective viscosity of salt. The results presented in this thesis suggest that it is highly likely that salt structures

Analysis of water content in salt deposits: its application to radioactive waste storage

International Nuclear Information System (INIS)

Cuevas Muller, C. de la.

1993-01-01

The salt deposits as radioactive storage medium are analyzed. This report studies the physical-chemical characteristics of water into salts deposits, its implications for the safety of the repository, and the transport water release mechanism. The last part analyzes the geochemical numerical data of correlation analysis, geostatistics analysis and interpretation of statistical data

Laboratory scale vitrification of low-level radioactive nitrate salts and soils from the Idaho National Engineering Laboratory

International Nuclear Information System (INIS)

Shaw, P.; Anderson, B.

1993-07-01

INEL has radiologically contaminated nitrate salt and soil waste stored above and below ground in Pad A and the Acid Pit at the Radioactive Waste Management Complex. Pad A contain uranium and transuranic contaminated potassium and sodium nitrate salts generated from dewatered waste solutions at the Rocky Flats Plant. The Acid Pit was used to dispose of liquids containing waste mineral acids, uranium, nitrate, chlorinated solvents, and some mercury. Ex situ vitrification is a high temperature destruction of nitrates and organics and immobilizes hazardous and radioactive metals. Laboratory scale melting of actual radionuclides containing INEL Pad A nitrate salts and Acid Pit soils was performed. The salt/soil/additive ratios were varied to determine the range of glass compositions (resulted from melting different wastes); maximize mass and volume reduction, durability, and immobilization of hazardous and radioactive metals; and minimize viscosity and offgas generation for wastes prevalent at INEL and other DOE sites. Some mixtures were spiked with additional hazardous and radioactive metals. Representative glasses were leach tested and showed none. Samples spiked with transuranic showed low nuclide leaching. Wasteforms were two to three times bulk densities of the salt and soil. Thermally co-processing soils and salts is an effective remediation method for destroying nitrate salts while stabilizing the radiological and hazardous metals they contain. The measured durability of these low-level waste glasses approached those of high-level waste glasses. Lab scale vitrification of actual INEL contaminated salts and soils was performed at General Atomics Laboratory as part of the INEL Waste Technology Development and Environmental Restoration within the Buried Waste Integrated Demonstration Program

Pilot-Scale Removal Of Fluoride From Legacy Plutonium Materials Using Vacuum Salt Distillation

International Nuclear Information System (INIS)

Pierce, R. A.; Pak, D. J.

2012-01-01

Between September 2009 and January 2011, the Savannah River National Laboratory (SRNL) and HB-Line designed, developed, tested, and successfully deployed a system for the distillation of chloride salts. In 2011, SRNL adapted the technology for the removal of fluoride from fluoride-bearing salts. The method involved an in situ reaction between potassium hydroxide (KOH) and the fluoride salt to yield potassium fluoride (KF) and the corresponding oxide. The KF and excess KOH can be distilled below 1000°C using vacuum salt distillation (VSD). The apparatus for vacuum distillation contains a zone heated by a furnace and a zone actively cooled using either recirculated water or compressed air. During a vacuum distillation operation, a sample boat containing the feed material is placed into the apparatus while it is cool, and the system is sealed. The system is evacuated using a vacuum pump. Once a sufficient vacuum is attaned, heating begins. Volatile salts distill from the heated zone to the cooled zone where they condense, leaving behind the non-volatile material in the feed boat. Studies discussed in this report were performed involving the use of non-radioactive simulants in small-scale and pilot-scale systems as well as radioactive testing of a small-scale system with plutonium-bearing materials. Aspects of interest include removable liner design considerations, boat materials, in-line moisture absorption, and salt deposition

Concepts and Technologies for Radioactive Waste Disposal in Rock Salt

Directory of Open Access Journals (Sweden)

Wernt Brewitz

2007-01-01

Full Text Available In Germany, rock salt was selected to host a repository for radioactive waste because of its excellent mechanical properties. During 12 years of practical disposal operation in the Asse mine and 25 years of disposal in the disused former salt mine Morsleben, it was demonstrated that low-level wastes (LLW and intermediate-level wastes (ILW can be safely handled and economically disposed of in salt repositories without a great technical effort. LLW drums were stacked in old mining chambers by loading vehicles or emplaced by means of the dumping technique. Generally, the remaining voids were backfilled by crushed salt or brown coal filter ash. ILW were lowered into inaccessible chambers through a borehole from a loading station above using a remote control.Additionally, an in-situ solidification of liquid LLW was applied in the Morsleben mine. Concepts and techniques for the disposal of heat generating high-level waste (HLW are advanced as well. The feasibility of both borehole and drift disposal concepts have been proved by about 30 years of testing in the Asse mine. Since 1980s, several full-scale in-situ tests were conducted for simulating the borehole emplacement of vitrified HLW canisters and the drift emplacement of spent fuel in Pollux casks. Since 1979, the Gorleben salt dome has been investigated to prove its suitability to host the national final repository for all types of radioactive waste. The “Concept Repository Gorleben” disposal concepts and techniques for LLW and ILW are widely based on the successful test operations performed at Asse. Full-scale experiments including the development and testing of adequate transport and emplacement systems for HLW, however, are still pending. General discussions on the retrievability and the reversibility are going on.

Using the characteristics of the structure of the upper Frasnian salt bearing formation in prospecting for oil deposits in the Pripyat depression

Energy Technology Data Exchange (ETDEWEB)

Yeroshina, D.M.; Kislik, V.Z.; Sinichka, A.M.; Vysotskiy, E.A.

1984-01-01

The possibility is shown of using the structure of the upper Frasnian salt bearing formation to establish ancient depressions and uplifts. A gradual wedge out of the lower strata of rock salt towards the domes of the ancient uplifts occurs. It is recommended that several reflecting levels be built up in the base of the salt bearing formation to record the behavior of these strata.

Salt removal from tanks containing high-level radioactive waste

International Nuclear Information System (INIS)

Kiser, D.L.

1981-01-01

At the Savannah River Plant (SRP), there are 23 waste storage tanks containing high-level radioactive wastes that are to be retired. These tanks contain about 23 million liters of salt and about 10 million liters of sludge, that are to be relocated to new Type III, fully stress-relieved tanks with complete secondary containment. About 19 million liters of salt cake are to be dissolved. Steam jet circulators were originally proposed for the salt dissolution program. However, use of steam jet circulators raised the temperature of the tank contents and caused operating problems. These included increased corrosion risk and required long cooldown periods prior to transfer. Alternative dissolution concepts were investigated. Examination of mechanisms affecting salt dissolution showed that the ability of fresh water to contact the cake surface was the most significant factor influencing dissolution rate. Density driven and mechanical agitation techniques were developed on a bench scale and then were demonstrated in an actual waste tank. Actual waste tank demonstrations were in good agreement with bench-scale experiments at 1/85 scale. The density driven method utilizes simple equipment, but leaves a cake heel in the tank and is hindered by the presence of sludge or Zeolite in the salt cake. Mechanical agitation overcomes the problems found with both steam jet circulators and the density driven technique and is the best method for future waste tank salt removal

Flotation of copper-bearing shale in solutions of inorganic salts and organic reagents

Directory of Open Access Journals (Sweden)

Ratajczak Tomasz

2017-01-01

Full Text Available Flotation data on copper-bearing shale in aqueous solutions of inorganic electrolytes (NaCl, Na2SO4, KPF6, NH4Cl and organic reagents (ethylamine, propylamine as frothers were presented and discussed. The relationships between shale flotation, surface tension of aqueous solution and foam height during bubbling with air in the flotation system were presented. It has been found that flotation of shale in the presence of inorganic salts the yield was directly proportional to the surface tension of the aqueous solution of salt and inversely proportional to the height of the foam. On the other hand, for organic reagents solutions (short chain amines, a reverse effect has been observed in relation to the inorganic compounds studied, that is the yield of copper-bearing shale flotation and the foam height were inversely proportional to the surface tension of the amine solution.

De-chlorination and solidification of radioactive LiCl waste salt by using SiO_2-Al_2O_3-P_2O_5 (SAP) inorganic composite including B_2O_3 component

International Nuclear Information System (INIS)

Lee, Ki Rak; Park, Hwan-Seo; Cho, In-Hak; Choi, Jung-Hoon; Eun, Hee-Chul; Lee, Tae-Kyo; Han, Seung Youb; Ahn, Do-Hee

2017-01-01

SAP (SiO_2-Al_2O_3-P_2O_5) composite has been recently studied in KAERI to deal with the immobilization of radioactive salt waste, one of the most problematic wastes in the pyro-chemical process. Highly unstable salt waste was successfully converted into stable compounds by the dechlorination process with SAPs, and then a durable waste form with a high waste loading was produced when adding glassy materials to dechlorination product. In the present study, U-SAP composite which is SAP bearing glassy component (Boron) was synthesized to remove the adding and mixing steps of glassy materials for a monolithic wasteform. With U-SAPs prepared by a sol-gel process, a series of wasteforms were fabricated to identify a proper reaction condition. Physical and chemical properties of dechlorination products and U-SAP wasteforms were characterized by XRD, DSC, SEM, TGA and PCT-A. A U-SAP wasteform showed suitable properties as a radioactive wasteform such as dense surface morphology, high waste loading, and high durability at the optimized U-SAP/salt ratio 2.

Blending Of Radioactive Salt Solutions In Million Gallon Tanks

Energy Technology Data Exchange (ETDEWEB)

Leishear, Robert A.; Lee, Si Y.; Fowley, Mark D.; Poirier, Michael R.

2012-12-10

Research was completed at Savannah River National Laboratory (SRNL) to investigate processes related to the blending of radioactive, liquid waste, salt solutions in 4920 cubic meter, 25.9 meter diameter storage tanks. One process was the blending of large salt solution batches (up to 1135 ? 3028 cubic meters), using submerged centrifugal pumps. A second process was the disturbance of a settled layer of solids, or sludge, on the tank bottom. And a third investigated process was the settling rate of sludge solids if suspended into slurries by the blending pump. To investigate these processes, experiments, CFD models (computational fluid dynamics), and theory were applied. Experiments were performed using simulated, non-radioactive, salt solutions referred to as supernates, and a layer of settled solids referred to as sludge. Blending experiments were performed in a 2.44 meter diameter pilot scale tank, and flow rate measurements and settling tests were performed at both pilot scale and full scale. A summary of the research is presented here to demonstrate the adage that, ?One good experiment fixes a lot of good theory?. Experimental testing was required to benchmark CFD models, or the models would have been incorrectly used. In fact, CFD safety factors were established by this research to predict full-scale blending performance. CFD models were used to determine pump design requirements, predict blending times, and cut costs several million dollars by reducing the number of required blending pumps. This research contributed to DOE missions to permanently close the remaining 47 of 51 SRS waste storage tanks.

Radioactive waste disposal in the Gorleben salt deposit

International Nuclear Information System (INIS)

Gizycki, P. von

1985-01-01

In the opinion of five experts, the protective function of the overlying rock as a barrier has turned out to be questionable after borings and measurements carried through at Gorleben. Moreover, the results have also raised doubts about the geological safety of the salt deposit as a barrier in the long run. The geological multibarrier concept must be discarded. Not only critics, but also 3 advocates from the field of official research on radioactive waste disposal state their opinion. (DG) [de

The safe disposal of radioactive wastes in geologic salt formations

International Nuclear Information System (INIS)

Kuehn, K.; Proske, R.

Geologic salt formations appear to be particularly suitable for final storage. Their existance alone - the salt formations in Northern Germany are more than 200 million years old - is proof of their stability and of their isolation from biological cycles. In 1967 the storage of LAW and later, in 1972, of MAW was started in the experimental storage area Asse, south-east of Braunschweig, after the necessary technical preparations had been made. In more than ten years of operation approx. 114,000 drums of slightly active and 1,298 drums of medium-active wastes were deposited without incident. Methods have been developed for filling the available caverns with wastes and salt to ensure the security of long term disposal without supervision. Tests with electric heaters for simulation of heat-generating highly active wastes confirm the good suitability of salt formations for storing these wastes. Safety analyses for the operating time as well as for the long term phase after closure of the final storage area, which among others also comprise the improbable ''greatest expected accident'', namely break through of water, are carried out and confirm the safety of ultimate storage of radioactive wastes in geological salt formations. (orig./HP) [de

Low disposal of radioactive wastes in salt formations of the Federal Republic of Germaany

International Nuclear Information System (INIS)

Albrecht, E.

1980-01-01

The salt formations of northern Europe are generally suitable for the storage of radioactive wastes because the region is largely free from earthquakes and the salt formations known as diapires provide effective hydrological sealing. The Federal Republic of Germany employed the Asse Salt Mine of Lower Saxony for research in waste storage. More recently, exploratory work has begun on the construction of a large recycling and disposal plant at the Gorleben salt dome. The geology, hydrology, rock mechanics, and seismicity of the two sites are briefly discussed, including a discussion of experiences gained so far from the Asse site. 11 refs

An improvement study on the closed chamber distillation system for recovery of renewable salts from salt wastes containing radioactive rare earth compounds

International Nuclear Information System (INIS)

Eun, H.C.; Cho, Y.Z.; Lee, T.K.; Kim, I.T.; Park, G.I.; Lee, H.S.

2013-01-01

In this paper, an improvement study on the closed chamber distillation system for recovery of renewable salts from salt wastes containing radioactive rare earth compounds was performed to determine optimum operating conditions. It was very important to maintain the pressure in the distillation chamber below 10 Torr for a high efficiency (salt recovery >99 %) of the salt distillation. This required increasing the salt vaporization and condensation rates in the distillation system. It was confirmed that vaporization and condensation rates could be improved controlling the given temperature of top of the condensation chamber. In the distillation tests of the salt wastes containing rare earth compounds, the operation time at a given temperature was greatly reduced changing the given temperature of top of the condensation chamber from 780 to 700 deg C. (author)

Blending of Radioactive Salt Solutions in Million Gallon Tanks - 13002

International Nuclear Information System (INIS)

Leishear, Robert A.; Lee, Si Y.; Fowley, Mark D.; Poirier, Michael R.

2013-01-01

Research was completed at Savannah River National Laboratory (SRNL) to investigate processes related to the blending of radioactive, liquid waste, salt solutions in 4920 cubic meter, 25.9 meter diameter storage tanks. One process was the blending of large salt solution batches (up to 1135 - 3028 cubic meters), using submerged centrifugal pumps. A second process was the disturbance of a settled layer of solids, or sludge, on the tank bottom. And a third investigated process was the settling rate of sludge solids if suspended into slurries by the blending pump. To investigate these processes, experiments, CFD models (computational fluid dynamics), and theory were applied. Experiments were performed using simulated, non-radioactive, salt solutions referred to as supernates, and a layer of settled solids referred to as sludge. Blending experiments were performed in a 2.44 meter diameter pilot scale tank, and flow rate measurements and settling tests were performed at both pilot scale and full scale. A summary of the research is presented here to demonstrate the adage that, 'One good experiment fixes a lot of good theory'. Experimental testing was required to benchmark CFD models, or the models would have been incorrectly used. In fact, CFD safety factors were established by this research to predict full-scale blending performance. CFD models were used to determine pump design requirements, predict blending times, and cut costs several million dollars by reducing the number of required blending pumps. This research contributed to DOE missions to permanently close the remaining 47 of 51 SRS waste storage tanks. (authors)

Blending of Radioactive Salt Solutions in Million Gallon Tanks - 13002

Energy Technology Data Exchange (ETDEWEB)

Leishear, Robert A.; Lee, Si Y.; Fowley, Mark D.; Poirier, Michael R. [Savannah River National Laboratory, Aiken. S.C., 29808 (United States)

2013-07-01

Research was completed at Savannah River National Laboratory (SRNL) to investigate processes related to the blending of radioactive, liquid waste, salt solutions in 4920 cubic meter, 25.9 meter diameter storage tanks. One process was the blending of large salt solution batches (up to 1135 - 3028 cubic meters), using submerged centrifugal pumps. A second process was the disturbance of a settled layer of solids, or sludge, on the tank bottom. And a third investigated process was the settling rate of sludge solids if suspended into slurries by the blending pump. To investigate these processes, experiments, CFD models (computational fluid dynamics), and theory were applied. Experiments were performed using simulated, non-radioactive, salt solutions referred to as supernates, and a layer of settled solids referred to as sludge. Blending experiments were performed in a 2.44 meter diameter pilot scale tank, and flow rate measurements and settling tests were performed at both pilot scale and full scale. A summary of the research is presented here to demonstrate the adage that, 'One good experiment fixes a lot of good theory'. Experimental testing was required to benchmark CFD models, or the models would have been incorrectly used. In fact, CFD safety factors were established by this research to predict full-scale blending performance. CFD models were used to determine pump design requirements, predict blending times, and cut costs several million dollars by reducing the number of required blending pumps. This research contributed to DOE missions to permanently close the remaining 47 of 51 SRS waste storage tanks. (authors)

Radioactive waste and special waste disposal in salt domes - phoney waste management solutions

International Nuclear Information System (INIS)

Grimmel, E.

1990-01-01

The paper tries to make aware of the fact that an indefinite safe disposal of anthropogeneous wastes in underground repositories is impossible. Suspicion is raised that the Gorleben-Rambow salt dome has never been studied for its suitability as a repository, but that it was simply taken for granted. Safety analyses are meant only to conceal uncertainty. It is demanded to immediately opt out of the ultimate disposal technique for radioactive and special wastes in salt caverns. (DG) [de

Characteristics of solidified products containing radioactive molten salt waste.

Science.gov (United States)

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

2007-11-01

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

Radiolysis salt phenomenology: application to storage of high level radioactive waste

International Nuclear Information System (INIS)

Akram, Najib

1993-01-01

In France, rock salt is a candidate repository for highly radioactive waste. Rock salt contains water and adsorbed gases which can be released in boreholes after heating due to vitrified wastes. In addition, waste-induced irradiation in near-field conditions induce radiolytic reactions which also contribute to gas release. The aim of this work is to understand and evaluate the effects of heat and irradiation produced by waste containers in a deep disposal, primarily concerning gas production. This is justified by the impact of gases on long-term safety: toxicity, explosibility, chemical reactivity, pressure build-up. We have evidenced the influence of integrated dose, filling gases, temperature and grain size on an homogeneous medium (Asse Mine rock salt). We have then studied heterogeneous samples, which allowed to determine the influence of the chemical and mineralogical composition of rock salt (bedded rock salt from the Mine de Potasse d'Alsace). The role played by organic matter on gas production is important, leading for instance to high consumption rates of oxygen. Through this study, we have also considered the behaviour of clay-rich materials under irradiation. Our results constitute important bases for the future modelling of the phenomena which will take place in the near-field of a rock salt-type repository, especially concerning its long-term safety. (author) [fr

Release consequence analysis for a hypothetical geologic radioactive waste repository in salt

International Nuclear Information System (INIS)

1979-08-01

One subtask conducted under the INFCE program is to evaluate and compare the health and safety impacts of different fuel cycles in which all radioactive wastes (except those from mining and milling) are placed in a geologic repository in salt. To achieve this objective, INFCE Working Group 7 examined the radiologic dose to humans from geologic repositories containing waste arisings as defined for seven reference fuel cycles. This report examines the release consequences for a generic waste repository in bedded salt. The top of the salt formation and the top of the repository are assumed to be 250 and 600 m, respectively, below the surface. The hydrogeologic structure above the salt consists of two aquifers and two aquitards. The aquifers connect to a river 6.2 km from the repository. The regional gradient to the river is 1 m/km in all aquifers. Hydrologic, transport, and dose models were used to model two release scenarios for each fuel cycle, one without a major disturbance and one in which a major geologic perturbation breached the repository immediately after it was sealed. The purpose of the modeling was to predict the rate of transport of radioactive contaminants from the repository through the geosphere to the biosphere, and to determine the potential dose to humans. Of the many radionuclides in the waste, only 129 I and 226 Ra arrived at the river in sufficient concentrations for a measurable dose calculation. Radionuclide concentrations in the ground water pose no threat to man because the ground water is a concentrated brine and it is diluted by a factor of 10 6 to 10 7 upon entering the river

De-chlorination and solidification of radioactive LiCl waste salt by using SiO{sub 2}-Al{sub 2}O{sub 3}-P{sub 2}O{sub 5} (SAP) inorganic composite including B{sub 2}O{sub 3} component

Energy Technology Data Exchange (ETDEWEB)

Lee, Ki Rak; Park, Hwan-Seo; Cho, In-Hak; Choi, Jung-Hoon; Eun, Hee-Chul; Lee, Tae-Kyo; Han, Seung Youb; Ahn, Do-Hee [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2017-09-15

SAP (SiO{sub 2}-Al{sub 2}O{sub 3}-P{sub 2}O{sub 5}) composite has been recently studied in KAERI to deal with the immobilization of radioactive salt waste, one of the most problematic wastes in the pyro-chemical process. Highly unstable salt waste was successfully converted into stable compounds by the dechlorination process with SAPs, and then a durable waste form with a high waste loading was produced when adding glassy materials to dechlorination product. In the present study, U-SAP composite which is SAP bearing glassy component (Boron) was synthesized to remove the adding and mixing steps of glassy materials for a monolithic wasteform. With U-SAPs prepared by a sol-gel process, a series of wasteforms were fabricated to identify a proper reaction condition. Physical and chemical properties of dechlorination products and U-SAP wasteforms were characterized by XRD, DSC, SEM, TGA and PCT-A. A U-SAP wasteform showed suitable properties as a radioactive wasteform such as dense surface morphology, high waste loading, and high durability at the optimized U-SAP/salt ratio 2.

Salt Repository Project site study plan for background environmental radioactivity: Revision 1

International Nuclear Information System (INIS)

1987-12-01

The Site Study Plan for Background Environmental Radioactivity describes a field program consisting of an initial radiological survey and a radiological sampling program. The field program includes measurement of direct radiation and collection and analysis of background radioactivity samples of air, precipitation, soil, water, milk, pasture grass, food crops, meat, poultry, game, and eggs. The plan describes for each study the need for the study, the study design, data management, and use, schedule of proposed activities, and quality assurance requirements. These studies will provide data needed to satisfy requirements contained in, or derived from, the Salt Repository Project Requirements Document. 43 refs., 10 figs., 7 tabs

The HAW-Project. Test disposal of highly radioactive radiation sources in the Asse salt mine. Final report

International Nuclear Information System (INIS)

Rothfuchs, T.; Cuevas, C. de las; Donker, H.; Feddersen, H.K.; Garcia-Celma, A.; Gies, H.; Goreychi, M.; Graefe, V.; Heijdra, J.; Hente, B.; Jockwer, N.; LeMeur, R.; Moenig, J.; Mueller, K.; Prij, J.; Regulla, D.; Smailos, E.; Staupendahl, G.; Till, E.; Zankl, M.

1995-01-01

In order to improve the final concept for the disposal of high-level radioactive waste (HAW) in boreholes drilled into salt formation plans were developed a couple of years ago for a full scale testing of the complete technical system of an underground repository. To satisfy the test objectives, thirty highly radioactive radiation sources were planned to be emplaced in six boreholes located in two test galleries at the 800-m-level in the Asse salt mine. A duration of testing of approximately five years was envisaged. Because of licensing uncertainties the German Federal Government decided on December 3rd, 1992 to stop all activities for the preparation of the test disposal immediately. In the course of the preparation of the test disposal, however, a system, necessary for handling of the radiation sources was developed and installed in the Asse salt mine and two non-radioactive reference tests with electrical heaters were started in November 1988. These tests served for the investigation of thermal effects in comparison to the planned radioactive tests. An accompanying scientific investigation programme performed in situ and in the laboratory comprises the estimation and observation of the thermal, radiation-induced, and mechanical interaction between the rock salt and the electrical heaters and the radiation sources, respectively. The laboratory investigations are carried out at Braunschweig (FRG), Petten (NL), Saclay (F) and Barcelona (E). As a consequence of the premature termination of the project the working programme was revised. The new programme agreed to by the project partners included a controlled shutdown of the heater tests in 1993 and a continuation of the laboratory activities until the end of 1994. (orig.)

Forecasting the space-time stability of radioactive waste isolation in salt formations

International Nuclear Information System (INIS)

Anderson, E.B.; Karelin, A.I.; Krivokhatsiy, A.S.; Savonenkov, V.G.

1992-01-01

The possibilities to use salt formations for radioactive waste isolation are realized by creating shaft-type underground repositories in these rocks in Germany and the USA. The burial safety of low- and intermediate-level wastes for several hundred years have been substantiated for the sites chosen. Specialists of different countries presented positive properties of rock salt as a medium for isolation of radionuclides. A rich experience in building subsurface structures for different purposes in salts is accumulated in our country. Detailed investigations of salt formation have shown that far from all the saliferous areas and structures may be used for constructing burial sites. One of the reasons for this limitation is a sharp difference of individual deposits by their compositions, structures, the character of deposition and the conditions of formation. The geological criteria of safety acquire special significance in connection with the necessity to isolate radionuclides having the half-loves more than 1000 years. The time intervals required for stable isolation make up millions of years and cover great cycles of the evolution of the Earth surface and biosphere

Requirements for a long-term safety certification for chemotoxic substances stored in a final storage facility for high radioactive and heat-generating radioactive waste in rock salt formations

International Nuclear Information System (INIS)

Tholen, M.; Hippler, J.; Herzog, C.

2007-01-01

Within the scope of a project funded by the German Federal Ministry of Economics and Technology (Bundesministerium fuer Wirtschaft und Technologie, BMWi), a safety certification concept for a future permanent final storage for high radioactive and heat-generating radioactive waste (HAW disposal facility) in rock salt formations is being prepared. For a reference concept, compliance with safety requirements in regard to operational safety as well as radiological and non-radiological protection objectives related to long-term safety, including ground water protection, will be evaluated. This paper deals with the requirements for a long-term safety certification for the purpose of protecting ground water from chemotoxic substances. In particular, longterm safety certifications for the permanent disposal of radioactive waste in a HAW disposal facility in rock salt formations and for the dumping of hazardous waste in underground storage facilities in rock salt formations are first discussed, followed by an evaluation as to whether these methods can be applied to the long-term safety certification for chemotoxic substances. The authors find it advisable to apply the long-term safety certification for underground storage facilities to the long-term safety certification for chemotoxic substances stored in a HAW disposal facility in rock salt formations. In conclusion, a corresponding certification concept is introduced. (orig.)

Temperature distributions in a salt formation used for the ultimate disposal of high-level radioactive waste

International Nuclear Information System (INIS)

Ploumen, P.

1980-01-01

In the Federal Republic of Germany the works on waste disposal is focussed on the utilization of a salt formation for ultimate disposal of radioactive wastes. Heat released from the high-level waste will be dissipated in the salt and the surrounding geologic formations. The occuring temperature distributions will be calculated with computer codes. A survey of the developed computer codes will be shown; the results for a selected example, taking into account the loading sequence of the waste, the mine ventilation as well as an air gap between the waste and the salt, will be discussed. Furthermore it will be shown that by varying the disposal parameters, the maximum salt temperature can be below any described value. (Auth.)

The HAW-Project: Test disposal of highly radioactive radiation sources in the Asse salt mine

International Nuclear Information System (INIS)

1992-04-01

Two electrical heater tests were already started in November 1988 and are continuously surveyed in respect of the thermomechanical and geochemical response of the rock mass. Also the handling system necessary for the emplacement of 30 radioactive canisters (Sr-90 and Cs-137 sources) was developed and succesfully tested. This system consists of six multiple transport and storage casks of the type Castor-GSF-5, two above ground/below ground shuttle transport casks of the type Asse TB1, an above ground transfer station, an underground transport vehicle, a disposal machine, and a borehole slider. A laboratory investigation program on radiation effects in salt is being performed in advance to the radioactive canister emplacement. This program includes the investigation of thermally and radiolytically induced water and gas release from the rock salt and the radiolytical decomposition of salt minerals. For gamma dose and dose rate measurements in the test field measuring systems consisting of ionisation chambers as well as solid state dosemeters were developed and tested. Thermomechanical computer code validation is performed by calculational predictions and parallel investigation of the stress and displacement fields in the underground test field. (orig./HP)

Technical support for GEIS: radioactive waste isolation in geologic formations. Volume 8. Repository preconceptual design studies: salt

International Nuclear Information System (INIS)

1978-04-01

This volume, Volume 8 ''Repository Preconceptual Design Studies: Salt,'' is one of a 23-volume series, ''Technical Support for GEIS: Radioactive Waste Isolation in Geologic Formations,'' Y/OWI/TM-36, which supplements the ''Contribution to Draft Generic Environmental Impact Statement on Commercial Waste Management: Radioactive Waste Isolation in Geologic Formations,'' Y/OWI/TM-44. The series provides a more complete technical basis for the preconceptual designs, resource requirements, and environmental source terms associated with isolating commercial LWR wastes in underground repositories in salt, granite, shale and basalt. Wastes are considered from three fuel cycles: uranium and plutonium recycling, no recycling of spent fuel and uranium-only recycling. This document describes a preconceptual design for a nuclear waste storage facility in salt. The facility design consists of several chambers excavated deep within a geologic formation together with access shafts and supportive surface structures. The facility design provides for: receiving and unloading waste containers; lowering them down shafts to the mine level; transporting them to the proper storage area, and emplacing them in mined storage rooms. Drawings of the facility design are contained in TM-36/9, ''Drawings for Repository Preconceptual Design Studies: Salt.''

ICP-MS nebulizer performance for analysis of SRS high salt simulated radioactive waste tank solutions (number-sign 3053)

International Nuclear Information System (INIS)

Jones, V.D.

1997-01-01

High Level Radioactive Waste Tanks at the Savannah River Site are high in salt content. The cross-flow nebulizer provided the most stable signal for all salt matrices with the smallest signal loss/suppression due to this matrix. The DIN exhibited a serious lack of tolerance for TDS; possibly due to physical de-tuning of the nebulizer efficiency

Flotation of copper-bearing shale in solutions of inorganic salts and organic reagents

OpenAIRE

Ratajczak Tomasz

2017-01-01

Flotation data on copper-bearing shale in aqueous solutions of inorganic electrolytes (NaCl, Na2SO4, KPF6, NH4Cl) and organic reagents (ethylamine, propylamine) as frothers were presented and discussed. The relationships between shale flotation, surface tension of aqueous solution and foam height during bubbling with air in the flotation system were presented. It has been found that flotation of shale in the presence of inorganic salts the yield was directly proportional to the surface tensio...

Mineral sources of water and their influence on the safe disposal of radioactive wastes in bedded salt deposits

International Nuclear Information System (INIS)

Fallis, S.M.

1973-12-01

With the increased use of nuclear energy, there will be subsequent increases in high-level radioactive wastes such as Sr 90 , Cs 137 , and Pu 239 . Several agencies have considered the safest possible means to store or dispose of wastes in geologic environments such as underground storage in salt deposits, shale beds, abandoned dry mines, and in clay and shale pits. Salt deposits have received the most favorable attention because they exist in dry environments and because of other desirable properties of halite (its plasticity, gamma-ray shielding, heat dissipation ability, low mining cost, and worldwide abundance). Much work has been done on bedded salt deposits, particularly the Hutchinson Salt Member of the Wellington Formation at Lyons, Kansas. Salt beds heated by the decay of the radioactive wastes may release water by dehydration of hydrous minerals commonly present in evaporite sequences or water present in other forms such as fluid inclusions. More than 80 hydrous minerals are known to occur in evaporite deposits. The occurrences, total water contents (up to 63%) and dehydration temperatures (often less that 150 0 C) of these minerals are given. Since it is desirable to dispose of radioactive wastes in a dry environment, care must be taken that large quantities of water are not released through the heating of hydrous minerals. Seventy-four samples from four cores taken at Lyons, Kansas, were analyzed by x-ray diffraction. The minerals detected were halite, anhydrite, gypsum, polyhalite, dolomite, magnesite, quartz, feldspar, and the clay minerals illite, chlorite, kaolinite, vermiculite, smectite, mixed-layer clay, and corrensite (interstratified chlorite-vermiculite). Of these, gypsum, polyhalite and the clay minerals are all capable of releasing water when heated

Molten salt oxidation of mixed wastes: Separation of radioactive materials and Resource Conservation and Recovery Act (RCRA) materials

International Nuclear Information System (INIS)

Bell, J.T.; Haas, P.A.; Rudolph, J.C.

1995-01-01

The Oak Ridge National Laboratory (ORNL) is participating in a program to apply a molten salt oxidation (MSO) process to treatment of mixed (radioactive and RCRA) wastes. The salt residues from the MSO treatment will require further separations or other processing to prepare them for final disposal. A bench-scale MSO apparatus is being installed at ORNL and will be operated on real Oak Ridge wastes. The treatment concepts to be tested and demonstrated on the salt residues from real wastes are described

Fundamental study on the salt distillation from the mixtures of rare earth precipitates and LiCl-KCl eutectic salt

International Nuclear Information System (INIS)

Yang, H. C.; Eun, H. C.; Cho, Y. Z.; Lee, H. S.; Kim, I. T.

2008-01-01

An electrorefining process of spent nuclear fuel generates waste salt containing some radioactive metal chlorides. The most effective method to reduce salt waste volume is to separate radioactive metals from non-radioactive salts. A promising approach is to change radioactive metal chlorides into salt-insoluble oxides by an oxygen sparging. Following this, salt distillation process is available to effectively separate the precipitated particulate metal oxides from salt. This study investigated the distillation rates of LiCl-KCl eutectic salt under different vacuums at elevated temperatures. The first part study investigated distillation rates of eutectic salt under different vacuums at high temperatures by using thermo-gravimetric furnace system. In the second part, we tested the removal of eutectic salt from the RE precipitates by using the laboratory vacuum distillation furnace system. Investigated variables were the temperature of mixture, the degree of vacuum and the time

Rock salt as a medium for long-term isolation of radioactive wastes - a reassessment

International Nuclear Information System (INIS)

Chaturvedi, L.

1985-01-01

Rock salt has been regarded as a suitable medium for the permanent disposal of high and medium level radioactive wastes since the National Academy of Sciences recommended it in 1957. As a result of detained site-specific studies conducted for the Waste Isolation Pilot Plant (WIPP) project in New Mexico, however, several potential problems which are unique to bedded salt deposits have emerged. These include 1) the need to delineate the extent and rate of past dissolution and projections for the future, 2) the origin and significance of brines often found underlying the salt beds, 3) the rate and volume of migration of brine from the salt crystals towards the heat producing waste canisters, 4) the creep rates and implications for retrievability, and 5) the existence of potash and oil and gas resources with implications of human intrusion in the future. These questions will also be faced for sites in salt domes with added complications due to more complex structure and hydrology. The experience at WIPP shows that the site characterization process for high level waste repositories in bedded or dome salt should aim at identifying the important issues of site suitability early in the process and a clear program should be established to address these issues

Mineral sources of water and their influence on the safe disposal of radioactive wastes in bedded salt deposits

Energy Technology Data Exchange (ETDEWEB)

Fallis, S.M.

1973-12-01

With the increased use of nuclear energy, there will be subsequent increases in high-level radioactive wastes such as Sr/sup 90/, Cs/sup 137/, and Pu/sup 239/. Several agencies have considered the safest possible means to store or dispose of wastes in geologic environments such as underground storage in salt deposits, shale beds, abandoned dry mines, and in clay and shale pits. Salt deposits have received the most favorable attention because they exist in dry environments and because of other desirable properties of halite (its plasticity, gamma-ray shielding, heat dissipation ability, low mining cost, and worldwide abundance). Much work has been done on bedded salt deposits, particularly the Hutchinson Salt Member of the Wellington Formation at Lyons, Kansas. Salt beds heated by the decay of the radioactive wastes may release water by dehydration of hydrous minerals commonly present in evaporite sequences or water present in other forms such as fluid inclusions. More than 80 hydrous minerals are known to occur in evaporite deposits. The occurrences, total water contents (up to 63%) and dehydration temperatures (often less that 150/sup 0/C) of these minerals are given. Since it is desirable to dispose of radioactive wastes in a dry environment, care must be taken that large quantities of water are not released through the heating of hydrous minerals. Seventy-four samples from four cores taken at Lyons, Kansas, were analyzed by x-ray diffraction. The minerals detected were halite, anhydrite, gypsum, polyhalite, dolomite, magnesite, quartz, feldspar, and the clay minerals illite, chlorite, kaolinite, vermiculite, smectite, mixed-layer clay, and corrensite (interstratified chlorite-vermiculite). Of these, gypsum, polyhalite and the clay minerals are all capable of releasing water when heated.

User's manual and guide to SALT3 and SALT4: two-dimensional computer codes for analysis of test-scale underground excavations for the disposal of radioactive waste in bedded salt deposits

International Nuclear Information System (INIS)

Lindner, E.N.; St John, C.M.; Hart, R.D.

1984-02-01

SALT3 and SALT4 are two-dimensional analytical/displacement-discontinuity codes designed to evaluate temperatures, deformation, and stresses associated with underground disposal of radioactive waste in bedded salt. These codes were developed by the University of Minnesota for the Office of Nuclear Waste Isolation in 1979. The present documentation describes the mathematical equations of the physical system being modeled, the numerical techniques utilized, and the organization of these computer codes. The SALT3 and SALT4 codes can simulate: (a) viscoelastic behavior in pillars adjacent to excavations; (b) transversely isotropic elastic moduli such as those exhibited by bedded or stratified rock; and (c) excavation sequence. Major advantages of these codes are: (a) computational efficiency; (b) the small amount of input data required; and (c) a creep law based on laboratory experimental data for salt. The main disadvantage is that some of the assumptions in the formulation of the codes, i.e., the homogeneous elastic half-space and temperature-independent material properties, render it unsuitable for canister-scale analysis or analysis of lateral deformation of the pillars. The SALT3 and SALT4 codes can be used for parameter sensitivity analyses of two-dimensional, repository-scale, thermomechanical response in bedded salt during the excavation, operational, and post-closure phases. It is especially useful in evaluating alternative patterns and sequences of excavation or waste canister placement. SALT3 is a refinement of an earlier code, SALT, and includes a fully anelastic creep model and thermal stress routine. SALT4 is a later version, and incorporates a revised creep model which is strain-hardening

Treating radioactive effluent

International Nuclear Information System (INIS)

Kirkham, I.A.

1981-01-01

In the treatment of radioactive effluent it is known to produce a floc being a suspension of precipitates carrying radioactive species in a mother liquor containing dissolved non-radioactive salts. It is also known and accepted practice to encapsulate the floc in a solid matrix by treatment with bitumen, cement and the like. In the present invention the floc is washed with water prior to encapsulation in the solid matrix whereby to displace the mother liquor containing the dissolved non-radioactive salts. This serves to reduce the final amount of solidified radioactive waste with consequent advantages in the storage and disposal thereof. (author)

Recent studies on radiation damage formation in synthetic NaCl and natural rock salt for radioactive waste disposal applications

International Nuclear Information System (INIS)

Swyler, K.J.; Klaffky, R.W.; Levy, P.W.

1980-01-01

Radiation damage formation in natural rock salt is described as a function of irradiation temperature and plastic deformation. F-center formation decreases with increasing temperature while significant colloidal sodium formation occurs over a restricted temperature range around 150 0 C. Plastic deformation increases colloid formation; it is estimated that colloid concentrations may be increased by a factor of 3 if the rock salt near radioactive waste disposal canisters is heavily deformed. Optical bandshape analysis indicates systematic differences between the colloids formed in synthetic and natural rock salts

Measurements of the background radioactivity in a salt mine in Remolinos (Zaragoza, Spain)

International Nuclear Information System (INIS)

Gerona, G.; Morales, A.; Nunez-Lagos, R.; Puimedon, J.; Villar, J.A.

1986-01-01

Measurements of the background radioactivity in a salt mine in Remolinos (Zaragoza) have been performed. The site was 200 mwe deep in a gallery far (1 Km) from the actual mining area. The detector has been a 120 cm 3 Ge(Li) with conventional electronic and magnetic read-out. Neutron induced radiation was detected by surrounding the detector with borax. Different shieldings have been studied. Results are compared with laboratory data of Zaragoza. (author)

TRUEX partitioning from radioactive ICPP sodium bearing waste

International Nuclear Information System (INIS)

Herbst, R.S.; Brewer, K.N.; Tranter, T.J.; Todd, T.A.

1995-03-01

The Idaho Chemical Processing Plant (ICPP) located at the Idaho National Engineering Laboratory in Southeast Idaho is currently evaluating several treatment technologies applicable to waste streams generated over several decades of-nuclear fuel reprocessing. Liquid sodium bearing waste (SBW), generated primarily during decontamination activities, is one of the waste streams of interest. The TRansUranic EXtraction (TRUEX) process developed at Argonne National Laboratory is currently being evaluated to separate the actinides from SBW. On a mass basis, the amount of the radioactive species in SBW are low relative to inert matrix components. Thus, the advantage of separations is a dramatic decrease in resulting volumes of high activity waste (HAW) which must be dispositioned. Numerous studies conducted at the ICPP indicate the applicability of the TRUEX process has been demonstrated; however, these studies relied on a simulated SBW surrogate for the real waste. Consequently, a series of batch contacts were performed on samples of radioactive ICPP SBW taken from tank WM-185 to verify that actual waste would behave similarly to the simulated waste. The test results with SBW from tank WM-185 indicate the TRUEX solvent effectively extracts the actinides from the samples of actual waste. Gross alpha radioactivity, attributed predominantly to Pu and Am, was reduced from 3.14E+04 dps/mL to 1.46 dps/mL in three successive batch contacts with fresh TRUEX solvent. This reduction corresponds to a decontamination factor of DF = 20,000 or 99.995% removal of the gross a activity in the feed. The TRUEX solvent also extracted the matrix components Zr, Fe, and Hg to an appreciable extent (D Zr > 10, D Fe ∼ 2, D Hg ∼6). Iron co-extracted with the actinides can be successfully scrubbed from the organic with 0.2 M HNO 3 . Mercury can be selectively partitioned from the actinides with either sodium carbonate or nitric acid (≥ 5 M HNO 3 ) solutions

Molten salt oxidation of ion-exchange resins doped with toxic metals and radioactive metal surrogates

International Nuclear Information System (INIS)

Yang, Hee-Chul; Cho, Yong-Jun; Yoo, Jae-Hyung; Kim, Joon-Hyung; Eun, Hee-Chul

2005-01-01

Ion-exchange resins doped with toxic metals and radioactive metal surrogates were test-burned in a bench-scale molten salt oxidation (MSO) reactor system. The purposes of this study are to confirm the destruction performance of the two-stage MSO reactor system for the organic ion-exchange resin and to obtain an understanding of the behavior of the fixed toxic metals and the sulfur in the cationic exchange resins. The destruction of the organics is very efficient in the primary reactor. The primarily destroyed products such as carbon monoxide are completely oxidized in the secondary MSO reactor. The overall collection of the sulfur and metals in the two-stage MSO reactor system appeared to be very efficient. Over 99.5% of all the fixed toxic metals (lead and cadmium) and radioactive metal surrogates (cesium, cobalt, strontium) remained in the MSO reactor bottom. Thermodynamic equilibrium calculations and the XRD patterns of the spent salt samples revealed that the collected metals existed in the form of each of their carbonates or oxides, which are non-volatile species at the MSO system operating conditions. (author)

Enhancement of natural radioactivity in soils and salt-marshes surrounding a non-nuclear industrial complex

International Nuclear Information System (INIS)

Bologon, J.P.; Garca-Tenorio, R.; Garca-Leon, M.

1995-01-01

The existence of a very high extension (about 1000 ha) of phosphogypsum piles, sited in the estuary formed by the mouths of the Tinto and Odiel rivers (SW Spain), produce a quite local, but unambiguous radioactive impact in the surrounding salt-marshes. In these piles the main by-product formed in the manufacture of phosphoric acid is stored. The radioactive impact is generated by the deposition and accumulation of radionuclides from the uranium series that previously had been mainly leached or dissolved from the piles by waters that temporally can cover or cross them. Other means of impact, especially through the atmosphere, have been evaluated as negligible or not detectable

In situ-experiments on the disposal of high-level radioactive wastes (HAW) at the Asse salt mine Federal Republic of Germany

International Nuclear Information System (INIS)

Kuhn, K.; Rothfuchs, T.

1989-01-01

Deep geological salt formations are considered as being the most suitable medium for the disposal of radioactive wastes in the Federal Republic of Germany (FRG). This paper reports how, in order to develop and to prove the necessary disposal techniques, the Asse Salt Mine in the northern part of Germany is being used as a national R and D facility for the execution of representative in situ-tests. Besides the test-wise disposal of low-and medium-level radioactive waste, a series of in situ experiments was performed on the disposal of high-level radioactive waste (HAW). The so-called HAW repository is being performed from 1983 through 1994 will be the most important pilot test for the HAW repository in the FRG. During this experiment, 30 vitrified high-level radioactive heat and radiation sources will be emplaced in six underground boreholes. The duration of testing will be approximately five years. In addition to the investigations of the interactions of the heat and radiation sources and the host rock, a complete handling system for HAW-canisters is being developed and proved

Microbial Influence on the Performance of Subsurface, Salt-Based Radioactive Waste Repositories. An Evaluation Based on Microbial Ecology, Bioenergetics and Projected Repository Conditions

International Nuclear Information System (INIS)

Swanson, J.S.; Reed, D.T.; Cherkouk, A.; Arnold, T.; Meleshyn, A.; Patterson, Russ

2018-01-01

For the past several decades, the Nuclear Energy Agency Salt Club has been supporting and overseeing the characterisation of rock salt as a potential host rock for deep geological repositories. This extensive evaluation of deep geological settings is aimed at determining - through a multidisciplinary approach - whether specific sites are suitable for radioactive waste disposal. Studying the microbiology of granite, basalt, tuff, and clay formations in both Europe and the United States has been an important part of this investigation, and much has been learnt about the potential influence of microorganisms on repository performance, as well as about deep subsurface microbiology in general. Some uncertainty remains, however, around the effects of microorganisms on salt-based repository performance. Using available information on the microbial ecology of hyper-saline environments, the bioenergetics of survival under high ionic strength conditions and studies related to repository microbiology, this report summarises the potential role of microorganisms in salt-based radioactive waste repositories

Problems and risks involved in the projected storage of radioactive waste in a salt dome in the northwest of the FRG

International Nuclear Information System (INIS)

Mauthe, F.

1979-01-01

Current planning envisages long-term intermediate storage of radioactive waste and the exploration of the Gorleben salt dome by deep drilling in order to start appropriate mining work in case of favourable drilling results. The statements presented here on the problem of the 'Feasibility of ultimate storage of radioactive waste in salt deposits' (subject selected by the Government of the land Lower-Saxony) are aimed at informing the general public about the difficulties and problems involved in this waste disposal project and critically assess the arguments put forward by industry and licensing authorities in order to gain acceptance for this politically delicate project; the argumentation discussed here mainly refers to the field of geological science. (orig.) [de

Analysis of the geological stability of a hypothetical radioactive waste repository in a bedded salt formation

International Nuclear Information System (INIS)

Tierney, M.S.; Lusso, F.; Shaw, H.R.

1978-01-01

This document reports on the development of mathematical models used in preliminary studies of the long-term safety of radioactive wastes deeply buried in bedded salt formations. Two analytical approaches to estimating the geological stability of a waste repository in bedded salt are described: (a) use of probabilistic models to estimate the a priori likelihoods of release of radionuclides from the repository through certain idealized natural and anthropogenic causes, and (b) a numerical simulation of certain feedback effects of emplacement of waste materials upon ground-water access to the repository's host rocks. These models are applied to an idealized waste repository for the sake of illustration

Salt brickwork as long-term sealing in salt formations

International Nuclear Information System (INIS)

Walter, F.; Yaramanci, U.

1993-01-01

Radioactive wastes can be disposed of in deep salt formations. Rock salt is a suitable geologic medium because of its unique characteristics. Open boreholes, shafts and drifts are created to provide physical access to the repository. Long-term seals must be emplaced in these potential pathways to prevent radioactive release into the biosphere. The sealing materials must be mechanically and, most important, geochemically stable within the host rock. Salt bricks made from compressed salt-powder are understood to be the first choice long-term sealing material. Seals built of salt bricks will be ductile. Large sealing systems are built by combining the individual bricks with mortar. Raw materials for mortar are fine-grained halite powder and ground saliferous clay. This provides for the good adhesive strength of the mortar to the bricks and the high shear-strength of the mortar itself. To test the interaction of rock salt with an emplaced long-term seal, experiments will be carried out in situ, in the Asse salt mine in Germany. Simple borehole sealing experiments will be performed in horizontal holes and a complicated drift sealing experiment is planned, to demonstrate the technology of sealing a standard size drift or shaft inside a disturbed rock mass. Especially, the mechanical stability of the sealing system has to be demonstrated

Decontamination method for radioactively contaminated material

International Nuclear Information System (INIS)

Shoji, Yuichi; Mizuguchi, Hiroshi; Sakai, Hitoshi; Komatsubara, Masaru

1998-01-01

Radioactively contaminated materials having surfaces contaminated by radioactive materials are dissolved in molten salts by the effect of chlorine gas. The molten salts are brought into contact with a low melting point metal to reduce only radioactive materials by substitution reaction and recover them into the low melting point metal. Then, a low melting point metal phase and a molten salt phase are separated. The low melting point metal phase is evaporated to separate the radioactive materials from molten metals. On the other hand, other metal ions dissolved in the molten salts are reduced into metals by electrolysis at an anode and separated from the molten salts and served for regeneration. The low melting point metals are reutilized together with contaminated lead, after subjected to decontamination, generated from facilities such as nuclear power plant or lead for disposal. Since almost all materials including the molten salts and the molten metals can be enclosed, the amount of wastes can be reduced. In addition, radiation exposure of operators who handle them can be reduced. (T.M.)

Preliminary area selection considerations for radioactive waste repositories in bedded salt

International Nuclear Information System (INIS)

Wagoner, J.L.; Steinborn, T.L.

1979-01-01

This guide describes an approach to selection of areas of bedded salt which contain potentially suitable sites for the storage of radioactive waste. To evaluate a site selected by a license applicant, it is necessary to understand the technical site characteristics which should be considered in the preliminary phase of site selection. These site characteristics are presented here in checklist form, and each item is accompanied by a discussion which explains its significance. These qualitative considerations are used first to select an area of interest within a broad geologic or geomorphic region. Once an area has been selected, more quantitative information must be acquired to determine whether the proposed site meets the resultations for storage of nuclear waste

Accumulated energy determination in salts rocks irradiated by means of thermoluminescence techniques: application to the high level radioactive wastes repositories analysis

International Nuclear Information System (INIS)

Dies, J.; Ortega. J.; Tarrasa. F.; Cuevas, C.

1995-01-01

The report summarizes the study carried out to develop the radiation effects on salt rocks in order to repository the high level radioactive wastes. The study is structured into 3 main aspects: 1.- Analysis of irradiation experiences in Haw project of Pet ten reactor. 2.- Irradiation of salt sample of CESAR industrial irradiator. 3.- Correlation study between the accumulated energy, termoluminescence answer and the defect concentration

Geochemical processes in marine salt deposits: Their significance and their implications in connection with disposal of radioactive waste within salt domes

Energy Technology Data Exchange (ETDEWEB)

Herrmann, A G [Goettingen Univ. (Germany, F.R.). Geochemisches Inst.

1980-01-01

Attempts to effect permanent disposal of radioactive wastes in marine evaporites should do nothing to disturb, either in the short or the long term, the present relative stability of such bodies of rock. It is necessary to take account of all of the geochemical and physico-chemical reactions known to have been involved in the processes which formed the evaporites before proceeding to an acceptable strategy for disposal of radionucleides. These processes can be represented as three kinds of metamorphism: 1. solution metamorphism, 2. thermal metamorphism, 3. dynamic metamorphism. In all of the evaporite occurrences in Germany such processes have been influential in altering, on occasion significantly, the primary mineralogical composition and have also promoted a considerable degree of transposition of material. Given similar geochemical and physico-chemical premises, these metamorphic processes could become effective now or in the future. It is therefore necessary to discuss the following criteria when examining salt domes as permanent repositories of highly radioactive substances: (1) Temperatures <= 90/sup 0/ +- 10/sup 0/C at the contact between waste containers and rock salt; (2) Temperatures <= 75/sup 0/C within zones of carnallite rocks; (3) Immobilisation of high-level waste in crystalline forms whenever possible; (4) Systems of additional safety barriers around the waste containers or the unreprocessed spent fuel elements. The geochemical and physical effectiveness of the barriers within an evaporite environment must be guaranteed. For example: Ni-Ti-alloys, corundum, ceramic, anhydrite.

Study of composite adsorbent synthesis and characterization for the removal of Cs in the high-salt and high-radioactive wastewater

Energy Technology Data Exchange (ETDEWEB)

Kim, Jimin; Lee, Keun Young; Kim, Kwang Wook; Lee, Eil Hee; Chung, Dong Yong; Moon, Jei Kwon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Hyun, Jae Hyuk [Chungnam National University, Daejeon (Korea, Republic of)

2017-03-15

For the removal of cesium (Cs) from high radioactive/high salt-laden liquid waste, this study synthesized a highly efficient composite adsorbent (potassium cobalt ferrocyanide (PCFC)-loaded chabazite (CHA)) and evaluated its applicability. The composite adsorbent used CHA, which could accommodate Cs as well as other molecules, as a supporting material and was synthesized by immobilizing the PCFC in the pores of CHA through stepwise impregnation/precipitation with CoCl{sub 2} and K{sub 4}Fe (CN){sub 6} solutions. When CHA, with average particle size of more than 10 μm, is used in synthesizing the composite adsorbent, the PCFC particles were immobilized in a stable form. Also, the physical stability of the composite adsorbent was improved by optimizing the washing methodology to increase the purity of the composite adsorbent during the synthesis. The composite adsorbent obtained from the optimal synthesis showed a high adsorption rate of Cs in both fresh water (salt-free condition) and seawater (high-salt condition), and had a relatively high value of distribution coefficient (larger than 10{sup 4} mL·g{sup -1}) regardless of the salt concentration. Therefore, the composite adsorbent synthesized in this study is an optimized material considering both the high selectivity of PCFC on Cs and the physical stability of CHA. It is proved that this composite adsorbent can remove rapidly Cs contained in high radioactive/high salt-laden liquid waste with high efficiency.

Method for volume reduction and encapsulation of water-bearing, low-level radioactive wastes

International Nuclear Information System (INIS)

1982-01-01

The invention relates to the processing of water-bearing wastes, especially those containing radioactive materials from nuclear power plants like light-water moderated and cooled reactors. The invention provides a method to reduce the volume of wastes like contaminated coolants and to dispose them safely. According to the invention, azeotropic drying is applied to remove the water. Distilation temperatures are chosen to be lower than the lowest boiling point of the mixture components. In the preferred version, a polymerizing monomer is used to obtain the azeotropic mixture. In doing so, encapsulation is possible by combination with a co-reactive polymer that envelopes the waste material. (G.J.P.)

Safety assessment of radioactive waste disposal into geological formations; a preliminary application of fault tree analysis to salt deposits

International Nuclear Information System (INIS)

Bertozzi, B.; D'Alessandro, M.; Girardi, F.; Vanossi, M.

1978-01-01

The methodology of the fault tree analysis (FTA) has been widely used at the Joint Research Centre of Ispra in nuclear reactor safety studies. The aim of the present work consisted in studying the applicability of this methodology to geological repositories of radioactive wastes, including criteria and approaches for the quantification of probalities of primary events. The present work has just an illustrative purpose. Two ideal cases of saline formations, I.E. a bedded salt and a diapir were chosen as potential disposal sites for radioactive waste. On the basis of arbitrarily assumed hydrogeological features of the salt formations and their surrounding environment, possible phenomena capable of causing the waste to be released from each formation have been discussed and gathered following the logical schemes of the FTA. The assessment of probability values for release events due to natural causes as well as to human actions, over different time periods, up to one million years, has been discussed

Salt supply to and significance of asymmetric salt diapirs

DEFF Research Database (Denmark)

Koyi, H.; Burliga, S.; Chemia, Zurab

2012-01-01

Salt diapirs can be asymmetric both internally and externally reflecting their evolution history. As such, this asymmetry bear a significant amount of information about the differential loading (± lateral forces) and in turn the salt supply that have shaped the diapir. In two dimensions......, In this study we compare results of analogue and numerical models of diapirs with two natural salt diapris (Klodawa and Gorleben diapirs) to explain their salt supply and asymmetric evolution. In a NW-SE section, the Gorleben salt diapir possesses an asymmetric external geometry represented by a large...... southeastern overhang due to salt extrusion during Middle Cretaceous followed by its burial in Tertiary. This external asymmetry is also reflected in the internal configuration of the diapir which shows different rates of salt flow on the two halves of the structure. The asymmetric external and internal...

Generic aspects of salt repositories

International Nuclear Information System (INIS)

Laughon, R.B.

1979-01-01

The history of geological disposal of radioactive wastes in salt is presented from 1957 when a panel of the National Academy of Sciences-National Research Council recommended burial in bedded salt deposits. Early work began in the Kansas, portion of the Permian Basin where simulated wastes were placed in an abandoned salt mine at Lyons, Kansas, in the late 1960's. This project was terminated when the potential effect of nearby solution mining activities could not be resolved. Evaluation of bedded salts resumed a few years later in the Permian Basin in southeastern New Mexico, and search for suitable sites in the 1970's resulted in the formation of the National Waste Terminal Storage Program in 1976. Evaluation of salt deposits in many regions of the United States has been virtually completed and has shown that deposits having the greatest potential for radioactive waste disposal are those of the largest depositional basins and salt domes of the Gulf Coast region

Radioactive Waste Isolation in Salt: Peer review of documents dealing with geophysical investigations

International Nuclear Information System (INIS)

McGinnis, L.D.; Bowen, R.H.

1987-03-01

The Salt Repository Project, a US Department of Energy program to develop a mined repository in salt for high-level radioactive waste, is governed by a complex and sometimes inconsistent array of laws, administrative regulations, guidelines, and position papers. In conducting multidisciplinary peer reviews of contractor documents in support of this project, Argonne National Laboratory has needed to inform its expert reviewers of these governmental mandates, with particular emphasis on the relationship between issues and the technical work undertaken. This report acquaints peer review panelists with the regulatory framework as it affects their reviews of site characterization plans and related documents, including surface-based and underground test plans. Panelists will be asked to consider repository performance objectives and issues as they judge the adequacy of proposed geophysical testing. All site-specific discussions relate to the Deaf Smith County site in Texas, which was approved for site characterization by the President in May 1986. Natural processes active at the Deaf Smith County site and the status of geophysical testing near the site are reviewed briefly. 25 refs., 4 figs., 5 tabs

Technical bases for establishing a salt test facility

International Nuclear Information System (INIS)

1985-05-01

The need for a testing facility in which radioactive materials may be used in an underground salt environment is explored. No such facility is currently available in salt deposits in the United States. A salt test facility (STF) would demonstrate the feasibility of safely storing radioactive waste in salt and would provide data needed to support the design, construction, licensing, and operation of a radioactive waste repository in salt. Nineteen issues that could affect long-term isolation of waste materials in a salt repository are identified from the most pertinent recent literature. The issues are assigned an overall priority and a priority relative to the activities of the STF. Individual tests recommended for performance in the STF to resolve the 19 issues are described and organized under three groups: waste package performance, repository design and operation, and site characterization and evaluation. The requirements for a salt test facility are given in the form of functional criteria, and the approach that will be used in the design, execution, interpretation, and reporting of tests is discussed

Geosphere migration studies as support for the comparison of candidate sites for disposal of radioactive waste in rock-salt

International Nuclear Information System (INIS)

Glasbergen, P.; Hassanizadeh, S.M.; Noordijk, H.; Sauter, F.

1988-01-01

The Dutch research program on the geological disposal of radioactive waste was designed to supply a basis for the selection of combinations of three factors, i.e., type of rock-salt formation, site, and disposal technique, satisfying radiological standards and other criteria for final disposal. The potential sites have been grouped according to the type of rock-salt formation (e.g. bedded salt and salt domes) and two classes of depth below the surface of the ground. Values for geohydrological parameters were obtained by extrapolation of data from existing boreholes and analysis of the sedimentary environment. A three-dimensional model of groundwater flow and contaminant transport, called METROPOL, has been developed. To investigate the effect of high salinity on nuclide transport properly, a theoretical experimental study was carried out. Use of a thermodynamic approach showed that terms related to salt mass fraction have to be added to Darcy's and Fick's laws. An experimental study to investigate effects of these modifications is in progress. 8 refs.; 8 figs.; 1 table

Technical support for GEIS: radioactive waste isolation in geologic formations. Volume 16. Repository preconceptual design studies: BPNL waste forms in salt

International Nuclear Information System (INIS)

1978-04-01

This volume, Volume 16, ''Repository Preconceptual Design Studies: BPNL Waste Forms in Salt,'' is one of a 23 volume series, ''Technical Support for GEIS: Radioactive Waste Isolation in Geologic Formations,'' Y/OWI/TM-36, which supplements the ''Contribution to Draft Generic Environmental Impact Statement on Commercial Waste Management: Radioactive Waste Isolation in Geologic Formations,'' Y/OWI/TM-44. The series provide a more complete technical basis for the preconceptual designs, resource requirements, and environmental source terms associated with isolating commercial LWR wastes in underground repositories in salt, granite, shale and basalt. Wastes are considered from three fuel cycles: uranium and plutonium recycling, no recycling of spent fuel and uranium-only recycling. This document describes a preconceptual design for a nuclear waste storage facility in salt. The waste forms assumed to arrive at the repository were supplied by Battelle Pacific Northwest Laboratories (BPNL). The facility design consists of several chambers excavated deep within a geologic formation together with access shafts and supportive surface structures. The facility design provides for: receiving and unloading waste containers; lowering them down shafts to the mine level; transporting them to the proper storage area and emplacing them in mined storage rooms. Drawings of the facility design are contained in TM-36/17, ''Drawings for Repository Preconceptual Design Studies: BPNL Waste Forms in Salt.''

Disposition of the fluoride fuel and flush salts from the Molten Salt Reactor experiment at Oak Ridge National Laboratory

International Nuclear Information System (INIS)

Peretz, F.J.

1996-01-01

The Molten Salt Reactor Experiment (MSRE) is an 8 MW reactor that was operated at Oak Ridge National Laboratory (ORNL) from 1965 through 1969. The reactor used a unique liquid salt fuel, composed of a mixture of LIF, BeF 2 , ZrF 4 , and UF 4 , and operated at temperatures above 600 degrees C. The primary fuel salt circulation system consisted of the reactor vessel, a single fuel salt pump, and a single primary heat exchanger. Heat was transferred from the fuel salt to a coolant salt circuit in the primary heat exchanger. The coolant salt was similar to the fuel salt, except that it contains only LiF (66%) and BeF, (34%). The coolant salt passed from the primary heat exchanger to an air-cooled radiator and a coolant salt pump, and then returned to the primary heat exchanger. Each of the salt loops was provided with drain tanks, located such that the salt could be drained out of either circuit by gravity. A single drain tank was provided for the non-radioactive coolant salt. Two drain tanks were provided for the fuel salt. Since the fuel salt contained radioactive fuel, fission products, and activation products, and since the reactor was designed such that the fuel salt could be drained immediately into the drain tanks in the event of a problem in the fuel salt loop, the fuel salt drain tanks were provided with a system to remove the heat generated by radioactive decay. A third drain tank connected to the fuel salt loop was provided for a batch of flush salt. This batch of salt, similar in composition to the coolant salt, was used to condition the fuel salt loop after it had been exposed to air and to flush the fuel salt loop of residual fuel salt prior to accessing the reactor circuit for maintenance or experimental activities. This report discusses the disposition of the fluoride fuel and flush salt

Saltstone: cement-based waste form for disposal of Savannah River Plant low-level radioactive salt waste

International Nuclear Information System (INIS)

Langton, C.A.

1984-01-01

Defense waste processing at the Savannah River Plant will include decontamination and disposal of approximately 400 million liters of waste containing NaNO 3 , NaOH, Na 2 SO 4 , and NaNO 2 . After decontamination, the salt solution is classified as low-level waste. A cement-based waste form, saltstone, has been designed for disposal of Savannah River Plant low-level radioactive salt waste. Bulk properties of this material have been tailored with respect to salt leach rate, permeability, and compressive strength. Microstructure and mineralogy of leached and unleached specimens were characterized by SEM and x-ray diffraction analyses. The disposal system for the DWPF salt waste includes reconstitution of the crystallized salt as a solution containing 32 wt % solids. This solution will be decontaminated to remove 137 Cs and 90 Sr and then stabilized in a cement-based waste form. Laboratory and field tests indicate that this stabilization process greatly reduces the mobility of all of the waste constitutents in the surface and near-surface environment. Engineered trenches for subsurface burial of the saltstone have been designed to ensure compatibility between the waste form and the environment. The total disposal sytem, saltstone-trench-surrounding soil, has been designed to contain radionuclides, Cr, and Hg by both physical encapsulation and chemical fixation mechanisms. Physical encapsulation of the salts is the mechanism employed for controlling N and OH releases. In this way, final disposal of the SRP low-level waste can be achieved and the quality of the groundwater at the perimeter of the disposal site meets EPA drinking water standards

Basic reasons and the practice of using deep water-bearing levels for liquid radioactive waste disposal

International Nuclear Information System (INIS)

Spitsyn, V.I.; Pimenov, M.K.; Balukova, V.D.; Leontichuk, A.S.; Kokorin, I.N.; Yudin, F.P.; Rakov, N.A.

1978-01-01

Speculations are presented on the development and organization of liquid radioactive waste underground disposal in deep water-bearing levels completely isolated from other levels and the surface. Major requirements are formulated that are laid down to low-, moderate-and high-radioactive wastes subject to the disposal. Geological and hydrological conditions as well as the scheme and design features of pilot field facilities are described, where works on high-active waste disposal were started in 1972. In 1972 and 1973 450 and 1050 m 3 of the wastes (7.5 and 53 MCi) respecrespectively were disposed. The first results of the pilot disposal and the 3-year surveillance over the plate-collector condition and the performance of the facilities have reaffirmed the feasibility, medical and radiation safety and economic attractiveness of the disposal of wastes with up to 10-25 Ci/l specific activity

Waste treatment using molten salt oxidation

International Nuclear Information System (INIS)

Navratil, J.D.; Stewart, A.E.

1996-01-01

MSO technology can be characterized as a submerged oxidation process; the basic concept is to introduce air and wastes into a bed of molten salt, oxidize the organic wastes in the molten salt, use the heat of oxidation to keep the salt molten and remove the salt for disposal or processing and recycling. The molten salt (usually sodium carbonate at 900-1000 C) provides four waste management functions: providing a heat transfer medium, catalyzing the oxidation reaction, preventing the formation of acid gases by forming stable salts, and efficiently capturing ash particles and radioactive materials by the combined effects of wetting, encapsulation and dissolution. The MSO process requires no wet scrubbing system for off-gas treatment. The process has been developed through bench-scale and pilot-scale testing, with successful destruction demonstration of a wide variety of hazardous and mixed (radioactive and hazardous wastes). (author). 24 refs, 2 tabs, 2 figs

Testing of Air Pulse Agitators to Support Design of Savannah River Site Highly Radioactive Processing at the Salt Waste Processing Facility

International Nuclear Information System (INIS)

Gallego, R.M.; Stephens, A.B.; Wilkinson, R.H.; Dev, H.; Suggs, P.C.

2006-01-01

The Salt Waste Processing Facility (SWPF) is intended to concentrate the highly radioactive constituents from waste salt solutions at the Savannah River Site (SRS). Air Pulse Agitators (APAs) were selected for process mixing in high-radiation locations at the SWPF. This technology has the advantage of no moving parts within the hot cell, eliminating potential failure modes and the need for maintenance within the high-radiation environment. This paper describes the results of APA tests performed to gain operational and performance data for the SWPF design. (authors)

Underground disposal of radioactive wastes

Energy Technology Data Exchange (ETDEWEB)

NONE

1979-08-15

Disposal of low- and intermediate-level radioactive wastes by shallow land burial, emplacement in suitable abandoned mines, or by deep well injection and hydraulic fracturing has been practised in various countries for many years. In recent years considerable efforts have been devoted in most countries that have nuclear power programmes to developing and evaluating appropriate disposal systems for high-level and transuranium-bearing waste, and to studying the potential for establishing repositories in geological formations underlaying their territories. The symposium, organized jointly by the IAEA and OECD's Nuclear Energy Agency in cooperation with the Geological Survey of Finland, provided an authoritative account of the status of underground disposal programmes throughout the world in 1979. It was evidence of the experience that has been gained and the comprehensive investigations that have been performed to study various options for the underground disposal of radioactive waste since the last IAEA/NEA symposium on this topic (Disposal of Radioactive Waste into the Ground) was held in 1967 in Vienna. The 10 sessions covered the following topics: National programme and general studies, Disposal of solid waste at shallow depth and in rock caverns, underground disposal of liquid waste by deep well injection and hydraulic fracturing, Disposal in salt formations, Disposal in crystalline rocks and argillaceous sediments, Thermal aspects of disposal in deep geological formations, Radionuclide migration studies, Safety assessment and regulatory aspects.

Geochemistry of U-Th- REE bearing minerals, in radioactive pegmatite in Um Swassi-Dara area, north eastern desert, Egypt

International Nuclear Information System (INIS)

Ali, B. H.

2007-01-01

Some of the pegmatites in the north Eastern Desert of Egypt have high radioactive values, between them the studied radioactive pegmatites which are clustered just in the western margin of Um Swassi-Dara hosted monzogranites. In zoned pegmatite the alteration zones locate between quartz core and intermediate zone are characterizing with the abundance of rare-earth minerals, anderbergite, cenosite, Y-allanite and uranium, thorium minerals such as euxenite, ferro-columbite and complex titanium-yetrum oxides (Kobbite). This zone is a result of many alteration processes developed from volatile-rich magmatic fluids and/or hydrothermal solution which evolved from late differentiated magmatic fluid and lead to increase of U, Th, Zr, Nb, Ti and REE bearing minerals. Such a distinctive alkaline mineralization suite, possibly related to an alkali fluid phase, is superimposed on a more normal, less alkaline group of minerals such as cassiterite, chalcopyrite, and galena. Nb-Ta-Ti minerals bearing U and Th, define a sequence of oxide, cyclosilicate and silicate minerals, showing the effect of hydrothermal overprinting with extreme REE enrichment of the fluids. It can be concluded that the studied mineralization took place in three overlapping stages

Bibliography of studies for the Salt Repository Project Office of the Civilian Radioactive Waste Management Program, April 1978-May 1986

International Nuclear Information System (INIS)

1986-10-01

DOE/CH/10140-05 is an annotated bibliography of approved reports that have been produced for the US Department of Energy Salt Repository Project Office of the Civilian Radioactive Waste Management Program since April 1978. This document is intended for use by the US Department of Energy, State and local officials, the US Nuclear Regulatory Commission, contractors to the Office of Nuclear Waste Isolation, concerned citizens, and others who need a comprehensive listing of reports related to a nuclear waste repository in salt. This document consists of a main report listing, appendixes with Work Breakdown Structure lists, and a topical index

Bibliography of studies for the Salt Repository Project Office of the Civilian Radioactive Waste Management Program, April 1978-December 1986

International Nuclear Information System (INIS)

1987-06-01

This document is an annotated bibliography of approved reports that have been produced for the US Department of Energy Salt Repository Project Office of the Civilian Radioactive Waste Management Program since April 1978. This document is intended for use by the US Department of Energy, State and local officials, the US Nuclear Regulatory Commission, contractors to the Office of Nuclear Waste Isolation, concerned citizens, and others who need a comprehensive listing of reports related to a nuclear waste repository in salt. This document consists of a main report listing, appendixes with Work Breakdown Structure lists, and a topical index

Solid waste disposal into salt mines

International Nuclear Information System (INIS)

Repke, W.

1981-01-01

The subject is discussed as follows: general introduction to disposal of radioactive waste; handling of solid nuclear waste; technology of final disposal, with specific reference to salt domes; conditioning of radioactive waste; safety barriers for radioactive waste; practice of final disposal in other countries. (U.K.)

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

Federal Republic of Germany R and D programme: A special issue of the journal radioactive waste management and the nuclear fuel cycle

International Nuclear Information System (INIS)

Merz, E.R.

1986-01-01

This book examines the issues of radioactive waste management and the nuclear fuel cycle in the Federal Republic of Germany. Topics considered include the challenges of waste handling and disposal, the borosilicate glass for Pamela, the treatment and conditioning of transuranelement bearing wastes in the Federal Republic of Germany, conditioning of low and intermediate level wastes, volume reduction of low level solid radioactive waste by incineration and compaction in the Federal Republic of Germany, MAW test emplacement in boreholes, treatment and disposal of special radioactive wastes comprising tritium, carbon 14, krypton 85 and iodine 129, and the German Project: ''Safety Studies for Nuclear Waste Management: Development of Safety Assessment Methodology for Final Disposal of Nuclear Waste in a Salt Dome

Radioactivity in environmental samples

International Nuclear Information System (INIS)

Fornaro, Laura

2001-01-01

The objective of this practical work is to familiarize the student with radioactivity measures in environmental samples. For that were chosen samples a salt of natural potassium, a salt of uranium or torio and a sample of drinkable water

Water-bearing explosive compositions

Energy Technology Data Exchange (ETDEWEB)

Gay, G M

1970-12-21

An explosive water-bearing composition, with high detonation velocity, comprises a mixture of (1) an inorganic oxidizer salt; (2) nitroglycerine; (3) nitrocellulose; (4) water; and (5) a water thickening agent. (11 claims)

Principal aspects of petrographical examination of rock salts to assess their suitability for radioactive waste disposal

International Nuclear Information System (INIS)

Shekhunova, S.B.

1995-01-01

To solve the problem of high-level radioactive waste (HLRW) isolation in Ukraine a preparatory stage of feasibility study as to the construction of a pilot laboratory has been completed. Salty formations are considered as possible host rocks for HLRW isolation. 7 salt formations located in 5 regions of Ukraine have been examined and was found that only two, i.e. the Upper Devonian and Lower Permian halogenic formations of the Dnieper-Donets Depression appeared to have considerable promise for these purposes. In these two formations 4 zones with 12 candidate-sites were selected. The promising zones are located both in bedded salt and in salt domes. Analytical treatment our previous studies as well as a special-purpose research have resulted in designing packages of the schematic information models for the zones and some candidate-sites. Now we are preparing to start exploration drilling at several promising structures. Research has been carried out by the Institute of Geological Sciences (National Academy of Sciences of Ukraine) on budget and contract financial basis with the participation of branch institutes and the State Committee on Nuclear power Utilization (Goskomatom). The drilling and geophysical data were presented by Goskomgeologiya production organizations

Field experiments in salt formations

International Nuclear Information System (INIS)

Kuehn, K.

1986-01-01

Field experiments in salt formations started as early as 1965 with Project Salt Vault in the Lyons Mine, Kansas, U.S.A., and with the purchase of the Asse salt mine by the German Federal Government. Underground tests concentrated on the heat dissipation around buried high-level radioactive wastes and the geomechanical consequences of their disposal. Near-field investigations cover the properties of water and gas release, radiolysis and corrosion. Further objectives of field experiments are the development and underground testing of a handling system for high-level wastes. The performance of an underground test disposal for such wastes is not only considered to be necessary for technical and scientific reasons but also for improving public acceptance of the concept of radioactive waste disposal. (author)

Methods and results of the investigation of the thermomechanical behaviour of rock salt with regard to the final disposal of high-level radioactive wastes

International Nuclear Information System (INIS)

Wieczorek, K.; Klarr, K.

1993-01-01

This report summarizes the knowledge about thermal and mechanical behaviour of rock salt that has been accumulated by various R and D institutions in Germany from laboratory and in situ investigations. An important objective is to give a comprehensive overview of the investigation methods and instruments available and to discuss these methods and instruments with regard to their applicability and reliability for the investigation of the thermomechanical effects of high level radioactive waste emplacement in rock salt formations. The report is focused on the activities of the GSF-Institut fur Tieflagerung in the Asse mine regarding the disposal of high and intermediate level radioactive waste during the last decades. The design and the results of the most important in situ experiments are presented and discussed in detail. The results are compared to model calculations in order to evaluate the reliability of both the measurements and the calculation results. The relevance of the results for the situation in Spain is discussed in a separate chapter. As the investigations in Germany have been performed in domal salt, while the Spanish concept is based on waste disposal in bedded salt, significant differences in the thermomechanical behaviour cannot be excluded. The investigation methods, however, will be applicable. (Author)

Sintered bentonite ceramics for the immobilization of cesium- and strontium-bearing radioactive waste

Science.gov (United States)

Ortega, Luis Humberto

The Advanced Fuel Cycle Initiative (AFCI) is a Department of Energy (DOE) program, that has been investigating technologies to improve fuel cycle sustainability and proliferation resistance. One of the program's goals is to reduce the amount of radioactive waste requiring repository disposal. Cesium and strontium are two primary heat sources during the first 300 years of spent nuclear fuel's decay, specifically isotopes Cs-137 and Sr-90. Removal of these isotopes from spent nuclear fuel will reduce the activity of the bulk spent fuel, reducing the heat given off by the waste. Once the cesium and strontium are separated from the bulk of the spent nuclear fuel, the isotopes must be immobilized. This study is focused on a method to immobilize a cesium- and strontium-bearing radioactive liquid waste stream. While there are various schemes to remove these isotopes from spent fuel, this study has focused on a nitric acid based liquid waste. The waste liquid was mixed with the bentonite, dried then sintered. To be effective sintering temperatures from 1100 to 1200°C were required, and waste concentrations must be at least 25 wt%. The product is a leach resistant ceramic solid with the waste elements embedded within alumino-silicates and a silicon rich phase. The cesium is primarily incorporated into pollucite and the strontium into a monoclinic feldspar. The simulated waste was prepared from nitrate salts of stable ions. These ions were limited to cesium, strontium, barium and rubidium. Barium and rubidium will be co-extracted during separation due to similar chemical properties to cesium and strontium. The waste liquid was added to the bentonite clay incrementally with drying steps between each addition. The dry powder was pressed and then sintered at various temperatures. The maximum loading tested is 32 wt. percent waste, which refers to 13.9 wt. percent cesium, 12.2 wt. percent barium, 4.1 wt. percent strontium, and 2.0 wt. percent rubidium. Lower loadings of waste

Conditions for the test emplacement of intermediate-level radioactive wastes in chamber 8a of the 511 m level of the Asse Salt Mine

International Nuclear Information System (INIS)

1984-01-01

The Gesellschaft fuer Strahlen- und Umweltforschung mbH (GSF) emplaces intermediate-level radioactive wastes which accumulate in an activity involving the use of radioactive materials that is licensed or reported in the Federal Republic of Germany or which are stored on an interim basis by the appropriate licensing or inspection agencies in chamber 8a of the 511 m level of the Asse Salt Mine in Remlingen near Wolfenbuettel in conjunction with an engineering test program. The type and form of the intermediate-level wastes must conform to certain conditions so that there are no hazards to personnel and the repository during transfer and subsequent storage. It is therefore necessary for the radioactive wastes to be treated and packaged before delivery in such a way that they satisfy the conditions presented in this document. The GSF shall inform the companies and organizations delivering wastes about its experiences with emplacement operations. The Conditions for the Test Emplacement of Intermediate-Level Radioactive Wastes in Chamber 8a of the 511 m Level of the Asse Salt Mine must be adapted to conform to the latest state of science and the art. The GSF must therefore reserve the right to modify the conditions, allowing for an appropriate transition period

Selenium in the Blackfoot, Salt, and Bear River Watersheds

Science.gov (United States)

Hamilton, S.J.; Buhl, K.J.

2005-01-01

Nine stream sites in the Blackfoot River, Salt River, and Bear River watersheds in southeast Idaho, USA were sampled in May 2001 for water, surficial sediment, aquatic plants, aquatic invertebrates, and fish. Selenium was measured in these aquatic ecosystem components, and a hazard assessment was performed on the data. Water quality characteristics such as pH, hardness, and specific conductance were relatively uniform among the nine sites. Of the aquatic components assessed, water was the least contaminated with selenium because measured concentrations were below the national water quality criterion of 5 μ g/L at eight of the nine sites. In contrast, selenium was elevated in sediment, aquatic plants, aquatic invertebrates, and fish from several sites, suggesting deposition in sediments and food web cycling through plants and invertebrates. Selenium was elevated to concentrations of concern in fish at eight sites (> 4 μ g/g in whole body). A hazard assessment of selenium in the aquatic environment suggested a moderate hazard at upper Angus Creek (UAC) and Smoky Creek (SC), and high hazard at Little Blackfoot River (LiB), Blackfoot River gaging station (BGS), State Land Creek (SLC), upper (UGC) and lower Georgetown Creek (LGC), Deer Creek (DC), and Crow Creek (CC). The results of this study indicate that selenium concentrations from the phosphate mining area of southeast Idaho were sufficiently elevated in several ecosystem components to cause adverse effects to aquatic resources in southeastern Idaho.

Hydrometallurgical treatment of plutonium. Bearing salt baths waste

International Nuclear Information System (INIS)

Bros, P.; Gozlan, J.P.; Lecomte, M.; Bourges, J.

1993-01-01

The salt flux issuing from the electrorefining of plutonium metal alloy in salt baths (KCI + NaCI) poses a difficult problem of the back-end alpha waste management. An alternative to the salt process promoted by Los Alamos Laboratory is to develop a hydrometallurgical treatment. A new process based on the electrochemistry technique in aqueous solution has been defined and tested successfully in the CEA. The diagram of the process exhibits two principal steps: in the head-end, a dissolution in HNO 3 medium accompanied with an electrolytic dechlorination leading to a quantitative elimination of chloride as CI 2 gas followed by its trapping one soda lime cartridge, a complete oxidative dissolution of the refractory Pu residues by electrogenerated Ag(II), in the back-end: the Pu and Am recoveries by chromatographic extractions. (authors). 10 figs., 9 refs

Effect of using FLiBe and FLiNaBe molten salts bearing plutonium fluorides on the neutronic performance of PACER

International Nuclear Information System (INIS)

Acir, Adem

2012-01-01

In this paper, the effects of using FLiBe and FLiNaBe Molten Salts Bearing Plutonium Fluorides on the neutronic performance of the PACER are investigated. The optimum radial thickness for tritium self-sufficiency of the blankets addition of plutonium fluorides to FLiNaBe (LiF-/NaF BeF 2 ) and FLiBe (LiF-/BeF 2 ) of a dual purpose modified PACER concept are determined. The calculations are carried out with the one dimensional transport code XSDRNPM/SCALE5. The tritium breeding capacities of FLiNaBe and FLiBe with addition of plutonium fluorides in molten salt zone are investigated and compared. The optimum molten salt zone thickness is computed as 155 cm for tritium self-sufficiency of the blankets using FLiBe +1% PuF 4 whereas, the optimum thickness with FLiNaBe +1% PuF 4 is calculated as 170 cm. In addition, neutron transport calculations have been performed to evaluate the energy multiplication factor, total fission rate, displacement per atom and helium gas generation for optimal radial thickness in the blanket. Also, the tritium production and the radiation damage limits should be evaluated together in a fusion blanket for determining the optimum thickness of molten salt layer. (orig.)

Underground storage of radioactive wastes

International Nuclear Information System (INIS)

Dietz, D.N.

1977-01-01

An introductory survey of the underground disposal of radioactive wastes is given. Attention is paid to various types of radioactive wastes varying from low to highly active materials, as well as mining techniques and salt deposits

Possible salt mine and brined cavity sites for radioactive waste disposal in the northeastern southern peninsula of Michigan

International Nuclear Information System (INIS)

Landes, K.K.; Bourne, H.L.

1976-01-01

A reconnaissance report on the possibilities for disposal of radioactive waste covers Michigan only, and is more detailed than an earlier one involving the northeastern states. Revised ''ground rules'' for pinpointing both mine and dissolved salt cavern sites for waste disposal include environmental, geologic, and economic factors. The Michigan basin is a structural bowl of Paleozoic sediments resting on downwarped Precambrian rocks. The center of the bowl is in Clare and Gladwin Counties, a short distance north of the middle of the Southern Peninsula. The strata dip toward this central area, and some stratigraphic sequences, including especially the salt-containing Silurian section, increase considerably in thickness in that direction. Lesser amounts of salt are also present in the north central part of the Lower Peninsula. Michigan has been an oil and gas producing state since 1925 and widespread exploration has had two effects on the selection of waste disposal sites: (1) large areas are leased for oil and gas; and (2) the borehole concentrations, whether producing wells, dry holes, or industrial brine wells that penetrated the salt section, should be avoided. Two types of nuclear waste, low level and high level, can be stored in man-made openings in salt beds. The storage facilities are created by (1) the development of salt mines where the depths are less than 3000 ft, and (2) cavities produced by pumping water into a salt bed, and bringing brine back out. The high level waste disposal must be confined to mines of limited depth, but the low level wastes can be accommodated in brine cavities at any depth. Seven potential prospects have been investigated and are described in detail

Waterproofing improvement of radioactive waste asphalt solid

International Nuclear Information System (INIS)

Adachi, Katsuhiko; Yamaguchi, Takashi; Ikeoka, Akira.

1981-01-01

Purpose: To improve the waterproofing of asphalt solid by adding an alkaline earth metal salt and, further, paraffin, into radioactive liquid waste when processing asphalt solidification of the radioactive liquid waste. Method: Before processing molten asphalt solidification of radioactive liquid waste, soluble salts of alkaline earth metal such as calcium chloride, magnesium chloride, or the like is added to the radioactive liquid waste. Paraffin having a melting point of higher than 60 0 C, for example, is added to the asphalt, and waterproofing can be remarkably improved. The waste asphalt solid thus fabricated can prevent the swelling thereof, and can improve its waterproofing. (Yoshihara, H.)

Processes and parameters involved in modeling radionuclide transport from bedded salt repositories. Final report. Technical memorandum

International Nuclear Information System (INIS)

Evenson, D.E.; Prickett, T.A.; Showalter, P.A.

1979-07-01

The parameters necessary to model radionuclide transport in salt beds are identified and described. A proposed plan for disposal of the radioactive wastes generated by nuclear power plants is to store waste canisters in repository sites contained in stable salt formations approximately 600 meters below the ground surface. Among the principal radioactive wastes contained in these canisters will be radioactive isotopes of neptunium, americium, uranium, and plutonium along with many highly radioactive fission products. A concern with this form of waste disposal is the possibility of ground-water flow occurring in the salt beds and endangering water supplies and the public health. Specifically, the research investigated the processes involved in the movement of radioactive wastes from the repository site by groundwater flow. Since the radioactive waste canisters also generate heat, temperature is an important factor. Among the processes affecting movement of radioactive wastes from a repository site in a salt bed are thermal conduction, groundwater movement, ion exchange, radioactive decay, dissolution and precipitation of salt, dispersion and diffusion, adsorption, and thermomigration. In addition, structural changes in the salt beds as a result of temperature changes are important. Based upon the half-lives of the radioactive wastes, he period of concern is on the order of a million years. As a result, major geologic phenomena that could affect both the salt bed and groundwater flow in the salt beds was considered. These phenomena include items such as volcanism, faulting, erosion, glaciation, and the impact of meteorites. CDM reviewed all of the critical processes involved in regional groundwater movement of radioactive wastes and identified and described the parameters that must be included to mathematically model their behavior. In addition, CDM briefly reviewed available echniques to measure these parameters

Reconsolidated Salt as a Geotechnical Barrier

International Nuclear Information System (INIS)

Hansen, Francis D.; Gadbury, Casey

2015-01-01

Salt as a geologic medium has several attributes favorable to long-term isolation of waste placed in mined openings. Salt formations are largely impermeable and induced fractures heal as stress returns to equilibrium. Permanent isolation also depends upon the ability to construct geotechnical barriers that achieve nearly the same high-performance characteristics attributed to the native salt formation. Salt repository seal concepts often include elements of reconstituted granular salt. As a specific case in point, the Waste Isolation Pilot Plant recently received regulatory approval to change the disposal panel closure design from an engineered barrier constructed of a salt-based concrete to one that employs simple run-of-mine salt and temporary bulkheads for isolation from ventilation. The Waste Isolation Pilot Plant is a radioactive waste disposal repository for defense-related transuranic elements mined from the Permian evaporite salt beds in southeast New Mexico. Its approved shaft seal design incorporates barrier components comprising salt-based concrete, bentonite, and substantial depths of crushed salt compacted to enhance reconsolidation. This paper will focus on crushed salt behavior when applied as drift closures to isolate disposal rooms during operations. Scientific aspects of salt reconsolidation have been studied extensively. The technical basis for geotechnical barrier performance has been strengthened by recent experimental findings and analogue comparisons. The panel closure change was accompanied by recognition that granular salt will return to a physical state similar to the halite surrounding it. Use of run-of-mine salt ensures physical and chemical compatibility with the repository environment and simplifies ongoing disposal operations. Our current knowledge and expected outcome of research can be assimilated with lessons learned to put forward designs and operational concepts for the next generation of salt repositories. Mined salt

Reconsolidated Salt as a Geotechnical Barrier

Energy Technology Data Exchange (ETDEWEB)

Hansen, Francis D. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Gadbury, Casey [USDOE Carlsbad Field Office, NM (United States)

2015-11-01

Salt as a geologic medium has several attributes favorable to long-term isolation of waste placed in mined openings. Salt formations are largely impermeable and induced fractures heal as stress returns to equilibrium. Permanent isolation also depends upon the ability to construct geotechnical barriers that achieve nearly the same high-performance characteristics attributed to the native salt formation. Salt repository seal concepts often include elements of reconstituted granular salt. As a specific case in point, the Waste Isolation Pilot Plant recently received regulatory approval to change the disposal panel closure design from an engineered barrier constructed of a salt-based concrete to one that employs simple run-of-mine salt and temporary bulkheads for isolation from ventilation. The Waste Isolation Pilot Plant is a radioactive waste disposal repository for defense-related transuranic elements mined from the Permian evaporite salt beds in southeast New Mexico. Its approved shaft seal design incorporates barrier components comprising salt-based concrete, bentonite, and substantial depths of crushed salt compacted to enhance reconsolidation. This paper will focus on crushed salt behavior when applied as drift closures to isolate disposal rooms during operations. Scientific aspects of salt reconsolidation have been studied extensively. The technical basis for geotechnical barrier performance has been strengthened by recent experimental findings and analogue comparisons. The panel closure change was accompanied by recognition that granular salt will return to a physical state similar to the halite surrounding it. Use of run-of-mine salt ensures physical and chemical compatibility with the repository environment and simplifies ongoing disposal operations. Our current knowledge and expected outcome of research can be assimilated with lessons learned to put forward designs and operational concepts for the next generation of salt repositories. Mined salt

Devoluming method of acidic radioactive liquid waste and processing system therefor

International Nuclear Information System (INIS)

Shirai, Takamori; Honda, Tadahiro

1998-01-01

Radioactive liquid wastes such as liquid wastes discharged from chemical decontamination (containing free acids, metal salts dissolved in acids, not-dissolved iron rust and radioactive metals) are introduced to an acid recovering device using a diffusion permeation membrane and separated to a deacidified liquid and separated acid liquid. The separated acid liquid mainly comprising free acids is recovered to a tank for recovered acids, and used repeatedly for removing crud. The deacidified liquid mainly comprising salts is concentrated in a reverse osmosis membrane (RO) concentration device. RO concentrated liquid containing radioactive metals is dried, and salts are decomposed in a drying/salt-decomposing device and separated into metal oxides and a mixed gas of an acidic gas and steams. The gas is cooled in an acid absorbing device and recovered as free acids. The metal oxides containing radioactive metals are solidified. (I.N.)

Method of solidifying radioactive wastes

International Nuclear Information System (INIS)

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

1984-01-01

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

Accelerator molten-salt breeder reactor

International Nuclear Information System (INIS)

Furukawa, Kazuo; Kuroi, Hideo; Kato, Yoshio; Oomichi, Toshihiko.

1979-01-01

Purpose: To obtain fission products and to transmute transuranium elements and other radioactive wastes by the use of Accelerator Molten-Salt Breeder Reactor. Constitution: Beams from an accelerator pipe at one end of a target vessel is injected through a window into target molten salts filled inside of the target vessel. The target molten salts are subjected to pump recycling or spontaneous convection while forcively cooled by blanket molten salts in an outer vessel. Then, energy is recovered from the blanket molten salts or the target molten salts at high temperatures through electric power generation or the like. Those salts containing such as thorium 232 and uranium 238 are used as the blanket molten salts so that fission products may be produced by neutrons generated in the target molten salts. PbCl 2 -PbF 2 and LiF-BeF 2 -ThF 4 can be used as the target molten salts and as the blanket molten salts respectively. (Seki, T.)

Geology and undiscovered resource assessment of the potash-bearing Pripyat and Dnieper-Donets Basins, Belarus and Ukraine

Science.gov (United States)

Cocker, Mark D.; Orris, Greta J.; Dunlap, Pamela; Lipin, Bruce R.; Ludington, Steve; Ryan, Robert J.; Słowakiewicz, Mirosław; Spanski, Gregory T.; Wynn, Jeff; Yang, Chao

2017-08-03

Undiscovered potash resources in the Pripyat Basin, Belarus, and Dnieper-Donets Basin, Ukraine, were assessed as part of a global mineral resource assessment led by the U.S. Geological Survey (USGS). The Pripyat Basin (in Belarus) and the Dnieper-Donets Basin (in Ukraine and southern Belarus) host stratabound and halokinetic Upper Devonian (Frasnian and Famennian) and Permian (Cisuralian) potash-bearing salt. The evaporite basins formed in the Donbass-Pripyat Rift, a Neoproterozoic continental rift structure that was reactivated during the Late Devonian and was flooded by seawater. Though the rift was divided, in part by volcanic deposits, into the separate Pripyat and Dnieper-Donets Basins, both basins contain similar potash‑bearing evaporite sequences. An Early Permian (Cisuralian) sag basin formed over the rift structure and was also inundated by seawater resulting in another sequence of evaporite deposition. Halokinetic activity initiated by basement faulting during the Devonian continued at least into the Permian and influenced potash salt deposition and structural evolution of potash-bearing salt in both basins.Within these basins, four areas (permissive tracts) that permit the presence of undiscovered potash deposits were defined by using geological criteria. Three tracts are permissive for stratabound potash-bearing deposits and include Famennian (Upper Devonian) salt in the Pripyat Basin, and Famennian and Cisuralian (lower Permian) salt in the Dnieper-Donets Basin. In addition, a tract was delineated for halokinetic potash-bearing Famennian salt in the Dnieper-Donets Basin.The Pripyat Basin is the third largest source of potash in the world, producing 6.4 million metric tons of potassium chloride (KCl) (the equivalent of about 4.0 million metric tons of potassium oxide or K2O) in 2012. Potash production began in 1963 in the Starobin #1 mine, near the town of Starobin, Belarus, in the northwestern corner of the basin. Potash is currently produced from

In situ investigations on the impact of heat production and gamma radiation with regard to high-level radioactive waste disposal in rock salt formations

International Nuclear Information System (INIS)

Rothfuchs, T.

1986-01-01

Deep geological formations especially rock salt formations, are considered worldwide as suitable media for the final disposal of radioactive high-level waste (HLW). In the Federal Republic of Germany, the Institut fur Tieflagerung of the Gesellschaft fur Strahlen- und Umweltforschung mbH Munchen operates the Asse Salt Mine as a pilot facility for testing the behavior of an underground nuclear waste repository. The tests are performed using heat and radiation sources to simulate disposed HLW canisters. The measured data obtained since 1965 show that the thermomechanical response of the salt formation and the physical/chemical changes in the vicinity of disposal boreholes are not a serious concern and that their long-term consequences can be estimated based on theoretical considerations and in-situ investigations

Molten salt oxidation of mixed wastes: Separation of radioactive materials and Resource Conservation and Recovery Act (RCRA) materials

International Nuclear Information System (INIS)

Bell, J.T.; Haas, P.A.; Rudolph, J.C.

1993-01-01

The Oak Ridge National Laboratory (ORNL) is involved in a program to apply a molten salt oxidation (MSO) process to the treatment of mixed wastes at Oak Ridge and other Department of Energy (DOE) sites. Mixed wastes are defined as those wastes that contain both radioactive components, which are regulated by the atomic energy legislation, and hazardous waste components, which are regulated under the Resource Conservation and Recovery Act (RCRA). A major part of our ORNL program involves the development of separation technologies that are necessary for the complete treatment of mixed wastes. The residues from the MSO treatment of the mixed wastes must be processed further to separate the radioactive components, to concentrate and recycle residues, or to convert the residues into forms acceptable for final disposal. This paper is a review of the MSO requirements for separation technologies, the information now available, and the concepts for our development studies

Barium borosilicate glass - a potential matrix for immobilization of sulfate bearing high-level radioactive liquid waste

International Nuclear Information System (INIS)

Kaushik, C.P.; Mishra, R.K.; Sengupta, P.; Kumar, Amar; Das, D.; Kale, G.B.; Raj, Kanwar

2006-01-01

Borosilicate glass formulations adopted worldwide for immobilization of high-level radioactive liquid waste (HLW) is not suitable for sulphate bearing HLW, because of its low solubility in such glass. A suitable glass matrix based on barium borosilicate has been developed for immobilization of sulphate bearing HLW. Various compositions based on different glass formulations were made to examine compatibility with waste oxide with around 10 wt% sulfate content. The vitrified waste product obtained from barium borosilicate glass matrix was extensively evaluated for its characteristic properties like homogeneity, chemical durability, glass transition temperature, thermal conductivity, impact strength, etc. using appropriate techniques. Process parameters like melt viscosity and pour temperature were also determined. It is found that SB-44 glass composition (SiO 2 : 30.5 wt%, B 2 O 3 : 20.0 wt%, Na 2 O: 9.5 wt% and BaO: 19.0 wt%) can be safely loaded with 21 wt% waste oxide without any phase separation. The other product qualities of SB-44 waste glass are also found to be on a par with internationally adopted waste glass matrices. This formulation has been successfully implemented in plant scale

Management situation and prospect of radioactive waste

International Nuclear Information System (INIS)

Han, Pil Jun

1985-04-01

This book tell US that management situation and prospect of radioactive waste matter, which includes importance of energy, independence, limitation of fossil fuel energy, density of nuclear energy, strategy of supply of energy resource in Korea, nuclear energy development and radioactive waste matter, summary of management of radioactive waste, statistics of radioactive waste, disposal principle of radioactive waste, management on radioactive waste after using, disposal of Trench, La Marche in French, and Asse salt mine in Germany.

Effect of state of tetraoctylammonium and trioctylpropylammonium salts in extracts on coextraction of micro- and macroelements

International Nuclear Information System (INIS)

Bagreev, V.V.; Kardivarenko, L.M.; Zolotov, Yu.A.

1988-01-01

State effect of halide simple and metal-bearing tetraoctyl- and trioctylpropylammonium salts in benzene and nitrobenzene on inhibition of indium and cobalt trace amounts extraction from HCl solutions by the extractable macrocomponents-gallium and zinc, respectively, is investigated. Dissociation constants (K dis ) for metal-bearing salts and tetraoctylammonium and trioctylpropylammonium bromides in nitrobenzene as well are calculated. It is shown, that inhibition of trace elements extraction is the more higher, the more is the difference between K dis for alkylammonium metal-bearing and simple salts

Formation of Microbial Mats and Salt in Radioactive Paddy Soils in Fukushima, Japan

Directory of Open Access Journals (Sweden)

Kazue Tazaki

2015-12-01

Full Text Available Coastal areas in Minami-soma City, Fukushima, Japan, were seriously damaged by radioactive contamination from the Fukushima Daiichi Nuclear Power Plant (FDNPP accident that caused multiple pollution by tsunami and radionuclide exposure, after the Great East Japan Earthquake, on 11 March 2011. Some areas will remain no-go zones because radiation levels remain high. In Minami-soma, only 26 percent of decontamination work had been finished by the end of July in 2015. Here, we report the characterization of microbial mats and salt found on flooded paddy fields at Karasuzaki, Minami-soma City, Fukushima Prefecture, Japan which have been heavily contaminated by radionuclides, especially by Cs (134Cs, 137Cs, 40K, Sr (89Sr, 90Sr, and 91 or 95Zr even though it is more than 30 km north of the FDNPP. We document the mineralogy, the chemistry, and the micro-morphology, using a combination of micro techniques. The microbial mats were found to consist of diatoms with mineralized halite and gypsum by using X-ray diffraction (XRD. Particular elements concentrated in microbial mats were detected using scanning electron microscopy equipped with energy dispersive spectroscopy (SEM-EDS and X-ray fluorescence (XRF. The objective of this contribution is to illustrate the ability of various diatoms associated with minerals and microorganisms which are capable of absorbing both radionuclides and stable isotopes from polluted paddy soils in extreme conditions. Ge semiconductor analysis of the microbial mats detected 134Cs, 137Cs, and 40K without 131I in 2012 and in 2013. Quantitative analysis associated with the elemental content maps by SEM-EDS indicated the possibility of absorption of radionuclide and stable isotope elements from polluted paddy soils in Fukushima Prefecture. In addition, radionuclides were detected in solar salts made of contaminated sea water collected from the Karasuzaki ocean bath, Minami-soma, Fukushima in 2015, showing high Zr content associated

Bases, Assumptions, and Results of the Flowsheet Calculations for the Decision Phase Salt Disposition Alternatives

Energy Technology Data Exchange (ETDEWEB)

Elder, H.H.

2001-07-11

The HLW salt waste (salt cake and supernate) now stored at the SRS must be treated to remove insoluble sludge solids and reduce the soluble concentration of radioactive cesium radioactive strontium and transuranic contaminants (principally Pu and Np). These treatments will enable the salt solution to be processed for disposal as saltstone, a solid low-level waste.

Characterization of the effects of continuous salt processing on the performance of molten salt fusion breeder blankets

International Nuclear Information System (INIS)

Patterson-Hine, F.A.; Davidson, J.W.; Klein, D.E.; Lee, J.D.

1985-01-01

Several continuous salt processing options are available for use in molten salt fusion breeder blanket designs: fluorination only, fluorination plus reductive extraction, and fluorination, plus reductive extraction, plus metal transfer. The effects of processing on blanket performance have been assessed for these three levels of processing and various equilibrium uranium concentrations in the salt. A one-dimensional model of the blanket was used in the neutronics analysis, which incorporated transport calculations with time-dependent isotope generation and depletion calculations. The method of salt processing was found to have little affect on the level of radioactivity, toxicity, or the thermal behavior of the salt during operation of the reactor. The processing rates necessary to maintain the desired uranium concentrations in the suppressed-fission environment were quite low, which permitted only long-lived species to be removed from the salt. The effects of the processing therefore became apparent only after the radioactivity due to the short-lived species diminished. The effect of the additional processing (reductive extraction and metal transfer) could be seen after approximately 1 year of decay, but were not significant at times closer to shutdown. The reduced radioactivity and corresponding heat deposition were thus of no consequence in accident or maintenance situations. Net fissile production in the Be/MS blanket concept at a fusion power level of 3000 MW at 70% capacity ranged from 5100 kg/year to 5170 kg/year for uranium concentrations of 0.11% and 1.0% 233 U in thorium, respectively, with fluorination-only processing. The addition of processing by reductive extraction resulted in 5125 kg/year for the 0.11% 233 U case and 5225 kg/year for the 1.0% 233 U case

Pressure-induced brine migration in consolidated salt in a repository

International Nuclear Information System (INIS)

Hwang, Y.; Chambre, P.L.; Lee, W.W.L.; Pigford, T.H.

1987-06-01

This report describes a mathematical model for brine migration through intact salt near a radioactive waste package emplaced in salt. Solutions indicate limited movement following ten years emplacement

Bituminization of liquid radioactive wastes. Part 1

International Nuclear Information System (INIS)

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

1991-01-01

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

Natural radioactivity levels in soil samples around the flood affected salt field area, Kelambakkam, Chennai, Tamilnadu, India using gamma ray spectrometry

International Nuclear Information System (INIS)

Rajalakshmi, A.; Chandrasekaran, A.; Thangam, V.; Jananee, B.

2018-01-01

Humans are exposed to natural radiation from external sources, which include radionuclides in the earth and cosmic radiation. Gamma Ray spectroscopic technique was used to assess the natural radioactivity in soils around the flood affected salt field area, Kelambakkam Chennai, Tamilnadu, India. The activity concentration of 238 U, 232 Th, 40 K and absorbed dose rate of soil samples were calculated to assess the radiation hazards in the study area

Summary report on salt dissolution review meeting, March 29--30, 1977

International Nuclear Information System (INIS)

Johnson, K.S.; Brokaw, A.L.; Gilbert, J.F.; Saberian, A.; Snow, R.H.; Walters, R.F.

1977-01-01

It is the unanimous conclusion of the Ad Hoc Committee that radioactive waste can be stored in salt and underground repository sites sufficiently removed from natural and/or man-made dissolution areas so that the waste will not be liberated during its hazardous period at projected rates of future salt dissolution. To ensure long-term isolation of radioactive waste in salt formations, specific recommendations are given for needed research concerning (A) General Principles, (B) Basinal or Regional Studies, and (C) Site-Specific Studies, each stated in sequence of priority

Biosphere transport and radiation dose calculations resulting from radioactive waste stored in deep salt formation (PACOMA-project)

International Nuclear Information System (INIS)

Jong, E.J. de; Koester, H.W.; Vries, W.J. de; Lembrechts, J.F.

1990-03-01

Parts are presented of the results of a safety-assessment study of disposal of medium and low level radioactive waste in salt formations in the Netherlands. The study concerns several disposal concepts for 2 kinds of salt formation, a deep dome and a shallow dome. 7 cases were studied with the same Dutch inventory and 1 with a reference inventory R, in order to compare results with those of other PACOMA participants. The total activity of the reference inventory R is 30 percent lower than the Dutch inventory, but some long living nuclides such as I-129, Np-237 and U-238 have a considerably higher activity. This reference inventor R has been combined with the disposal concept of mined cavities in a shallow salt dome. In each case. the released fraction of stored radio-nuclides moves gradually with water through the geosphere to the bio-sphere where it enters a river. River water is used for sprinkler irrigation and for drinking by man and livestock. The dispersal of the radionuclides into the biosphere is calculated with the BIOS program of the NRPB. Subroutines linked to the program add doses via different pathways to obtain a maximum individual dose, a collective dose and an integrated collective dose. This study presents results of these calculations. (author). 11 refs.; 39 figs.; 111 tabs

Salt ingestion caves.

Directory of Open Access Journals (Sweden)

Lundquist Charles A.

2006-01-01

Full Text Available Large vertebrate herbivores, when they find a salt-bearing layer of rock, say in a cliff face, can produce sizable voids where, overgenerations, they have removed and consumed salty rock. The cavities formed by this natural animal process constitute a uniqueclass of caves that can be called salt ingestion caves. Several examples of such caves are described in various publications. Anexample in Mississippi U.S.A., Rock House Cave, was visited by the authors in 2000. It seems to have been formed by deer orbison. Perhaps the most spectacular example is Kitum Cave in Kenya. This cave has been excavated to a length over 100 metersby elephants. An ancient example is La Cueva del Milodon in Chile, which is reported to have been excavated by the now extinctmilodon, a giant ground sloth. Still other possible examples can be cited. This class of caves deserves a careful definition. First, thecavity in rock should meet the size and other conventions of the locally accepted definition of a cave. Of course this requirement differsin detail from country to country, particularly in the matter of size. The intent is to respect the local conventions. The characteristicthat human entry is possible is judged to be a crucial property of any recognized cave definition. Second, the cavity should besignificantly the result of vertebrate animal consumption of salt-bearing rock. The defining process is that rock removed to form thecave is carried away in the digestive track of an animal. While sodium salts are expected to be the norm, other salts for which thereis animal hunger are acceptable. Also some other speleogenesis process, such as solution, should not be excluded as long as it issecondary in formation of a cave in question.

Radioactive waste processing

International Nuclear Information System (INIS)

Dejonghe, P.

1978-01-01

This article gives an outline of the present situation, from a Belgian standpoint, in the field of the radioactive wastes processing. It estimates the annual quantity of various radioactive waste produced per 1000 MW(e) PWR installed from the ore mining till reprocessing of irradiated fuels. The methods of treatment concentration, fixation, final storable forms for liquid and solid waste of low activity and for high level activity waste. The storage of radioactive waste and the plutonium-bearing waste treatement are also considered. The estimated quantity of wastes produced for 5450 MW(e) in Belgium and their destination are presented. (A.F.)

Probability safety assessment of LOOP accident to molten salt reactor

International Nuclear Information System (INIS)

Mei Mudan; Shao Shiwei; Yu Zhizhen; Chen Kun; Zuo Jiaxu

2013-01-01

Background: Loss of offsite power (LOOP) is a possible accident to any type of reactor, and this accident can reflect the main idea of reactor safety design. Therefore, it is very important to conduct a study on probabilistic safety assessment (PSA) of the molten salt reactor that is under LOOP circumstance. Purpose: The aim is to calculate the release frequency of molten salt radioactive material to the core caused by LOOP, and find out the biggest contributor to causing the radioactive release frequency. Methods: We carried out the PSA analysis of the LOOP using the PSA process risk spectrum, and assumed that the primary circuit had no valve and equipment reliability data based on the existing mature power plant equipment reliability data. Results: Through the PSA analysis, we got the accident sequences of the release of radioactive material to the core caused by LOOP and its frequency. The results show that the release frequency of molten salt radioactive material to the core caused by LOOP is about 2×10 -11 /(reactor ·year), which is far below that of the AP1000 LOOP. In addition, through the quantitative analysis, we obtained the point estimation and interval estimation of uncertainty analysis, and found that the biggest contributor to cause the release frequency of radioactive material to the core is the reactor cavity cooling function failure. Conclusion: This study provides effective help for the design and improvement of the following molten salt reactor system. (authors)

Radioactivity. From radioelements to scientific applications

International Nuclear Information System (INIS)

2002-09-01

Radioactivity was not invented by man. It was discovered just over a century ago, in 1896, by the French physicist Henri Becquerel. He was attempting to find out whether the rays emitted by fluorescent uranium salts were the same as the X-rays discovered in 1895 by the German physicist Wilhelm Roentgen. He thought that the uranium salts, after being excited by light, emitted these X-rays. Imagine his surprise when, in Paris in March 1896, he discovered that photographic film had been exposed without 'Radioactivity was not invented by man. It is a natural phenomenon that was discovered at the end of the 19. century'. He concluded that uranium emitted invisible radiation, different from X-rays, spontaneously and inexhaustibly. The phenomenon he discovered was named radioactivity (from the Latin radius, meaning ray). Following Henri Becquerel's work, in 1898 Pierre and Marie Curie isolated polonium and radium, unknown radioactive elements present in uranium ore. (authors)

Radioactive waste isolation in salt: rationale and methodology for Argonne-conducted reviews of site characterization programs

International Nuclear Information System (INIS)

Harrison, W.; Ditmars, J.D.; Tisue, M.W.; Hambley, D.F.; Fenster, D.F.; Rote, D.M.

1985-07-01

Both regulatory and technical concerns must be addressed in Argonne-conducted peer reviews of site characterization programs for individual sites for a high-level radioactive waste repository in salt. This report describes the regulatory framework within which reviews must be conducted and presents background information on the structure and purpose of site characterization programs as found in US Nuclear Regulatory Commission (NRC) Regulatory Guide 4.17 and Title 10, Part 60, of the Code of Federal Regulations. It also presents a methodology to assist reviewers in addressing technical concerns relating to their respective areas of expertise. The methodology concentrates on elements of prime importance to the US Department of Energy's advocacy of a given salt repository system during the NRC licensing process. Instructions are given for reviewing 12 site characterization program elements, starting with performance objectives, performance issues, and levels of performance of repository subsystem components; progressing through performance assessment; and ending with plans for data acquisition and evaluation. The success of a site characterization program in resolving repository performance issues will be determined by judging the likelihood that the proposed data acquisition activities will reduce uncertainties in the performance predictions. 8 refs., 3 figs., 5 tabs

Method of decomposing treatment for radioactive organic phosphate wastes

International Nuclear Information System (INIS)

Uki, Kazuo; Ichihashi, Toshio; Hasegawa, Akira; Sato, Tatsuaki

1985-01-01

Purpose: To decompose the organic phosphoric-acid ester wastes containing radioactive material, which is produced from spent fuel reprocessing facilities, into inorganic materials using a simple device, under moderate conditions and at high decomposing ratio. Method: Radioactive organic phosphate wates are oxidatively decomposed by H 2 O 2 in an aqueous phosphoric-acid solution of metal phosphate salts. Copper phosphates are used as the metal phosphate salts and the decomposed solution of the radioactive organic phosphate wastes is used as the aqueous solution of the copper phosphate. The temperature used for the oxidizing decomposition ranges from 80 to 100 0 C. (Ikeda, J.)

Hydrologic environment of the Silurian salt deposits in parts of Michigan, Ohio, and New York

Science.gov (United States)

Norris, Stanley E.

1978-01-01

The aggregate thickness of evaporites (salt, gypsum, and anhydrite) in the Silurian Salina sequence in Michigan exceeds 1200 feet in areas near the periphery of the Michigan basin, where the salt beds are less than 3000 feet below land surface. In northeast Ohio the aggregate thickness of salt beds is as much as 200 feet in places, and in western New York it is more than 500 feet, where th beds are less than 3000 feet deep. The salt-bearing rocks dip regionally on the order of 50 feet per mile; those in Michigan dip toward the center of the Michigan basin, and those in Ohio and New York, in the Appalachian basin, dip generally southward. The rocks in both basins thicken downdip. Minor folds and faults occur in the salt-bearing rocks in all three states. Some of this defrmation has been attenuated or absorbed bo the salt beds. Occuring near the middle of thick sedimentary sequences, the salt beds are bounded aboe and below by beds containing water having dissolved-solids concentrations several times that seawter. The brines occur commonly in discrete zones of high permeability at specific places in the stratigraphic sequence. In northeast Ohio two prominent brine zones are recognized by the driller, the Devonian Oriskany Sandstone, or 'first water' zone, above the Salina Formation, and the Newburg or 'second water' zone below the Salina. In each aquifer there is a vertical component of hydraulic head, but little brine probably moves through the salt beds because their permeability is extremely low. Also, ther is little evidence of dissolution of the salt in areas distant from the outcrop, suggesting that if brine does move through the salt, movement is at a slow enough rate so that, in combination with the saturated or near-saturated condition of the water, it precludes significant dissolution. Principal brine movement is probably in the permeable zones in the direction of the hydraulic gradient. Two areas in Michigan and one area each in Ohio and New York appear

Development of a glass matrix for vitrification of sulphate bearing high level radioactive liquid waste

International Nuclear Information System (INIS)

Kaushik, C.P.; Mishra, R.K.; Thorat, Vidya; Ramchandran, M.; Amar Kumar; Ozarde, P.D.; Raj, Kanwar; Das, D.

2004-07-01

High level radioactive liquid waste (HLW) is generated during reprocessing of spent nuclear fuel. In the earlier reprocessing flow sheet ferrous sulphamate has been used for valancy adjustment of Pu from IV to III for effective separation. This has resulted in generation of HLW containing significance amount of sulphate. Internationally borosilicate glass matrix has been adopted for vitrification of HLW. The first Indian vitrification facility at Waste Immobilislition Plant (WIP), Tarapur a five component borosilicate matrix (SiO 2 :B 2 O 3 :Na 2 O : MnO : TiO 2 ) has been used for vitrification of waste. However at Trombay HLW contain significant amount of sulphate which is not compatible with standard borosilicate formulation. Extensive R and D efforts were made to develop a glass formulation which can accommodate sulphate and other constituents of HLW e.g., U, Al, Ca, etc. This report deals with development work of a glass formulations for immobilization of sulphate bearing waste. Different glass formulations were studied to evaluate the compatibility with respect to sulphate and other constituents as mentioned above. This includes sodium, lead and barium borosilicate glass matrices. Problems encountered in different glass matrices for containment of sulphate have also been addressed. A glass formulation based on barium borosilicate was found to be effective and compatible for sulphate bearing high level waste. (author)

The HAW-project: Demonstration facility for the disposal of high-level waste in salt

International Nuclear Information System (INIS)

Rothfuchs, T.; Duijves, K.A.

1990-04-01

The HAW-project plants the testwise emplacement of 30 vitrified highly radioactive canisters containing Cs-137 and Sr-90 at the 800 m level of the Asse salt mine for a testing period of approximately five years. The major objective of this project is the pilot testing and demonstration of safe methods for the final disposal of high-level radioactive waste (HAW) in geological salt formations. During the years 1985 to 1989 the underground test field was excavated, the measuring equipment installed, and two preceedings inactive electrical tests taken into operation. Furthermore, the components of a system for transportation and emplacement of highly radioactive canisters was fabricated, installed, and preliminarily tested. After some delays in the licensing procedure the emplacement of the 30 radioactive canisters is now envisaged for early 1991. For handling of the radioactive canisters and their emplacement into the boreholes a system consisting of a transport cask, a transport vehicle, a disposal machine, and of a borehole slider has been developed and will be tested. The actual scientific investigation programme is based on the estimation and observation of the interaction between the radioactive canisters and the rock salt. This programme includes measurement of thermally and radiolytically induced water and gas release from the rock salt and the radiolytical decomposition of salt minerals. Also the thermally induced stress and deformation fields in the surrounding rock mass will be investigated carefully. (orig./HP)

Dicarbonic acid anilides containing radioactive iodine (iodine 131, 123, 125, or 132) as well as their metal and amine salts; methods for the preparation of these compounds and of radioactive functional diagnostics containing them

International Nuclear Information System (INIS)

Buttermann, G.

1976-01-01

A method for the preparation of dicarbonic acid anilides containing radioactive iodine is described. The initial substances are N,N dimethyl-p-toluene sulfonamide, N,N bis-dimethyl aminosulfon, or dimethyl sulfon, or mixtures of these, which are heated in aqueous solution or in a melt with an alkali or alkaline earth radioiodide as carrier-free as possible. From the water-soluble salts of the obtained iodine-labelled dicarbonic acid anilides aqueous solutions are produced with 1 mg up to 5 g iodine-labelled dicarbonic acid anilide per 10 ml and an activity of 0.025 and 25 mCi per ml with physiologically compatible bases as radioactive functional diagnostics. (RB) [de

Geology and undiscovered resource assessment of the potash-bearing Central Asia Salt Basin, Turkmenistan, Uzbekistan, Tajikistan, and Afghanistan: Chapter AA in Global mineral resource assessment

Science.gov (United States)

Wynn, Jeff; Orris, Greta J.; Dunlap, Pamela; Cocker, Mark D.; Bliss, James D.

2016-03-23

Undiscovered potash resources in the Central Asia Salt Basin (CASB) of Turkmenistan, Uzbekistan, Tajikistan, and Afghanistan were assessed as part of a global mineral resource assessment led by the U.S. Geological Survey. The term “potash” refers to potassium-bearing, water-soluble salts derived from evaporite basins, where seawater dried up and precipitated various salt compounds; the word for the element “potassium” is derived from potash. Potash is produced worldwide at amounts exceeding 30 million metric tons per year, mostly for use in fertilizers. The term “potash” is used by industry to refer to potassium chloride, as well as potassium in sulfate, nitrate, and oxide forms. For the purposes of this assessment, the term “potash” refers to potassium ores and minerals and potash ore grades. Resource and production values are usually expressed by industry in terms of K2O (potassium oxide) or muriate of potash (KCl, potassium chloride).

Radiation damage studies on natural rock salt from various geological localities of interest to the radioactive waste disposal program

International Nuclear Information System (INIS)

Levy, P.W.

1981-01-01

As part of a program to investigate radiation damage in geological materials of interest to the radioactive waste disposal program, radiation damage, particularly radiation induced sodium metal colloid formation, has been studied in 14 natural rock salt samples. All measurements were made with equipment for making optical absorption and other measurements on samples, in a temperature controlled irradiation chamber, during and after 0.5 to 3.0 MeV electron irradiation. Samples were chosen for practical and scientific purposes, from localities that are potential repository sites and from different horizons at certain localities

SOLUTION MINING IN SALT DOMES OF THE GULF COAST EMBAYMENT

Energy Technology Data Exchange (ETDEWEB)

Griswold, G. B.

1981-02-01

Following a description of salt resources in the salt domes of the gulf coast embayment, mining, particularly solution mining, is described. A scenario is constructed which could lead to release of radioactive waste stored in a salt dome via inadvertent solution mining and the consequences of this scenario are analyzed.

Uranium- and thorium-bearing pegmatites of the United States

International Nuclear Information System (INIS)

Adams, J.W.; Arengi, J.T.; Parrish, I.S.

1980-04-01

This report is part of the National Uranium Resource Evaluation (NURE) Program designed to identify criteria favorable for the occurrence of the world's significant uranium deposits. This project deals specifically with uranium- and thorium-bearing pegmatites in the United States and, in particular, their distribution and origin. From an extensive literature survey and field examination of 44 pegmatite localities in the United States and Canada, the authors have compiled an index to about 300 uranium- and thorium-bearing pegmatites in the United States, maps giving location of these deposits, and an annotated bibliography to some of the most pertinent literature on the geology of pegmatites. Pegmatites form from late-state magma differentiates rich in volatile constituents with an attendant aqueous vapor phase. It is the presence of an aqueous phase which results in the development of the variable grain size which characterizes pegmatites. All pegmatites occur in areas of tectonic mobility involving crustal material usually along plate margins. Those pegmatites containing radioactive mineral species show, essentially, a similar distribution to those without radioactive minerals. Criteria such as tectonic setting, magma composition, host rock, and elemental indicators among others, all serve to help delineate areas more favorable for uranium- and thorium-bearing pegmatites. The most useful guide remains the radioactivity exhibited by uranium- and thorium-bearing pegmatites. Although pegmatites are frequently noted as favorable hosts for radioactive minerals, the general paucity and sporadic distribution of these minerals and inherent mining and milling difficulties negate the resource potential of pegmatites for uranium and thorium

Uranium- and thorium-bearing pegmatites of the United States

Energy Technology Data Exchange (ETDEWEB)

Adams, J.W.; Arengi, J.T.; Parrish, I.S.

1980-04-01

This report is part of the National Uranium Resource Evaluation (NURE) Program designed to identify criteria favorable for the occurrence of the world's significant uranium deposits. This project deals specifically with uranium- and thorium-bearing pegmatites in the United States and, in particular, their distribution and origin. From an extensive literature survey and field examination of 44 pegmatite localities in the United States and Canada, the authors have compiled an index to about 300 uranium- and thorium-bearing pegmatites in the United States, maps giving location of these deposits, and an annotated bibliography to some of the most pertinent literature on the geology of pegmatites. Pegmatites form from late-state magma differentiates rich in volatile constituents with an attendant aqueous vapor phase. It is the presence of an aqueous phase which results in the development of the variable grain size which characterizes pegmatites. All pegmatites occur in areas of tectonic mobility involving crustal material usually along plate margins. Those pegmatites containing radioactive mineral species show, essentially, a similar distribution to those without radioactive minerals. Criteria such as tectonic setting, magma composition, host rock, and elemental indicators among others, all serve to help delineate areas more favorable for uranium- and thorium-bearing pegmatites. The most useful guide remains the radioactivity exhibited by uranium- and thorium-bearing pegmatites. Although pegmatites are frequently noted as favorable hosts for radioactive minerals, the general paucity and sporadic distribution of these minerals and inherent mining and milling difficulties negate the resource potential of pegmatites for uranium and thorium.

Radioactive equilibrium of uranium-bearing ores in some problems of applied geology; Les equilibres radioactifs des menerais uraniferes dans quelques problemes de geologie appliquee

Energy Technology Data Exchange (ETDEWEB)

Coulomb, R; Girard, Ph; Goldsztein, M [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

1964-07-01

The state of equilibrium between several nuclides in radioactive relationship is determined with accuracy by the fundamental equations of radioactivity. It can be measured physically and expressed in suitable and internationally adopted units; Equilibrium - disequilibrium of uranium-bearing ores is a fairly complex phenomenon but the problem can be much simplified by well-chosen approximations in various practical field cases. The results of radiometric and radiochemical measurements lead to the interpretation of geochemical anomalies and may be used in the qualitative and quantitative estimation of uranium bearing deposits. (authors) [French] L'etat d'equilibre entre plusieurs radioelements en filiation se definit avec precision par les equations fondamentales de la radioactivite et peut etre determine par des mesures physiques dans des systemes d'unites commodes et internationalement adoptes. Le probleme general equilibre-desequilibre des minerais uraniferes est relativement complexe, mais peut se simplifier largement par des approximations judicieuses dans de nombreux cas particuliers rencontres concretement sur le terrain. Les resultats des mesures radiometriques et radiochimiques permettent l'interpretation des anomalies geochimiques et peuvent servir a l'estimation qualitative et quantitative des gisements de minerais uraniferes. (auteurs)

Aspects of the storage of radioactive waste

International Nuclear Information System (INIS)

Nienhuys, K.

1978-01-01

The expansion in the number of nuclear power stations in the netherlands is amongst other things, dependent on an acceptable policy for the storage of the waste from the stations. Consequently the idea has developed for storage in a salt-dome. The sub-committee on radioactive waste substances of the Interdepartmental Committee for Nuclear Energy has therefore given a mandate to initiate further research. For the risk analysis over the definitive storage of nuclear waste the sub-comittee produced a report in 1975, entitled 'Safety analysis for the underground storage of nuclear waste in salt-dome outcrops'. The analysis reveals a number of defective features. This makes especially clear that statements about the definitive storage of nuclear waste in salt domes can only be made with a great deal of uncertainty. There is no guarantee that the nuclear waste generated may be stowed away so that it will never return to the ionosphere. The speed whereby the nuclear waste may return would be dependent on a combination of events which cannot generally be calculated or assessed. The long term consequences of an irreversible radioactive contamination of the biosphere is not acceptable. There is insufficient proof that the storage of radioactive waste in salt domes is feasible. (G.C.)

Long-term interactions of full-scale cemented waste simulates with salt brines

Energy Technology Data Exchange (ETDEWEB)

Kienzler, B.; Borkel, C.; Metz, V.; Schlieker, M.

2016-07-01

Since 1967 radioactive wastes have been disposed of in the Asse II salt mine in Northern Germany. A significant part of these wastes originated from the pilot reprocessing plant WAK in Karlsruhe and consisted of cemented NaNO{sub 3} solutions bearing fission products, actinides, as well as process chemicals. With respect to the long-term behavior of these wastes, the licensing authorities requested leaching experiments with full scale samples in relevant salt solutions which were performed since 1979. The experiments aimed at demonstrating the transferability of results obtained with laboratory samples to real waste forms and at the investigation of the effects of the industrial cementation process on the properties of the waste forms. This research program lasted until 2013. The corroding salt solutions were sampled several times and after termination of the experiments, the solid materials were analyzed by various methods. The results presented in this report cover the evolution of the solutions and the chemical and mineralogical characterization of the solids including radionuclides and waste components, and the paragenesis of solid phases (corrosion products). The outcome is compared to the results of model calculations. For safety analysis, conclusions are drawn on radionuclide retention, evolution of the geochemical environment, evolution of the density of solutions, and effects of temperature and porosity of the cement waste simulates on cesium mobilization.

Long-term interactions of full-scale cemented waste simulates with salt brines

International Nuclear Information System (INIS)

Kienzler, B.; Borkel, C.; Metz, V.; Schlieker, M.

2016-01-01

Since 1967 radioactive wastes have been disposed of in the Asse II salt mine in Northern Germany. A significant part of these wastes originated from the pilot reprocessing plant WAK in Karlsruhe and consisted of cemented NaNO 3 solutions bearing fission products, actinides, as well as process chemicals. With respect to the long-term behavior of these wastes, the licensing authorities requested leaching experiments with full scale samples in relevant salt solutions which were performed since 1979. The experiments aimed at demonstrating the transferability of results obtained with laboratory samples to real waste forms and at the investigation of the effects of the industrial cementation process on the properties of the waste forms. This research program lasted until 2013. The corroding salt solutions were sampled several times and after termination of the experiments, the solid materials were analyzed by various methods. The results presented in this report cover the evolution of the solutions and the chemical and mineralogical characterization of the solids including radionuclides and waste components, and the paragenesis of solid phases (corrosion products). The outcome is compared to the results of model calculations. For safety analysis, conclusions are drawn on radionuclide retention, evolution of the geochemical environment, evolution of the density of solutions, and effects of temperature and porosity of the cement waste simulates on cesium mobilization.

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)

Idaho Nuclear Technology and Engineering Center (INTEC) Sodium Bearing Waste - Waste Incidental to Reprocessing Determination

International Nuclear Information System (INIS)

Jacobson, Victor Levon

2002-01-01

U.S. Department of Energy Manual 435.1-1, Radioactive Waste Management, Section I.1.C, requires that all radioactive waste subject to Department of Energy Order 435.1 be managed as high-level radioactive waste, transuranic waste, or low-level radioactive waste. Determining the radiological classification of the sodium-bearing waste currently in the Idaho Nuclear Technology and Engineering Center Tank Farm Facility inventory is important to its proper treatment and disposition. This report presents the technical basis for making the determination that the sodium-bearing waste is waste incidental to spent fuel reprocessing and should be managed as mixed transuranic waste. This report focuses on the radiological characteristics of the sodium-bearing waste. The report does not address characterization of the nonradiological, hazardous constituents of the waste in accordance with Resource Conservation and Recovery Act requirements

Radioactivity levels in soil of salt field area Kelambakkam, Tamil Nadu, India

International Nuclear Information System (INIS)

Ravisankar, R.; Rajalakshmi, A.; Manikandan, E.; Gajendiran, V.; Meenakshisundaram, V.

2006-01-01

Mother nature has gifted mankind with lot of precious gifts. Common salt is one of them. In the globe, Tamilnadu is one of the ideal locations for producing salt. Kelambakkam salt field area is one of the leading producers of salt in global market. The climate, soil and availability of brine are a great asset for producing quality salts. In the present work, the primordial radionuclides concentration in soil samples collected in and around the salt field area, Kelambakkam, Tamilnadu was measured using gamma ray spectrometer

The HAW-project: Demonstration facility for the disposal of high-level waste in salt

International Nuclear Information System (INIS)

Rothfuchs, T.; Duijves, K.A.; Mueller-Lyda, I.

1990-04-01

To satisfy the test objectives thirty highly radioactive canisters containing the radionuclides Cs-137 and Sr-90 will be emplaced in six boreholes located in two test galleries at the 800 m-level in the Asse salt mine. For handling of the radioactive canisters and their emplacement into the boreholes a system consisting of a transport cask, a transport vehicle, a disposal machine, and of a borehole slider has been developed. The actual scientific investigation programme is based on the estimation and observation of the interaction between the radioactive canisters and the rock salt. This programme includes measurement of thermally and radiolytically induced water and gas release from the rock salt and the radiolytical decomposition of salt minerals. Also the thermally induced stress and deformation fields in the surrounding rock mass will be investigated carefully. (orig./DG)

Residual radioactivity of treated green diamonds.

Science.gov (United States)

Cassette, Philippe; Notari, Franck; Lépy, Marie-Christine; Caplan, Candice; Pierre, Sylvie; Hainschwang, Thomas; Fritsch, Emmanuel

2017-08-01

Treated green diamonds can show residual radioactivity, generally due to immersion in radium salts. We report various activity measurements on two radioactive diamonds. The activity was characterized by alpha and gamma ray spectrometry, and the radon emanation was measured by alpha counting of a frozen source. Even when no residual radium contamination can be identified, measurable alpha and high-energy beta emissions could be detected. The potential health impact of radioactive diamonds and their status with regard to the regulatory policy for radioactive products are discussed. Copyright © 2017. Published by Elsevier Ltd.

Identification of release scenarios for a repository of radioactive waste in a salt dome in the Netherlands

International Nuclear Information System (INIS)

Glasbergen, P.; Hamstra, J.

1981-01-01

A review is presented of the long-term scenarios used in the safety analysis which was carried out for the disposal of radioactive waste in salt domes in the Netherlands. The long-term analysis involved the following natural processes or events: climatological and sea-level changes, glacial erosion, diapirism, subsidence, faulting and dissolution. The model calculations which were carried out showed the dominant parameters: the rate of diapirism and the rate of subsurface dissolution of rock salt. During the operational period the intrusion of water in the repository was considered to be the most hazardous event. Because the layout of the disposal mine, the disposal geometry and the disposal mining procedures were still under consideration, the first approach of a release scenario was made on a generic basis. A generic scenario is presented for the events during the flooding of the repository. The transport ways of water through the repository and its surroundings are indicated. It is concluded that release scenario analysis for long-term periods and for the operational period provides essential information to optimize the overall disposal system in an iterative process

Fast molten salt reactor-transmuter for closing nuclear fuel cycle on minor actinides

International Nuclear Information System (INIS)

Dudnikov, A. A.; Alekseev, P. N.; Subbotin, S. A.

2007-01-01

Creation fast critical molten salt reactor for burning-out minor actinides and separate long-living fission products in the closed nuclear fuel cycle is the most perspective and actual direction. The reactor on melts salts - molten salt homogeneous reactor with the circulating fuel, working as burner and transmuter long-living radioactive nuclides in closed nuclear fuel cycle, can serve as an effective ecological cordon from contamination of the nature long-living radiotoxic nuclides. High-flux fast critical molten-salt nuclear reactors in structure of the closed nuclear fuel cycle of the future nuclear power can effectively burning-out / transmute dangerous long-living radioactive nuclides, make radioisotopes, partially utilize plutonium and produce thermal and electric energy. Such reactor allows solving the problems constraining development of large-scale nuclear power, including fueling, minimization of radioactive waste and non-proliferation. Burning minor actinides in molten salt reactor is capable to facilitate work solid fuel power reactors in system NP with the closed nuclear fuel cycle and to reduce transient losses at processing and fabrications fuel pins. At substantiation MSR-transmuter/burner as solvents fuel nuclides for molten-salt reactors various salts were examined, for example: LiF - BeF2; NaF - LiF - BeF2; NaF-LiF ; NaF-ZrF4 ; LiF-NaF -KF; NaCl. RRC 'Kurchatov institute' together with other employees have developed the basic design reactor installations with molten salt reactor - burner long-living nuclides for fluoride fuel composition with the limited solubility minor actinides (MAF3 10 mol %) allows to develop in some times more effective molten salt reactor with fast neutron spectrum - burner/ transmuter of the long-living radioactive waste. In high-flux fast reactors on melts salts within a year it is possible to burn ∼300 kg minor actinides per 1 GW thermal power of reactor. The technical and economic estimation given power

Trial storage of high-level waste in the Asse II salt mine

International Nuclear Information System (INIS)

1984-01-01

This report covers a second phase of the work performed by GSF and KfK in the Asse II salt mine, with a view to disposal of radioactive waste in salt formations. New items of the research were geophysical investigations of the behaviour of heated salt and preparation of a trial storage in the Asse II salt mine

A review of in situ investigations in salt

International Nuclear Information System (INIS)

Kuehn, K.

1985-01-01

In situ investigations for the disposal of radioactive wastes in rock salt formations have the longest history in the field. Well known names are Project Salt Vault (PSV) which was performed in the Lyons Mine, Kansas/USA, and the Asse salt mine in Germany. The overall objective for in situ investigations is twofold: 1. To produce all necessary data for the construction and operation of repositories and 2. to produce all necessary data for a performance assessment for repositories

Understanding radioactive waste

International Nuclear Information System (INIS)

Murray, R.L.

1981-12-01

This document contains information on all aspects of radioactive wastes. Facts are presented about radioactive wastes simply, clearly and in an unbiased manner which makes the information readily accessible to the interested public. The contents are as follows: questions and concerns about wastes; atoms and chemistry; radioactivity; kinds of radiation; biological effects of radiation; radiation standards and protection; fission and fission products; the Manhattan Project; defense and development; uses of isotopes and radiation; classification of wastes; spent fuels from nuclear reactors; storage of spent fuel; reprocessing, recycling, and resources; uranium mill tailings; low-level wastes; transportation; methods of handling high-level nuclear wastes; project salt vault; multiple barrier approach; research on waste isolation; legal requiremnts; the national waste management program; societal aspects of radioactive wastes; perspectives; glossary; appendix A (scientific American articles); appendix B (reference material on wastes)

Understanding radioactive waste

Energy Technology Data Exchange (ETDEWEB)

Murray, R.L.

1981-12-01

This document contains information on all aspects of radioactive wastes. Facts are presented about radioactive wastes simply, clearly and in an unbiased manner which makes the information readily accessible to the interested public. The contents are as follows: questions and concerns about wastes; atoms and chemistry; radioactivity; kinds of radiation; biological effects of radiation; radiation standards and protection; fission and fission products; the Manhattan Project; defense and development; uses of isotopes and radiation; classification of wastes; spent fuels from nuclear reactors; storage of spent fuel; reprocessing, recycling, and resources; uranium mill tailings; low-level wastes; transportation; methods of handling high-level nuclear wastes; project salt vault; multiple barrier approach; research on waste isolation; legal requiremnts; the national waste management program; societal aspects of radioactive wastes; perspectives; glossary; appendix A (scientific American articles); appendix B (reference material on wastes). (ATT)

Long-term sealing of openings in salt formations

International Nuclear Information System (INIS)

Walter, F.; Stockmann, N.; Yaramanci, U.; Laurens, J.F.

1993-01-01

Radioactive wastes can be disposed of in deep salt formations. Rock salt is a suitable geologic medium because of its unique characteristics. Open boreholes, shafts and drifts are created to provide physical access to the repository. Long-term seals must be emplaced in those potential pathways to prevent radioactive release to the biosphere. The sealing materials must be mechanically and, most important, geochemically stable within the host rock. Salt bricks made of compressed salt-powder are understood to be the first choice long-term sealing material. Seals built from salt bricks will be ductile. The permeability of the salt bricks is assumed to be in the order of 2*10 -15 m 2 . Large sealing systems are built by combining the individual bricks with mortar. Raw materials for mortar are fine-grained halite powder and ground saliferous clay. The permeability of the mortar decreases with its salt content to approx. 2*10 -14 m 2 . Moistened saliferous clay may show temporary swelling. Sealing experiments will be carried out in the Asse salt mine. Long-term seals will be built into holes of 1 m diameter. The contact and merging of the brick-wall with the surrounding rock salt will be investigated in long-term tests. Within the in situ sealing program a number of geophysical methods are applied. Acoustic emission measurements are used to study the effects of high pressure gas injection and a geoelectrical observation program is aiming to estimate the permeability in and around the long-term seal. High frequency electromagnetic methods contribute to the knowledge of the petrophysical rock properties. 11 refs., 12 figs

Topotactic transformations of sodalite cages: synthesis and NMR study of mixed salt-free and salt-bearing sodalites.

Science.gov (United States)

Trill, Henning; Eckert, Hellmut; Srdanov, Vojislav I

2002-07-17

A series of mixed sodalite samples, Na(8)[Al(6)Si(6)O(24)]Br(x).(H(3)O(2))(2-x), with the unit cell stoichiometries varying in the 0 < x <2 region, was made by hydrothermal synthesis and subsequently transformed into Na(6+x)[Al(6)Si(6)O(24)]Br(x).(4H(2)O)(2-x) and Na(6+x)[Al(6)Si(6)O(24)]Br(x).circle(2-x) sodalites. Here, circle refers to an empty sodalite cage. The three series, referred hereafter to as the Br/basic, Br/hydro, and Br/dry series, were characterized by powder diffraction X-ray and by (23)Na, (27)Al, and (81)Br magic angle spinning (MAS) NMR and high-resolution triple quantum (TQ) MAS NMR spectroscopy. We determined that incorporation of Br(-) anions is 130 times more preferred than incorporation of H(3)O(2)(-) anions during the formation of sodalite cages, which permitted precise control of the halide content in the solid. Monotonic trends in chemical shifts were observed as a function of cage occupancy, reflecting continuous changes in structural parameters. A linear correlation between (81)Br chemical shift and lattice constant with a slope of -86 ppm/A was observed for all three series. Likewise, (23)Na chemical shifts for Na(+) cations in salt-bearing sodalite cages correlate linearly with the lattice constant. Both results indicate a universal dependence of the (23)Na and (81)Br chemical shifts on the Na-Br distance. The (27)Al chemical shifts of Br/basic and Br/hydro sodalites obey an established relation between delta(cs) and the average T-O-T bond angle of 0.72 ppm/degrees. Br/dry sodalites show two aluminum resonances, characterized by significantly different chemical shifts and quadrupolar interaction parameters. In that series, local symmetry distortions are evident from strong quadrupolar perturbations in the NMR spectra. P(Q) values for (27)Al vary between 0.8 MHz in Br/basic sodalites and 4.4 MHz in the Br/dry series caused by deviations from the tetrahedral symmetry of the salt-free sodalite cages. For (23)Na, P(Q) values of 0.8, 0

Mechanical behavior of New Mexico rock salt in triaxial compression up to 2000C

International Nuclear Information System (INIS)

Wawersik, W.R.; Hannum, D.W.

1978-01-01

An extensive experimental program is being conducted to determine the mechanical behavior of New Mexico rock salt in support of the structural design of a Radioactive Waste Isolation Pilot Plant (WIPP). In this initial report, three groups of tests are discussed to identify the relative and site-specific importance of deviator stress, confining pressure (mean stress), temperature, time (loading rate), and stress path. The three groups of experiments consist of (1) hydrostatic loading, (2) conventional triaxial compression tests (sigma 1 > sigma 2 = sigma 3 = const.), and (3) variable stress path tests including experiments at approximately constant sigma 1 and at constant mean stress. All data were generated on 100 mm diameter specimens. The rock salt exhibited nonlinear response under all loading conditions, practically zero initial elastic limit and an apparent inseparability of permanent deformations into time-independent and time-dependent components. Pressure and temperature did not alter the elastic constants but affected the principal strain ratio, the ratio volumetric strain/shear strain, rock salt ductility, and the ultimate stress. In particular, low pressure and temperature permitted pronounced dilatancy and loss in load bearing ability. Under such conditions the volumetric strains reach sizable fractions of the shear strains. Pressure remained important even at high temperature because it influenced the rate of shearing. Load path and stress history may be significant under deviatoric loading conditions and for large variations in pressure

Radioactive-waste isolation pilot plant

International Nuclear Information System (INIS)

Weart, W.D.

1977-01-01

The objective of the Waste Isolation Pilot Plant (WIPP) program is to demonstrate the suitability of bedded salt, specifically, the bedded salt deposits in the Los Medanos area of southeastern New Mexico, as a disposal medium for radioactive wastes. Our program responsibilities include site selection considerations, all aspects of design and development, technical guidance of facility operation, environmental impact assessment, and technical support to ERDA for developing public understanding of the facility

Geologic disposal of nuclear wastes: salt's lead is challenged

International Nuclear Information System (INIS)

Kerr, R.A.

1979-01-01

The types of radioactive waste disposal sites available are outlined. The use of salt deposits and their advantages are discussed. The reasons for the selection of the present site for the Waste Isolation Pilot Plant are presented. The possibilities of using salt domes along the Gulf Coast and not-salt rocks as nuclear waste repositories are also discussed. The sea bed characteristics are described and advantages of this type of site selection are presented

The electrochemical signature of functionalized single-walled carbon nanotubes bearing electroactive groups

International Nuclear Information System (INIS)

Le Floch, Fabien; Thuaire, Aurelie; Simonato, Jean-Pierre; Bidan, Gerard

2009-01-01

We report the modification and characterization of single-walled carbon nanotubes (SWCNTs) in view of molecular sensing applications. We found that ultrasonicated SWCNTs present sticking properties that make them adhere on electrode surfaces. This allows excellent characterization of SWCNTs by cyclic voltammetry (CV) before and after chemical functionalization with diazonium salts bearing electroactive groups. Bare SWCNTs presented distinct invariant shapes in CV, used as control curves, in comparison with functionalized SWCNTs for which specific signatures corresponding to the presence of grafted molecules were identified. According to the electronic substituents in the para position of the diazonium salts, divergent behaviours were observed for the grafting reactions. Diazonium salts having electrowithdrawing groups could be grafted without electrochemical induction whereas those bearing electron donating groups required a cathodic potential to generate the formation of the radical species.

The electrochemical signature of functionalized single-walled carbon nanotubes bearing electroactive groups

Energy Technology Data Exchange (ETDEWEB)

Le Floch, Fabien; Thuaire, Aurelie; Simonato, Jean-Pierre [LITEN/DTNM/LCRE, CEA-Grenoble 17 rue des Martyrs, 38054 Grenoble cedex 9 (France); Bidan, Gerard [INAC/DIR, CEA-Grenoble 17 rue des Martyrs, 38054 Grenoble cedex 9 (France)], E-mail: jean-pierre.simonato@cea.fr

2009-04-08

We report the modification and characterization of single-walled carbon nanotubes (SWCNTs) in view of molecular sensing applications. We found that ultrasonicated SWCNTs present sticking properties that make them adhere on electrode surfaces. This allows excellent characterization of SWCNTs by cyclic voltammetry (CV) before and after chemical functionalization with diazonium salts bearing electroactive groups. Bare SWCNTs presented distinct invariant shapes in CV, used as control curves, in comparison with functionalized SWCNTs for which specific signatures corresponding to the presence of grafted molecules were identified. According to the electronic substituents in the para position of the diazonium salts, divergent behaviours were observed for the grafting reactions. Diazonium salts having electrowithdrawing groups could be grafted without electrochemical induction whereas those bearing electron donating groups required a cathodic potential to generate the formation of the radical species.

Electrodialysis-ion exchange for the separation of dissolved salts

International Nuclear Information System (INIS)

Baroch, C.J.; Grant, P.J.

1995-01-01

The Department of Energy generates and stores a significant quantity of low level, high level, and mixed wastes. As some of the DOE facilities are decontaminated and decommissioned, additional and possibly different forms of wastes will be generated. A significant portion of these wastes are aqueous streams containing acids, bases, and salts, or are wet solids containing inorganic salts. Some of these wastes are quite dilute solutions, whereas others contain large quantities of nitrates either in the form of dissolved salts or acids. Many of the wastes are also contaminated with heavy metals, radioactive products, or organics. Some of these wastes are in storage because a satisfactory treatment and disposal processes have not been developed. This report describes the process of electrodialysis-ion exchange (EDIX) for treating aqueous wastes streams consisting of nitrates, sodium, organics, heavy metals, and radioactive species

Geology, hydrology, thickness and quality of salt at three alternate sites for disposal of radioactive waste in Kansas

International Nuclear Information System (INIS)

Bayne, C.K.; Brinkley, C.

1972-09-01

The three sites selected by the AEC for additional study for the disposal of radioactive wastes in Kansas are; Site A located in south-central Lincoln County, Site D-2 located in south-central Wichita County, and Site A-1 located in north-western Lincoln County. Results of the study show that all sites failed to meet the detailed criteria. Areas A and A-1 fail to meet the criteria concerning thickness and quality. Area D-2 fails to meet the criteria concerning quality and mineability of the salt. Areas west of Site A-1 and in south-central Harper County, in the authors' opinion, appear to be the best prospects for future study in Kansas

Assessment of tectonic hazards to waste storage in interior-basin salt domes

International Nuclear Information System (INIS)

Kehle, R.

1979-01-01

Salt domes in the northern Gulf of Mexico may make ideal sites for storage of radioactive waste because the area is tectonically quiet. The stability of such salt domes and the tectonic activity are discussed

Fixation of radioactive waste by reaction with clays: progress report

International Nuclear Information System (INIS)

Delegard, C.H.; Barney, G.S.

1975-07-01

Reactions of clay with Hanford-type radioactive wastes (liquids, salt cake, and sludge) were studied as a means of immobilization of radionuclides contained in the waste. Products of these reactions were identified as the crystalline sodium aluminosilicates, cancrinite and nepheline. Radionuclides are entrapped in these crystalline minerals. Conceptual flow diagrams for conversion of high-salt wastes to cancrinite and nepheline were defined and tested. The mineral products were evaluated for use as forms for long-term storage of radioactive waste

Ultrasonic testing of a sealing construction made of salt concrete in an underground disposal facility for radioactive waste

Energy Technology Data Exchange (ETDEWEB)

Krause, Martin; Effner, Ute Antonie; Milmann, Boris; Voelker, Christoph; Wiggenhauser, Herbert [Federal Institute for Materials Research and Testing (BAM), Berlin (Germany); Mauke, Ralf [The Federal Office for Radiation Protection, Salzgitter (Germany)

2015-07-01

For the closure of radioactive waste disposal facilities engineered barriers- so called ''drift seals'' are used. The purpose of these barriers is to constrain the possible infiltration of brine and to prevent the migration of radionuclides into the biosphere. In a rock salt mine a large scale in-situ experiment of a sealing construction made of salt concrete was set up to prove the technical feasibility and operability of such barriers. In order to investigate the integrity of this structure, non-destructive ultrasonic measurements were carried out. Therefore two different methods were applied at the front side of the test-barrier: 1 Reflection measurements from boreholes 2 Ultrasonic imaging by means of scanning ultrasonic echo methods This extended abstract is a short version of an article to be published in a special edition of ASCE Journal that will briefly describe the sealing construction, the application of the non-destructive ultrasonic measurement methods and their adaptation to the onsite conditions -as well as parts of the obtained results. From this a concept for the systematic investigation of possible contribution of ultrasonic methods for quality assurance of sealing structures may be deduced.

Modelling of the thermomechanical behaviour of salt rock

International Nuclear Information System (INIS)

Albers, G.; Graefe, V.; Korthaus, E.; Pudewillis, A.; Prij, J.

1986-01-01

The modelling of the thermomechanical behaviour of salt rock is examined, with respect to the disposal of radioactive waste in salt formations. The calculation methods and programmes currently available for the modelling are described. Some examples are given of calculations carried out in parallel with tests. Some results of modelling calculations for a repository are presented by way of illustration. (U.K.)

Thorium-based Molten Salt Reactor (TMSR) project in China

International Nuclear Information System (INIS)

Dai, Zhimin; Liu, Wei

2013-01-01

Making great efforts in development of nuclear energy is one of the long-term-plan in China's energy strategies. The advantages of Thorium-based nuclear energy are: rich resource in nature, less nuclear waste, low toxicity, nuclear non-proliferation and so on. Furthermore, China is a country with abundant thorium, thus it is necessary to develop the Thorium-based Molten Salt Reactor (TMSR) in China. Shanghai Institute of Applied Physics, Chinese Academy of Sciences (SINAP) had designed and constructed the first China's light-water reactor and developed a zero-power thorium-based molten salt reactor successfully in the early 1970s. The applied research project 'thorium molten salt reactor nuclear power system' by SINAP together with several other institutes had been accepted and granted by China government in 2011. The whole project has been divided into three stages: Firstly, built a 2 MW-zero-power high temperature solid molten salt reactor in 2015 and a 2 MW-zero-power high temperature liquid molten salt reactor in 2017. Secondly, in 2020 built a 10 MW high temperature liquid molten salt reactor. Thirdly, on the base of previous work, a 100 MW high temperature molten salt reactor should be achieving in 2030. After more than one years of efforts, a high quality scientific research team has been formed, which is able to design the molten salt reactor, the molten salt loop and related key equipment, the systems of molten salt preparation, purification and the radioactive gas removal. In the past one year, the initial physical design of high temperature molten salt reactor has been completed; the nuclear chemistry and radiation chemical laboratory has been built, a high temperature salt (HTS) loop and radioactive gas removal experiment device system have been successfully developed and constructed. Further, the preliminary study on reactor used carbon-carbon composite material has been investigated. (author)

Method for removing radioactive iodine and radioactive organic iodides from effluent gases

International Nuclear Information System (INIS)

1975-01-01

A method and composition for removing radioactive and organic iodides from an 131 I-containing off-gas stream is provided. The composition for removal by adsorption is a ceramic material with a surface area of from about 5 m 2 /g to about 250 m 2 /g impregnated with a metallic salt. The method for removing the iodine or iodide is accomplished by passing the off-gas stream over the ceramic material impregnated with the metallic salt. It finds special application in air filters for nuclear power plants

Final storage site for radioactive waste. Gorleben mine

International Nuclear Information System (INIS)

1995-02-01

Out of more than 20 salt stocks, the Gorleben salt stock was chosen. In addition to the preliminary information available on its size and depth, detailed exploratory investigations were carried out in order to test its suitability as a site for ultimate storage of all types of radioactive waste. (orig.) [de

Radiation induced F-center and colloid formation in synthetic NaCl and natural rock salt: applications to radioactive waste repositories

International Nuclear Information System (INIS)

Levy, P.W.; Loman, J.M.; Kierstead, J.A.

1983-01-01

Radiation damage, particularly Na metal colloid formation, has been studied in synthetic NaCl and natural rock salt using unique equipment for making optical absorption, luminescence and other measurements during irradiation with 1 to 3 MeV electrons. Previous studies have established the F-center and colloid growth phenomenology. At temperatures where colloids form most rapidly, 100 to 250 C, F-centers appear when the irradiation is initiated and increase at a decreasing rate to a plateau, reached at doses of 10 6 to 10 7 rad. Concomitant colloid growth is described by classical nucleation and growth curves with the transition to rapid growth occurring at 10 6 to 10 7 rad. The colloid growth rate is low at 100 C, increases markedly to a maximum at 150 to 175 C and decreases to a negligible rate at 225 C. At 1.2x10 8 rad/h the induction period is >10 4 sec at 100 C, 10 4 sec at 275 C. The colloid growth in salt from 14 localities is well described by C(dose)/sup n/ relations. Data on WIPP site salt (Los Medanos, NM, USA) has been used to estimate roughly the colloid expected in radioactive waste repositories. Doses of 1 to 2x10 10 rad, which will accumulate in salt adjacent to lightly shielded high level canisters in 200 to 500 years, will convert between 1 and 100% of the salt to Na colloids (and Cl) if back reactions or other limiting reactions do not occur. Each high level lightly shielded canister may ultimately be surrounded by 200 to 300 kg of colloid sodium. Low level or heavily shielded canisters may produce as little as 1 kg sodium

Solidification of liquid radioactive concentrates by fixation with cement

International Nuclear Information System (INIS)

Pekar, A.; Breza, M.; Timulak, J.; Krajc, T.

1985-01-01

In testing the technology of liquid radioactive wastes cementation, the effect was mainly studied of the content of boric acid and its salts on cement solidification, the effect of additives on radionuclide leachability and the effect of the salt content on the cementation product. On the basis of experimental work carried out on laboratory scale with model samples and samples of radioactive concentrate from the V-1 nuclear power plant, the following suitable composition of the cementation mixture was determined: 40% Portland cement, 40% zeolite containing material and 20% power plant ash. The most suitable ratio of liquid radioactive wastes and the cementation mixture is 0.5. As long as in such case the salt content of the concentrate ranges between 20 and 25%, the cementation product will have a maximum salt content of 10% and a leachability of the order of 10 -3 to 10 -4 g/cm 2 per day with a mechanical strength allowing safe handling. It was also found that the quality processing of the cement paste with degassing, e.g., by vibration, is more effective for the production of a pore-free cementation product than the application of various additives which are supposed to eliminate pore formation. (Z.M.)

Radioactivity of Consumer Products

Science.gov (United States)

Peterson, David; Jokisch, Derek; Fulmer, Philip

2006-11-01

A variety of consumer products and household items contain varying amounts of radioactivity. Examples of these items include: FiestaWare and similar glazed china, salt substitute, bananas, brazil nuts, lantern mantles, smoke detectors and depression glass. Many of these items contain natural sources of radioactivity such as Uranium, Thorium, Radium and Potassium. A few contain man-made sources like Americium. This presentation will detail the sources and relative radioactivity of these items (including demonstrations). Further, measurements of the isotopic ratios of Uranium-235 and Uranium-238 in several pieces of china will be compared to historical uses of natural and depleted Uranium. Finally, the presenters will discuss radiation safety as it pertains to the use of these items.

Salt repository design approach

International Nuclear Information System (INIS)

Matthews, S.C.

1983-01-01

This paper presents a summary discussion of the approaches that have been and will be taken in design of repository facilities for use with disposal of radioactive wastes in salt formations. Since specific sites have yet to be identified, the discussion is at a general level, supplemented with illustrative examples where appropriate. 5 references, 1 figure

Salt splitting with ceramic membranes

International Nuclear Information System (INIS)

Kurath, D.

1996-01-01

The purpose of this task is to develop ceramic membrane technologies for salt splitting of radioactively contaminated sodium salt solutions. This technology has the potential to reduce the low-level waste (LLW) disposal volume, the pH and sodium hydroxide content for subsequent processing steps, the sodium content of interstitial liquid in high-level waste (HLW) sludges, and provide sodium hydroxide free of aluminum for recycle within processing plants at the DOE complex. Potential deployment sites include Hanford, Savannah River, and Idaho National Engineering Laboratory (INEL). The technical approach consists of electrochemical separation of sodium ions from the salt solution using sodium (Na) Super Ion Conductors (NaSICON). As the name implies, sodium ions are transported rapidly through these ceramic crystals even at room temperatures

Radioactive waste isolation in salt: peer review of the D'Appolonia report on Schematic Designs for Penetration Seals for a Repository in the Permian Basin, Texas

International Nuclear Information System (INIS)

Hambley, D.F.; Stormont, J.C.; Russell, J.E.; Edgar, D.E.; Fenster, D.F.; Harrison, W.; Tisue, M.W.

1984-09-01

Argonne made the following recommedations for improving the reviewed reports. The authors of the report should: state the major assumptions of the study in Sec. 1.1 rather than later in the report; consider using salt for the shaft seals in salt horizons; reconsider whether keys are needed for the bulkheads; provide for interface grouting because use of expansive cement will not guarantee that interfaces will be impermeable; discuss the sealing schedule and, where appropriate, consider what needs to be done to ensure that emplaced radioactive waste could be retrieved if necessary; describe in more detail the sealing of the Dockum and Ogallala aquifers; consider an as low as reasonably achievable approach to performance requirements for the initial design phase; address the concerns in the 1983 US Nuclear Regulatory Commission document entitled Draft Technical Position: Borehole and Shaft Sealing of High-Level Nuclear Waste Repositories; cite the requirements for release of radioactivity by referring to specific clauses in the regulations of the US Environmental Protection Agency; and provide further explanation in the outline of future activities about materials development and verification testing. More emphasis on development of accelerated testing programs is also required

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.)

Landsat investigations of the northern Paradox basin, Utah and Colorado: implications for radioactive waste emplacement

Science.gov (United States)

Friedman, Jules D.; Simpson, Shirley L.

1978-01-01

present investigation for a potential radioactive waste-emplacement site in Salt Valley include confirmation of lack of permanent surface drainage and absence of agricultural or other development in the area of northern Salt Valley. On the other hand, the existence of diapirism, salt-karst landforms, and extensive lineamentation of the northern Paradox basin suggest regional tectonic instability at least in the geologic past. Future reactivation of diapiric or other halokinetic processes, including lateral flow, would lead to plastic behavior of the halite that might cause emplaced waste containers to migrate within the diapir. At Salt Valley, existing diapiric boundary faults and intersecting joint sets in sandstone units on the anticlinal flanks could, if the hydraulic gradient is suitable, provide conduits to the halite core for circulating ground water from adjacent Mesozoic sandstones in synclinal areas between the salt diapirs. Moreover, the loci of major lineament intersections might be areas of somewhat elevated seismic risk. If the salt barrier of Salt Valley anticline should fail in the future, potentially water-bearing Mesozoic fissile shales and friable to quartizitic sandstones would be the ultimate repository of the emplaced radioactive waste.

Experimental storage of high-level radioactive wastes in the Asse salt mine - technical aspects

International Nuclear Information System (INIS)

Gies, H.; Rothfuchs, T.; Feddersen, H.; Graefe, V.; Gross, S.; Hente, B.; Jockwer, N.; Kessels, W.; Schwaegermann, H.

1988-01-01

The work performed under this project in the Asse salt mine is an important milepost within the framework schedule of the 'Gorleben Poject' of Physikalisch-Technische Bundesanstalt (PTB). The project phase I (1982 - June 30, 1985) is about to be concluded at the time this report is published. The main points of interest of this project phase cover the planning of the experimental work, the design of experiments, and the first activities for developing the systems for handling the high-level radioactive wastes. The engineering development work has been advanced to the point where construction and manufacture of equipment can be started (transport containers Asse, TB1, collective transport containers, borehole gates, transport vehicles, waste positioning equipment, and borehole casing). Testing of the pipes for the last mentioned task with regard to the material's deformation behaviour will be done by the Dutch ECN as a sub-contractor. First laboratory experiments have been carried out on radiolysis gas formation, to complement the engineering work and the in-situ measuring programmes. (orig./RB) [de

Thermomechanical behaviour of salt rock. Project part 1

International Nuclear Information System (INIS)

Albrecht, H.; Hunsche, U.; Diekmann, N.; Ludwig, R.

1991-08-01

The present final report on the research project KWA 58019, part I, gives an overview of the research done from early in 1988 till mid-1991 in section B 2.13 of the Federal Office of Geosciences and Raw Materials, in the field of salt mechanics. This report contributes to the scientific foundations for dimensioning and safety analysis of a repository for radioactive wastes in a salt dome and for underground exploration of a salt dome. It covers the activities financed both by the research project and by earmarked funds. (orig.) [de

Organic waste processing using molten salt oxidation

Energy Technology Data Exchange (ETDEWEB)

Adamson, M. G., LLNL

1998-03-01

Molten Salt Oxidation (MSO) is a thermal means of oxidizing (destroying) the organic constituents of mixed wastes, hazardous wastes, and energetic materials while retaining inorganic and radioactive constituents in the salt. For this reason, MSO is considered a promising alternative to incineration for the treatment of a variety of organic wastes. The U. S. Department of Energy`s Office of Environmental Management (DOE/EM) is currently funding research that will identify alternatives to incineration for the treatment of organic-based mixed wastes. (Mixed wastes are defined as waste streams which have both hazardous and radioactive properties.) One such project is Lawrence Livermore National Laboratory`s Expedited Technology Demonstration of Molten Salt Oxidation (MSO). The goal of this project is to conduct an integrated demonstration of MSO, including off-gas and spent salt treatment, and the preparation of robust solid final forms. Livermore National Laboratory (LLNL) has constructed an integrated pilot-scale MSO treatment system in which tests and demonstrations are presently being performed under carefully controlled (experimental) conditions. The system consists of a MSO process vessel with dedicated off-gas treatment, a salt recycle system, feed preparation equipment, and equipment for preparing ceramic final waste forms. In this paper we describe the integrated system and discuss its capabilities as well as preliminary process demonstration data. A primary purpose of these demonstrations is to identify the most suitable waste streams and waste types for MSO treatment.

Preparation and Characterization of Novel Cationic Chitosan Derivatives Bearing Quaternary Ammonium and Phosphonium Salts and Assessment of Their Antifungal Properties.

Science.gov (United States)

Tan, Wenqiang; Li, Qing; Dong, Fang; Chen, Qiuhong; Guo, Zhanyong

2017-08-31

Chitosan is an abundant and renewable polysaccharide, its derivatives exhibit attractive bioactivities and the wide applications in various biomedical fields. In this paper, two novel cationic chitosan derivatives modified with quaternary phosphonium salts were successfully synthesized via trimethylation, chloride acetylation, and quaternization with tricyclohexylphosphine and triphenylphosphine. The structures and properties of synthesized products in the reactions were characterized by FTIR spectroscopy, ¹H-NMR, 31 P-NMR, elemental and thermogravimetric analysis. The antifungal activities of chitosan derivatives against four kinds of phytopathogens, including Phomopsis asparagi , Watermelon fusarium , Colletotrichum lagenarium , and Fusarium oxysporum were tested using the radial growth assay in vitro. The results revealed that the synthesized cationic chitosan derivatives showed significantly improved antifungal efficiency compared to chitosan. It was reasonably suggested that quaternary phosphonium groups enabled the obviously stronger antifungal activity of the synthesized chitosans. Especially, the triphenylphosphonium-functionalized chitosan derivative inhibited the growth of Phomopsis asparagi most effectively, with inhibitory indices of about 80% at 0.5 mg/mL. Moreover, the data demonstrated that the substituted groups with stronger electron-withdrawing ability relatively possessed greater antifungal activity. The results suggest the possibility that cationic chitosan derivatives bearing quaternary phosphonium salts could be effectively employed as novel antifungal biomaterials for application in the field of agriculture.

SAVANNAH RIVER SITE INCIPIENT SLUDGE MIXING IN RADIOACTIVE LIQUID WASTE STORAGE TANKS DURING SALT SOLUTION BLENDING

Energy Technology Data Exchange (ETDEWEB)

Leishear, R.; Poirier, M.; Lee, S.; Steeper, T.; Fowley, M.; Parkinson, K.

2011-01-12

This paper is the second in a series of four publications to document ongoing pilot scale testing and computational fluid dynamics (CFD) modeling of mixing processes in 85 foot diameter, 1.3 million gallon, radioactive liquid waste, storage tanks at Savannah River Site (SRS). Homogeneous blending of salt solutions is required in waste tanks. Settled solids (i.e., sludge) are required to remain undisturbed on the bottom of waste tanks during blending. Suspension of sludge during blending may potentially release radiolytically generated hydrogen trapped in the sludge, which is a safety concern. The first paper (Leishear, et. al. [1]) presented pilot scale blending experiments of miscible fluids to provide initial design requirements for a full scale blending pump. Scaling techniques for an 8 foot diameter pilot scale tank were also justified in that work. This second paper describes the overall reasons to perform tests, and documents pilot scale experiments performed to investigate disturbance of sludge, using non-radioactive sludge simulants. A third paper will document pilot scale CFD modeling for comparison to experimental pilot scale test results for both blending tests and sludge disturbance tests. That paper will also describe full scale CFD results. The final paper will document additional blending test results for stratified layers in salt solutions, scale up techniques, final full scale pump design recommendations, and operational recommendations. Specifically, this paper documents a series of pilot scale tests, where sludge simulant disturbance due to a blending pump or transfer pump are investigated. A principle design requirement for a blending pump is UoD, where Uo is the pump discharge nozzle velocity, and D is the nozzle diameter. Pilot scale test results showed that sludge was undisturbed below UoD = 0.47 ft{sup 2}/s, and that below UoD = 0.58 ft{sup 2}/s minimal sludge disturbance was observed. If sludge is minimally disturbed, hydrogen will not be

Hygroscopic salts and the potential for life on Mars.

Science.gov (United States)

Davila, Alfonso F; Duport, Luis Gago; Melchiorri, Riccardo; Jänchen, Jochen; Valea, Sergio; de Los Rios, Asunción; Fairén, Alberto G; Möhlmann, Diedrich; McKay, Christopher P; Ascaso, Carmen; Wierzchos, Jacek

2010-01-01

Hygroscopic salts have been detected in soils in the northern latitudes of Mars, and widespread chloride-bearing evaporitic deposits have been detected in the southern highlands. The deliquescence of hygroscopic minerals such as chloride salts could provide a local and transient source of liquid water that would be available for microorganisms on the surface. This is known to occur in the Atacama Desert, where massive halite evaporites have become a habitat for photosynthetic and heterotrophic microorganisms that take advantage of the deliquescence of the salt at certain relative humidity (RH) levels. We modeled the climate conditions (RH and temperature) in a region on Mars with chloride-bearing evaporites, and modeled the evolution of the water activity (a(w)) of the deliquescence solutions of three possible chloride salts (sodium chloride, calcium chloride, and magnesium chloride) as a function of temperature. We also studied the water absorption properties of the same salts as a function of RH. Our climate model results show that the RH in the region with chloride-bearing deposits on Mars often reaches the deliquescence points of all three salts, and the temperature reaches levels above their eutectic points seasonally, in the course of a martian year. The a(w) of the deliquescence solutions increases with decreasing temperature due mainly to the precipitation of unstable phases, which removes ions from the solution. The deliquescence of sodium chloride results in transient solutions with a(w) compatible with growth of terrestrial microorganisms down to 252 K, whereas for calcium chloride and magnesium chloride it results in solutions with a(w) below the known limits for growth at all temperatures. However, taking the limits of a(w) used to define special regions on Mars, the deliquescence of calcium chloride deposits would allow for the propagation of terrestrial microorganisms at temperatures between 265 and 253 K, and for metabolic activity (no growth) at

Spanish participation in the Haw Project: Laboratory investigations on Gamma irradiation effects in rock salt

International Nuclear Information System (INIS)

Cuevas, C. de las; Miralles, L.; Teixidor, P.; Garcia Veigas, J.; Dies, X.; Ortega, X.; Pueyo, J.J.

1993-01-01

In order to prove the safe disposal of high-level radioactive waste (HAW) in salt rock, a five years test disposal of thirty highly radioactive radiation sources is planned in the Asse salt mine, in the Federal Republic of Germany. The thirty radiation sources consist of steel canisters containing the vitrified radionuclides Caesium 137 and Strontium 90 in quantities sufficient to cover the bandwidth of heat generation and gamma radiation of real HAW. The radiation sources will be emplaced in six boreholes located in two galleries at the 800 m level. Two electrical heater tests were already started in November 1988 and are continuosly surveyed in respect of the rock mass. Also the handling system necessary for the emplacement of the radioactive canisters was developed and succesfully tested. A laboratory investigation programme on radiation effects in salt is being performed in advance to the radioactive canister emplacement. This programme includes the investigation of thermally and radiolytically induced water and gas release from the rock salt and the radiolytical decomposition of salt minerals. Part of this programme has been carried out since 1988 at the University of Barcelona, basically what refers to colloidal sodium determinations by light absorption measurements and microstructural studies on irradiated salt samples. For gamma dose and dose rate measurements in the test field, measuring systems consisting of ionisation chambers as well as solid state dosemeters were developed and tested. Thermomechanical computer code validation is performed by calculational predictions and parallel investigation of the stress and displacement fields in the underground test field

Brine flow in heated geologic salt.

Energy Technology Data Exchange (ETDEWEB)

Kuhlman, Kristopher L.; Malama, Bwalya

2013-03-01

This report is a summary of the physical processes, primary governing equations, solution approaches, and historic testing related to brine migration in geologic salt. Although most information presented in this report is not new, we synthesize a large amount of material scattered across dozens of laboratory reports, journal papers, conference proceedings, and textbooks. We present a mathematical description of the governing brine flow mechanisms in geologic salt. We outline the general coupled thermal, multi-phase hydrologic, and mechanical processes. We derive these processes governing equations, which can be used to predict brine flow. These equations are valid under a wide variety of conditions applicable to radioactive waste disposal in rooms and boreholes excavated into geologic salt.

Discussions about safety criteria and guidelines for radioactive waste management.

Science.gov (United States)

Yamamoto, Masafumi

2011-07-01

In Japan, the clearance levels for uranium-bearing waste have been established by the Nuclear Safety Commission (NSC). The criteria for uranium-bearing waste disposal are also necessary; however, the NSC has not concluded the discussion on this subject. Meanwhile, the General Administrative Group of the Radiation Council has concluded the revision of its former recommendation 'Regulatory exemption dose for radioactive solid waste disposal', the dose criteria after the institutional control period for a repository. The Standardization Committee on Radiation Protection in the Japan Health Physics Society (The Committee) also has developed the relevant safety criteria and guidelines for existing exposure situations, which are potentially applicable to uranium-bearing waste disposal. A new working group established by The Committee was initially aimed at developing criteria and guidelines specifically for uranium-bearing waste disposal; however, the aim has been shifted to broader criteria applicable to any radioactive wastes.

Site specific study for possible ongoing salt dome movement

International Nuclear Information System (INIS)

Thoms, R.L.; Manning, T.A.; Paille, L.K.; Gehle, R.M.

1977-01-01

U.S. Gulf Coast salt domes, among other geologic structures, currently are being considered for storage of commercial radioactive wastes. A major concern with dome storage of long lived radioactive wastes lies with the possible tectonic movement of the host dome. Any ongoing movement of a salt dome can be monitored with a site specific complementary system of field instrumentation and finite element modelling. Field instrumentation and accompanying finite element analyses for a study dome in northwest Louisiana are described. Site specific data and early experience associated with tiltmeters over the dome are presented. Also, recommendations are made for modifications and extensions of the field instrumentation and finite element modelling appropriate to the specific site under study

Bituminization of liquid radioactive waste. Part 3

International Nuclear Information System (INIS)

G'oshev, G.S.; Gradev, G.D.; Stefanova, I.G.; Milusheva, A.G.; Guteva, E.S.; Stefanov, G.I.

1991-01-01

The elaborated technology for bituminization of liquid radioactive wastes (salt concentrates) is characterized by the fact that the bituminization process takes place in two stages: concentration of the liquid residue and evaporation of the water with simultaneous homogeneous incorporation of the salts in the melted bitumen. An experimental installation for bituminization of salt concentrates was designed on the basis of this technology. The experience accumulated during the design and construction of the installation for bituminization of salt concentrates could be used for designing and constructing an industrial installation for bituminization of the liquid residue of the nuclear power plants. 2 tabs., 3 figs., 3 refs

Rheological stratification of the Hormuz Salt Formation in Iran - microstructural study of the dirty and pure rock salts from the Kuh-e-Namak (Dashti) salt diapir

Science.gov (United States)

Závada, Prokop; Desbois, Guillaume; Urai, Janos; Schulmann, Karel; Rahmati, Mahmoud; Lexa, Ondrej; Wollenberg, Uwe

2014-05-01

Significant viscosity contrasts displayed in flow structures of a mountain namakier (Kuh-e-Namak - Dashti), between 'weak' terrestrial debris bearing rock salt types and 'strong' pure rock salt types are questioned for deformation mechanisms using detailed quantitative microstructural study including crystallographic preferred orientation (CPO) mapping of halite grains. While the solid impurity rich ("dirty") rock salts contain disaggregated siltstone and dolomite interlayers, "clean" salts (debris free) reveal microscopic hematite and remnants of abundant fluid inclusions in non-recrystallized cores of porphyroclasts. Although flow in both, the recrystallized dirty and clean salt types is accommodated by combined mechanisms of pressure-solution creep (PS), grain boundary sliding (GBS) and dislocation creep accommodated grain boundary migration (GBM), their viscosity contrasts are explained by significantly slower rates of intergranular diffusion and piling up of dislocations at hematite inclusions in clean salt types. Porphyroclasts of clean salts deform by semi-brittle and plastic mechanisms with intra-crystalline damage being induced also by fluid inclusions that explode in the crystals at high fluid pressures. Boudins of clean salt types with coarse grained and original sedimentary microstructure suggest that clean rock salts are associated with dislocation creep dominated power law flow in the source layer and the diapiric stem. Rheological contrasts between both rock salt classes apply in general for the variegated and terrestrial debris rich ("dirty") Lower Hormuz and the "clean" rock salt forming the Upper Hormuz, respectively, and suggest that large strain rate gradients likely exist along horizons of mobilized salt types of different composition and microstructure.

The influence of salt aerosol on alpha radiation detection by WIPP continuous air monitors

Energy Technology Data Exchange (ETDEWEB)

Bartlett, W.T.; Walker, B.A. [Environmental Evaluation Group, Albuquerque, NM (United States)

1997-08-01

Waste Isolation Pilot Plant (WIPP) alpha continuous air monitor (CAM) performance was evaluated to determine if CAMs could detect accidental releases of transuranic radioactivity from the underground repository. Anomalous alpha spectra and poor background subtraction were observed and attributed to salt deposits on the CAM sampling filters. Microscopic examination of salt laden sampling filters revealed that aerosol particles were forming dendritic structures on the surface of the sampling filters. Alpha CAM detection efficiency decreased exponentially as salt deposits increased on the sampling filters, suggesting that sampling-filter salt was performing like a fibrous filter rather than a membrane filter. Aerosol particles appeared to penetrate the sampling-filter salt deposits and alpha particle energy was reduced. These findings indicate that alpha CAMs may not be able to detect acute releases of radioactivity, and consequently CAMs are not used as part of the WIPP dynamic confinement system. 12 refs., 12 figs., 1 tab.

Geology and salt deposits of the Michigan Basin

International Nuclear Information System (INIS)

Johnson, K.S.; Gonzales, S.

1976-07-01

The Silurian-age Salina salt, one of the greatest deposits of bedded rock salt in the world, underlies most of the Michigan basin and parts of the Appalachian basin in Ohio. Pennsylvania, New York, and West Virginia. Interest in this salt deposit has increased in recent years because there may be one or more areas where it could be used safely as a repository for the underground storage of high-level radioactive wastes. The general geology of the Michigan basin is summarized and the major salt deposits are described in the hope that these data will be useful in determining whether there are any areas in the basin that are sufficiently promising to warrant further detailed study. Distribution of the important salt deposits in the basin is limited to the Southern Peninsula of Michigan

Treating agent for urea containing radioactive materials

International Nuclear Information System (INIS)

Ogawa, Hiroshi; Maki, Kentaro.

1973-01-01

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

Assessment of a Salt Reduction Intervention on Adult Population Salt Intake in Fiji

Directory of Open Access Journals (Sweden)

Arti Pillay

2017-12-01

Full Text Available Reducing population salt intake is a global public health priority due to the potential to save lives and reduce the burden on the healthcare system through decreased blood pressure. This implementation science research project set out to measure salt consumption patterns and to assess the impact of a complex, multi-faceted intervention to reduce population salt intake in Fiji between 2012 and 2016. The intervention combined initiatives to engage food businesses to reduce salt in foods and meals with targeted consumer behavior change programs. There were 169 participants at baseline (response rate 28.2% and 272 at 20 months (response rate 22.4%. The mean salt intake from 24-h urine samples was estimated to be 11.7 grams per day (g/d at baseline and 10.3 g/d after 20 months (difference: −1.4 g/day, 95% CI −3.1 to 0.3, p = 0.115. Sub-analysis showed a statistically significant reduction in female salt intake in the Central Division but no differential impact in relation to age or ethnicity. Whilst the low response rate means it is not possible to draw firm conclusions about these changes, the population salt intake in Fiji, at 10.3 g/day, is still twice the World Health Organization’s (WHO recommended maximum intake. This project also assessed iodine intake levels in women of child-bearing age and found that they were within recommended guidelines. Existing policies and programs to reduce salt intake and prevent iodine deficiency need to be maintained or strengthened. Monitoring to assess changes in salt intake and to ensure that iodine levels remain adequate should be built into future surveys.

The influence of salt aerosol on alpha radiation detection by WIPP continuous air monitors

International Nuclear Information System (INIS)

Bartlett, W.T.; Walker, B.A.

1996-01-01

Alpha continuous air monitors (CAMs) will be used at the Waste Isolation Pilot Plant (WIPP) to measure airborne transuranic radioactivity that might be present in air exhaust or in work-place areas. WIPP CAMs are important to health and safety because they are used to alert workers to airborne radioactivity, to actuate air-effluent filtration systems, and to detect airborne radioactivity so that the radioactivity can be confined in a limited area. In 1993, the Environmental Evaluation Group (EEG) reported that CAM operational performance was affected by salt aerosol, and subsequently, the WIPP CAM design and usage were modified. In this report, operational data and current theories on aerosol collection were reviewed to determine CAM quantitative performance limitations. Since 1993, the overall CAM performance appears to have improved, but anomalous alpha spectra are present when sampling-filter salt deposits are at normal to high levels. This report shows that sampling-filter salt deposits directly affect radon-thoron daughter alpha spectra and overall monitor efficiency. Previously it was assumed that aerosol was mechanically collected on the surface of CAM sampling filters, but this review suggests that electrostatic and other particle collection mechanisms are more important than previously thought. The mechanism of sampling-filter particle collection is critical to measurement of acute releases of radioactivity. 41 refs

Treatment method for stabilization of radioactive exchange resin

International Nuclear Information System (INIS)

Hideo, Oni; Takashi, Miyake; Hitoshi, Miyamoto; Toshio, Funakoshi; Yuzo, Inagaki.

1988-01-01

This is a method for eluting radioactive nuclides from a radioactive ion exchange resin in which it has been absorbed. First, the Cs in this resin is extracted using a neutral salt solution which contains Na + . The Cs that has been transferred to the neutral salt solution is absorbed and expelled by inorganic ion exchangers. Then the Co, Fe, Mn and Sr in said resin are eluted using an acidic solution; the Co, Fe, Mn and Sr that have been transferred to the acidic solution are separated from that solution by means of a diffusion dialysis vat. This process is a unique characteristic of this ion exchange resin treatment method. 1 fig

Radioactive influence of some phosphogypsum piles located at the SW Spain in their surrounding soils and salt-marshes

Science.gov (United States)

Bolivar, J. P.; Mosqueda, F.; Vaca, F.; Garcia-Tenorio, R.; Martinez-Sanchez, M. J.; Perez-Sirvent, C.; Martinez-Lopez, S.

2012-04-01

In the SW of Spain, just in the confluence of the mouths of the Tinto and Odiel River and in the vicinity of Huelva town, there is a big industrial complex which includes between others an industry devoted during more than 40 years to the production of phosphoric acid, by treating sedimentary phosphate rock by the so-called "wet acid method". As a by-product of the mentioned process it have been produced historically huge amounts of a compound called phosphogypsum, which composition is mostly di-hydrate calcium sulphate containing some of the impurities of heavy metals and natural radionuclides originally present in the raw material. Due to the lack of market for this by-product, it has been mostly piled over some salt-marshes located in the vicinity of the industry, on the bank of the Tinto River. About 100 million tons of phosphogypsum have been piled in an area covering more than 1000 hectares, constituting a clear environmental and radiological anomaly in the zone. The phosphogypsum piles set do not conform obviously a close system. They are interacting with the nearby environment mostly by leaching waters releases from the waters accumulated in them either for its previous use in transporting in suspension the PG from the factory or by rainfall. These waters leaks contain in solution enhanced amounts of heavy metals and radionuclides that can provoke the chemical and radioactive contamination in surroundings soil and salt-marshes areas. In this communication the radioactive influence by the phosphogypsum piles in the surrounding terrestrial environment is evaluated. This contamination is mostly due to radionuclides belonging to the uranium series, which are present originally in the raw material treated in the industry, and afterwards in the generated phosphogypsum, in enhanced amounts in relation to typical soils. In addition, the different dynamics and behavior of different radionuclides will be discussed and analyzed. The gained information in this study

Technetium removal column flow testing with alkaline, high salt, radioactive tank waste

International Nuclear Information System (INIS)

Blanchard, D.L. Jr.; Kurath, D.E.; Golcar, G.R.; Conradson, S.D.

1996-01-01

This report describes two bench-scale column tests conducted to demonstrate the removal of Tc-99 from actual alkaline high salt radioactive waste. The waste used as feed for these tests was obtained from the Hanford double shell tank AW-101, which contains double shell slurry feed (DSSF). The tank sample was diluted to approximately 5 M Na with water, and most of the Cs-137 was removed using crystalline silicotitanates. The tests were conducted with two small columns connected in series, containing, 10 mL of either a sorbent, ABEC 5000 (Eichrom Industries, Inc.), or an anion exchanger Reillex trademark-HPQ (Reilly Industries, Inc.). Both materials are selective for pertechnetate anion (TcO 4 - ). The process steps generally followed those expected in a full-scale process and included (1) resin conditioning, (2) loading, (3) caustic wash to remove residual feed and prevent the precipitation of Al(OH) 3 , and (4) elution. A small amount of Tc-99m tracer was added as ammonium pertechnetate to the feed and a portable GEA counter was used to closely monitor the process. Analyses of the Tc-99 in the waste was performed using ICP-MS with spot checks using radiochemical analysis. Technetium x-ray absorption spectroscopy (XAS) spectra of 6 samples were also collected to determine the prevalence of non-pertechnetate species [e.g. Tc(IV)

Deep storage of radioactive waste from a geological point of view

International Nuclear Information System (INIS)

Venzlaff, Helmut

2015-01-01

For a deep storage of radioactive waste geologists gave their preference to salt prior to other rock complexes such as clay or granite. Major deposits from pure rock salt are particularly suitable to safely seal radioactive wastes from the biosphere because due to their plasticity they are free from fissures in which liquids and gases could circulate and because their thermal conductivity is higher than of other rocks. The geological stability of salt domes can be shown by their geological evolution. Thus the salt dome in Gorleben was formed 100 million years ago and is older than the Atlantic, the Alps or the ascent of the low mountain range. During this long period it survived ocean floods, mountain formations, earthquakes, volcanism and ice ages without considerably changing its shape. There are no geological reasons, why it should not remain stable during the next million years.

Deep storage of radioactive waste from a geological point of view

Energy Technology Data Exchange (ETDEWEB)

Venzlaff, Helmut [Federal Institute for Geo-Sciences and Raw Materials, Hannover (Germany)

2015-08-15

For a deep storage of radioactive waste geologists gave their preference to salt prior to other rock complexes such as clay or granite. Major deposits from pure rock salt are particularly suitable to safely seal radioactive wastes from the biosphere because due to their plasticity they are free from fissures in which liquids and gases could circulate and because their thermal conductivity is higher than of other rocks. The geological stability of salt domes can be shown by their geological evolution. Thus the salt dome in Gorleben was formed 100 million years ago and is older than the Atlantic, the Alps or the ascent of the low mountain range. During this long period it survived ocean floods, mountain formations, earthquakes, volcanism and ice ages without considerably changing its shape. There are no geological reasons, why it should not remain stable during the next million years.

Geologic investigation of the Virgin River Valley salt deposits, Clark County, southeastern Nevada, to investigate their suitability for possible storage of radioactive waste material as of September 1977

International Nuclear Information System (INIS)

1977-01-01

The results from a geologic investigation of the Virgin River Valley salt deposits, Clark County, southeastern Nevada, to examine their suitability for further study and consideration in connection with the possible storage of radioactive waste material are given. The results indicate that (1) approximately one-half of the salt body underlies the Overton Arm of Lake Mead and that the dry land portion of the salt body that has a thickness of 1,000 feet or more covers an area of about four and one-half square miles; (2) current tectonic activity in the area of the salt deposits is believed to be confined to seismic events associated with crustal adjustments following the filling of Lake Mead; (3) detailed information on the hydrology of the salt deposit area is not available at present but it is reported that a groundwater study by the U.S. Geological Survey is now in progress; (4) there is no evidence of exploitable minerals in the salt deposit area other than evaporites such as salt, gypsum, and possibly sand and gravel; (5) the salt deposit area is located inside the Lake Mead Recreation Area, outlined on the accompanying Location Plat, and several Federal, State, and Local agencies share regulatory responsibilities for the activities in the area; (6) other salt deposit areas of Arizona and Nevada, such as the Detrital Valley, Red Lake Dome, Luke Dome, and Mormon Mesa area, and several playa lake areas of central Nevada may merit further study; and (7) additional information, as outlined, is needed to more thoroughly evaluate the salt deposits of the Virgin River Valley and other areas referred to above

Decontaminated salt disposal as saltcrete in a landfill. Technical data summary

International Nuclear Information System (INIS)

1982-01-01

This technical data summary presents a reference process for immobilizing decontaminated salt solution from the 200-Area waste storage tanks with cement, and disposing of the final waste material (called saltcrete) by burial in trenches. The saltcrete will be protected from leaching by clay and will be placed at least 3 meters above the historic high water table and beneath at least 5 meters of soil overburden. The decontaminated salt solution is a waste material which remains after the bulk of the radionuclides have been removed from waste tank supernate. This removal is effected by contacting the waste supernate with sodium tetraphenyl boron (Na-TPB) and sodium titanate (NaTi 2 O 5 H). These materials remove (by precipitation) most of the 137 Cs and 90 Sr as well as many other radioactive and non-radioactive elements. These precipitates, along with many other sludges which reside in the HLW tanks will be incorporated in borosilicate glass for eventual disposal in a geologic repository. An ion exchange process will also be used for removal of 99 Tc. The decontaminated salt solution has sufficiently low levels of radioactivity that it can be disposed of on-site. The scope of the curent effort is to describe a process for blending decontaminated salt solution with cement to form a saltcrete product which has dimensional stability and relatively low leachability. The process is to be capable of solidifying 10 gpm of supernate. About 100 million gallons of salt solution is to be solidified

Natural radioactivity measurements and dose calculations to the public: Case of the uranium-bearing region of Poli in Cameroon

International Nuclear Information System (INIS)

Saidou; Bochud, Francois O.; Baechler, Sebastien; Moise, Kwato Njock; Merlin, Ngachin; Froidevaux, Pascal

2011-01-01

The objective of this work is to carry out a baseline study of the uranium-bearing region of Poli in which lies the uranium deposit of Kitongo, prior to its impending exploitation. This study required sampling soil, water and foodstuffs representative of the radioactivity exposure and food consumption patterns of the population of Poli. After sampling and radioactivity measurements were taken, our results indicated that the activities of natural series in soil and water samples are low. However, high levels of 210 Po and 210 Pb in foodstuffs (vegetables) were discovered and elevated activities of 40 K were observed in some soil samples. All components of the total dose were assessed and lead to an average value of 5.2 mSv/year, slightly higher than the average worldwide value of 2.4 mSv/year. Most of this dose is attributable to the ingestion dose caused by the high levels of 210 Po and 210 Pb contained in vegetables, food items which constitute an important part of the diet in Northern Cameroon. Consequently, bringing uranium ore from underground to the surface might lead to an increased dose for the population of Poli through a higher deposition of 222 Rn decay products on leafy vegetables.

Petrofabric changes in heated and irradiated salt from Project Salt Vault, Lyons, Kansas

International Nuclear Information System (INIS)

Holdoway, K.A.

1972-01-01

Rock salt was heated and irradiated in situ by implanted radioactive wastes during the Project Salt Vault experiment which was carried out at Lyons, Kansas, in the abandoned Carey Salt mine between 1965 and 1967. It was found that irradiation results in coloration of the salt, producing colors ranging from blue-black nearest the radiation source, to pale blue and purple farther from the source. Bleached areas are common in the radiation-colored salt, many representing trails produced by the migration of fluid inclusions towards the heat source. These visible trails are thought to have formed during the cooling down of the salt after the removal of the heaters and radiation sources. The distribution of primary structures in the salt suggests that little migration, if any, occurred during the course of the experiment. It is proposed that radiolysis of the brine within the inclusions may have led to the production of gases which impeded or prevented migration. Evidence of strain was observed in slip planes at 4 in. (10 cm) and between 5.5 and 10 in. (13.5 to 25.4 cm) from the array hole. Deformed bleached areas in the salt between the areas were slip planes are developed suggest that slight plastic deformation or flow may have occurred at 6 in. (15 cm) from the array hole. Differential thermal analysis shows that the maximum amount of stored energy also occurs at 6 in. (15 cm) from the array hole. This region may therefore represent the zone where the combined effect of stress and radiation was greatest

Cementation of liquid radioactive waste with high content of borate salts

International Nuclear Information System (INIS)

Gorbunova, O.

2015-01-01

The report reviews the ways of optimization of cementation of boron-containing liquid radioactive waste. The most common way to hardening the low-level liquid radioactive waste (LRW) is the cementation. However, boron-containing liquid radioactive waste with low pH values cannot be cemented without alkaline additives, to neutralize acid forms of borate compounds. Cement setting without additives happens only on 14-56 days, the compounds have low strength, and hence an insufficient reliability of radionuclides fixation in the cement matrix. The alkaline additives increase the volume of the final cement compound which enhances financial and operational costs. In order to control the speed of hardening of cement solution with a boron-containing liquid radioactive waste and to remove the components that prevent hardening of cement solution, it is proposed an electromagnetic treatment of LRW in the vortex layer of ferromagnetic particles. The results of infrared spectroscopy show, that electromagnetic treatment of liquid radioactive waste changes the ionic forms of the borates and raises the pH due to the dissociation of the oxygen and hydrogen bonds in the aqueous solutions of the boron compounds. The various types of ferromagnetic activators of the vortex layer have been investigated, including the highly dispersed nano-powders and the magnetic phases of the iron oxides. It has been determined the technological parameters of the electromagnetic treatment of liquid radioactive waste and the subsequent cementation of this type of LRW. By using the method of scanning electron microscopy it has been shown, that the nano-particles of magnetic phases of the ferric oxides are involved in phase formation of hydro-aluminum-calcium ferrites in the early stages of hardening and improving strength of the cement compounds with liquid radioactive waste. (authors)

Radiological consequences associated with human intrusion into radioactive waste repositories in salt formations

International Nuclear Information System (INIS)

Jacquier, P.

1989-01-01

The assessment of the radiological impact of human intrusion scenarios is extremely important in the case of repositories located in salt formations, since salt is obviously a valuable economic resource. Salt formations also represent a suitable medium for mining storage caverns for oil and gas. The scenario considered in this report is that of solution mining in salt formations to produce salt for human consumption. It is postulated that the salt is extracted by excavating a cavern through solution-mining and that, in the course of cavern enlargement, the waste is intercepted and drops to the bottom of the cavern. We have assumed that the intrusion takes place 500 or even 2 500 years after the repository has been sealed. The cases considered involve high-level vitrified waste or cemented alpha waste. The paper describes the assumptions on which the scenario is based and uses a simplified model to assess the radiological consequences associated with the ingestion of contaminated salt. The paper also provides details of a sensitivity/uncertainty analysis which identified several areas in which experimental studies should be either initiated or continued [fr

Method of removing radioactive waste from oil

International Nuclear Information System (INIS)

Belanger, R.L.

1986-01-01

This patent describes a method of removing particulates, radioactive contaminants, and moisture from oil, which consists of: straining out the particulates by passing the oil through a coarse filter screen to a receiving vessel; forming an upper stratum of oil and a lower stratum of sludge, consisting of mud, oil, particulates, and moisture, by heating the upper two-thirds of the receiving vessel; skimming off the stratum of oil from the receiving vessel; transferring the sludge from the receiving vessel to a container; transferring additional separated oil to the receiving vessel; conveying the oil skimmed from the receiving vessel to a mixing vessel; adding an effective amount of Calcium Hypochlorite crystals containing 65% free Chlorine to the mixing vessel to initiate salt formation with the radioactive contaminants; mixing the contents of the mixing vessel for at least ten minutes; transferring the mixture from the mixing vessel to a circulating heater; outputting the mixture from the circulating heater to a second mixing vessel; removing moisture from the oil; and filtering from the oil, the solid radioactive contaminant-salts and residual particulate matter

Effect of nitrite concentration on pit depth in carbon steel exposed to simulated radioactive waste

International Nuclear Information System (INIS)

Zapp, P.E.

1997-01-01

The growth of pits in carbon steel exposed to dilute (0.055 M nitrate-bearing) alkaline salt solutions that simulate radioactive waste was investigated in coupon immersion tests. Most coupons were tested in the as-received condition, with the remainder having been heat treated to produce an oxide film. Nitrite, which is an established pitting inhibitor in these solutions, was present in concentrations from 0 to 0.031 M to 0.16 M; the last concentration is known to prevent pitting initiation in the test solution at the 50 degrees C test temperature. The depths of the deepest pits on coupons of particular exposure conditions were measure microscopically and were analyzed as simple, type 1 extreme value statistical distributions, to predict the deepest expected pit in a radioactive waste tank subject to the test conditions. While the growth rate of pits could not be established from these tests, the absolute value of the deepest pits predicted is of the order of 100 mils after 448 days of exposure. The data indicate that even nitrite concentrations insufficient to prevent pitting have a beneficial effect on limiting the growth of deepest pits

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

International Nuclear Information System (INIS)

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

1986-01-01

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

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

International Nuclear Information System (INIS)

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

1980-01-01

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

Method to synthesize dense crystallized sodalite pellet for immobilizing halide salt radioactive waste

International Nuclear Information System (INIS)

Koyama, Tadafumi.

1994-01-01

A method is described for immobilizing waste chloride salts containing radionuclides such as cesium and strontium and hazardous materials such as barium. A sodalite intermediate is prepared by mixing appropriate amounts of silica, alumina and sodium hydroxide with respect to sodalite and heating the mixture to form the sodalite intermediate and water. Heating is continued to drive off the water to form a water-free intermediate. The water-free intermediate is mixed with either waste salt or waste salt which has been contacted with zeolite to concentrate the radionuclides and hazardous material. The waste salt-intermediate mixture is then compacted and heated under conditions of heat and pressure to form sodalite with the waste salt, radionuclides and hazardous material trapped within the sodalite cage structure. This provides a final product having excellent leach resistant capabilities

Salt Repository Project Waste Package Program Plan: Draft

International Nuclear Information System (INIS)

Carr, J.A.; Cunnane, J.C.

1986-01-01

Under the direction of the Office of Civilian Radioactive Waste Management (OCRWM) created within the DOE by direction of the Nuclear Waste Policy Act of 1982 (NWPA), the mission of the Salt Repository Project (SRP) is to provide for the development of a candidate salt repository for disposal of high-level radioactive waste (HLW) and spent reactor fuel in a manner that fully protects the health and safety of the public and the quality of the environment. In consideration of the program needs and requirements discussed above, the SRP has decided to develop and issue this SRP Waste Package Program Plan. This document is intended to outline how the SRP plans to develop the waste package design and to show, with reasonable assurance, that the developed design will satisfy applicable requirements/performance objectives. 44 refs., 16 figs., 16 tabs

Recent research and development of bearings for helium circulator

International Nuclear Information System (INIS)

Taniguchi, S.; Ezaki, Z.; Kawaguchi, K.; Matsumura, N.; Kozima, M.

1988-01-01

This paper mainly describes recent studies and successful applications of water lubricated bearing and gas lubricated bearing. Both types of bearing seem to be suitable for a turbo machine installed in an atomic energy plant - such as the helium circulator of a HTGR - not to be affected by radioactivity, so we have been attracted by them for about 10 years. The former was investigated theoretically taking account of turbulent flow due to the low viscosity of water, and compared with the experimental data. Good agreement was obtained, and a successful example applied to a small-sized high speed air compressor is shown. The latter was investigated using a large-sized bearing test rig simulated to an actual machine. The tilting pad journal bearing and the tilting pad thrust bearing were taken and improved for some aspects. These bearings have been taken into service on an actual circulator and are now operating successfully. Currently, a magnetic bearing is being studied to pay special attention to endurance for an earthquake and catcher bearing system. We would like to have an opportunity to present these results in the near future. (author). 5 refs, 15 figs, 2 tabs

Method of decomposing radioactive organic solvent wastes

International Nuclear Information System (INIS)

Uki, Kazuo; Ichihashi, Toshio; Hasegawa, Akira; Sato, Tatsuaki

1986-01-01

Purpose: To decompose radioactive organic solvent wastes or radioactive hydrocarbon solvents separated therefrom into organic materials under moderate conditions, as well as greatly decrease the amount of secondary wastes generated. Method: Radioactive organic solvent wastes comprising an organic phosphoric acid ester ingredient and a hydrocarbon ingredient as a diluent therefor, or radioactive hydrocarbon solvents separated therefrom are oxidatively decomposed by hydrogen peroxide in an aqueous phosphoric acid solution of phosphoric acid metal salts finally into organic materials to perform decomposing treatment for the radioactive organic solvent wastes. The decomposing reaction is carried out under relatively moderate conditions and cause less burden to facilities or the likes. Further, since the decomposed liquid after the treatment can be reused for the decomposing reaction as a catalyst solution secondary wastes can significantly be decreased. (Yoshihara, H.)

40 CFR Appendix B to Part 414 - Complexed Metal-Bearing Waste Streams

Science.gov (United States)

2010-07-01

... 414—Complexed Metal-Bearing Waste Streams Chromium Azo dye intermediates/Substituted diazonium salts + coupling compounds Vat dyes Acid dyes Azo dyes, metallized/Azo dye + metal acetate Acid dyes, Azo...

Minimization of radioactive material deposition in water-cooled nuclear reactors

International Nuclear Information System (INIS)

Ruiz, C.P.; Blaies, D.M.

1988-01-01

This patent describes the method for inhibiting the deposition of radioactive cobalt in a water-bearing vessel of a water-cooled nuclear reactor which comprises adding zinc ion to water entering the water-bearing vessel. The improvement contains a substantially lower proportion of the /sup 64/Zn isotope than naturally occurring zinc

High Salt Intake Attenuates Breast Cancer Metastasis to Lung.

Science.gov (United States)

Xu, Yijuan; Wang, Wenzhe; Wang, Minmin; Liu, Xuejiao; Lee, Mee-Hyun; Wang, Mingfu; Zhang, Hao; Li, Haitao; Chen, Wei

2018-04-04

Diet-related factors are thought to modify the risk of cancers, while the influence of high salt intake remains largely uncharacterized. Breast cancer is the most common cancer in women worldwide. In the present study, we examined the effect of salt intake on breast cancer by using a 4T1 mouse mammary tumor model. Unexpectedly, both the fitness and the survival rate of the tumor-bearing mice were improved by high salt intake. Similarly, high salt intake suppressed the primary tumor growth as well as metastasis to lung in mice. Mechanistically, high salt intake greatly reduced food intake and thus might exert antitumor effect through mimicking calorie restriction. Immunoblotting showed the lower proliferation marker Ki-67 and the higher expression of the tumor suppressor gene p53 in tumors of high salt intake mice. Importantly, high salt intake might induce hyperosmotic stress, which sensitized breast cancer cells to p53-dependent anoikis. Collectively, our findings raise the possibility that endogenous salt deposition might act as the first-line defense system against breast cancer progression as well as metastasis.

SALT4: a two-dimensional displacement discontinuity code for thermomechanical analysis in bedded salt deposits

International Nuclear Information System (INIS)

1983-04-01

SALT4 is a two-dimensional analytical/displacement-discontinuity code designed to evaluate temperatures, deformation, and stresses associated with underground disposal of radioactive waste in bedded salt. This code was developed by the University of Minnesota. This documentation describes the mathematical equations of the physical system being modeled, the numerical techniques utilized, and the organization of the computer code, SALT4. The SALT4 code takes into account: (1) viscoelastic behavior in the pillars adjacent to excavations; (2) transversely isotropic elastic moduli such as those exhibited by bedded or stratified rock; and (2) excavation sequence. Major advantages of the SALT4 code are: (1) computational efficiency; (2) the small amount of input data required; and (3) a creep law consistent with laboratory experimental data for salt. The main disadvantage is that some of the assumptions in the formulation of SALT4, i.e., temperature-independent material properties, render it unsuitable for canister-scale analysis or analysis of lateral deformation of the pillars. The SALT4 code can be used for parameter sensitivity analyses of two-dimensional, repository-scale, thermal and thermomechanical response in bedded salt during the excavation, operational, and post-closure phases. It is especially useful in evaluating alternative patterns and sequences of excavation or waste canister placement. SALT4 can also be used to verify fully numerical codes. This is similar to the use of analytic solutions for code verification. Although SALT4 was designed for analysis of bedded salt, it is also applicable to crystalline rock if the creep calculation is suppressed. In Section 1.5 of this document the code custodianship and control is described along with the status of verification, validation and peer review of this report

Integrated demonstration of molten salt oxidation with salt recycle for mixed waste treatment

International Nuclear Information System (INIS)

Hsu, P.C.

1997-01-01

Molten Salt Oxidation (MSO) is a thermal, nonflame process that has the inherent capability of completely destroying organic constituents of mixed wastes, hazardous wastes, and energetic materials while retaining inorganic and radioactive constituents in the salt. For this reason, MSO is considered a promising alternative to incineration for the treatment of a variety of organic wastes. Lawrence Livermore National Laboratory (LLNL) has prepared a facility and constructed an integrated pilot-scale MSO treatment system in which tests and demonstrations are performed under carefully controlled (experimental) conditions. The system consists of a MSO processor with dedicated off-gas treatment, a salt recycle system, feed preparation equipment, and equipment for preparing ceramic final waste forms. This integrated system was designed and engineered based on laboratory experience with a smaller engineering-scale reactor unit and extensive laboratory development on salt recycle and final forms preparation. In this paper we present design and engineering details of the system and discuss its capabilities as well as preliminary process demonstration data. A primary purpose of these demonstrations is identification of the most suitable waste streams and waste types for MSO treatment

Using of clay-salt slimes of 'Belaruskali' factory as a sorbents of radionuclide

International Nuclear Information System (INIS)

Maskalchuk, L.

2010-01-01

Document available in extended abstract form only. The effective sorbents for decrease of radionuclide migration in soil and prevention of pollution risk of soil and underground water by radionuclide, according to available practical experience on minimization of consequences of the Chernobyl NPP accident, are: clay minerals of layered structure of type 2:1, potash fertilizers for 137 Cs and potassium rocks for 90 Sr. The analysis of literary data shows, those there two base kinds of industrial waste are formed at sylvinite ore processing almost at all potassium plants of the world: - Solid halite rejects material - Liquid waste in the form of clay-salt slimes. There is about 9 % of halite waste from annual formation using in Belarus, clay-salt slimes (CSS) are not used in general and all the volume goes to slime storage. Clay-salt slimes are the waste products of potassium production being formed in the course of sylvinite ore conversion at processing plants of the Industrial centre 'Belaruskali'. Up to the present moment about 80 millions of tones of clay-salt slimes have been accumulated in Soligorsk industrial zone, and their annual formation makes up about 2.0-2.5 millions of tones. The volume of industrial waste collected in Republic of Belarus allows considering CSS as a possible source of low-cost raw material for reception of products with different functions. On the other hand by estimation of national and international experts such quantity of industrial waste, especially liquid, represents ecological danger. Taking into account this circumstance the situation with industrial waste disposal in Soligorsk industrial area of Belarus which was estimated by international experts as critical one and it needs the cardinal measures for further environment pollution prevention. There is considerable volume of liquid radioactive waste is formed at the Nuclear Power Plant operation. Modern tendencies of radioactive waste disposal are directed on

Radioactive wastes: sources, treatment, and disposal

International Nuclear Information System (INIS)

Wymer, R.G.; Blomeke, J.O.

1975-01-01

Sources, treatment, and disposal of radioactive wastes are analyzed in an attempt to place a consideration of the problem of permanent disposal at the level of established or easily attainable technology. In addition to citing the natural radioactivity present in the biosphere, the radioactive waste generated at each phase of the fuel cycle (mills, fabrication plants, reactors, reprocessing plants) is evaluated. The three treatment processes discussed are preliminary storage to permit decay of the short-lived radioisotopes, solidification of aqueous wastes, and partitioning the long-lived α emitters for separate and long-term storage. Dispersion of radioactive gases to the atmosphere is already being done, and storage in geologically stable structures such as salt mines is under active study. The transmutation of high-level wastes appears feasible in principle, but exceedingly difficult to develop

Technological study about a disposal measures of low-level radioactive waste including uranium and long-half-life radionuclides

International Nuclear Information System (INIS)

Sugaya, Toshikatsu; Nakatani, Takayoshi; Sakai, Akihiro; Sakamoto, Yoshiaki; Sasaki, Toshihisa; Nakamura, Yasuo

2017-02-01

Japan Atomic Energy Agency (JAEA) performed the technical studies contributed for the disposal measures of uranium-bearing waste with low concentration and intermediate depth disposal-based waste occurring from the process of the nuclear fuel cycle. (1) Study of the trench disposal of uranium-bearing waste. As a part of the study of disposal measures of the uranium-bearing waste, we carried out the safety assessment (exposure dose assessment) and derived the upper limit of radioactivity concentration of uranium which was allowed to be included in radioactive waste for trench disposal. (2) Preliminary study for the expansion of material applied to clearance in uranium-bearing waste. Currently, the clearance level of uranium handling facilities was derived from the radioactivity concentration of uranium corresponding to dose criterion about the exposure pathways of the reuse and recycle of metal. Therefore, we preliminarily evaluated whether metal and concrete were able to be applied to clearance by the method of the undergrounding disposal. (3) Study of the concentration limitation scenarios for the intermediate depth disposal-based waste. We carried out dose assessment of intermediate depth disposal of radioactive waste generated from JAEA about radioactive concentration limitation scenarios of which the concept was shown by the study team in Nuclear Regulation Authority. Based on the results, we discussed whether the waste was applied to radioactive waste conforming to concept of intermediate depth disposal. (author)

Hydrological methods preferentially recover cesium from nuclear waste salt cake

International Nuclear Information System (INIS)

Brooke, J.N.; Hamm, L.L.

1997-01-01

The Savannah River Site is treating high level radioactive waste in the form of insoluble solids (sludge), crystallized salt (salt cake), and salt solutions. High costs and operational concerns have prompted DOE to look for ways to improve the salt cake treatment process. A numerical model was developed to evaluate the feasibility of pump and treat technology for extracting cesium from salt cake. A modified version of the VAM3DCG code was used to first establish a steady-state flow field, then to simulate 30 days of operation. Simulation results suggest that efficient cesium extraction can be obtained with low displacement volumes. The actual extraction process will probably be less impressive because of nonuniform properties. 2 refs., 2 figs

The treatment and conditioning of transuranelement bearing wastes in the Federal Republic of Germany

International Nuclear Information System (INIS)

Krause, H.

1986-01-01

Transuranelement bearing wastes (TRU wastes) differ from other radioactive wastes (with the exception of high level wastes from reprocessing) primarily by the longevity and high radiotoxity of many of their radionuclides. The volumes and total TRU content of these wastes are still quite small. Due to the present absence of a repository for radioactive wastes in the FRG, no definitions of TRU wastes and no acceptance criteria for these wastes have been fixed so far. Anyway, as only waste disposal into deep geological formations is envisaged for the time being, the limits for the TRU content do not need to be as low as in countries practicing shallow land burial. During the experimental disposal in the Asse salt mine, wastes with a TRU-content <5μCi/g were considered as non-TRU waste. There is some probability that in the future a similar value may be fixed. The present practice in TRU waste management is primarily determined by this situation. However, this system is neither ideal from a fundamental point of view nor in the long range; and, therefore, research and development work is going on for the development of an advanced TRU waste management system which should meet the requirements of an industrial scale fast breeder fuel cycle, and also improve the acceptance of such a program by the public. (Auth.)

Method for treating radioactive liquids

International Nuclear Information System (INIS)

Komrow, R.R.; Pritchard, J.F.

1980-01-01

A process for treating and handling radioactive liquids and rendering such liquids safe for handling is disclosed. Transportation and disposal, the process comprises adding thereto a small amount of a water-insoluble alkali salt of an aqueous alkali saponified gelatinized-starch-polyacrylonitrile graft polymer, to form a solid, semi-solid or gel product

Radioactive mineral deposits

Energy Technology Data Exchange (ETDEWEB)

1948-01-01

This publication was designed as a guide for uranium and thorium prospectors in Australia. Physical properties, such as color, streak, luster, hardness, fracture, and specific gravity of the uranium and thorium-bearing minerals are summarized and the various methods suitable for detecting radioactivity in minerals are described. Two colored plates show samples of pitchblende (uraninite), autunite, carnotite, monazite, and others of the most important minerals sources of uranium and thorium.

Geohydrology of the Keechi, Mount Sylvan, Oakwood, and Palestine salt domes in the northeast Texas salt-dome basin

International Nuclear Information System (INIS)

Carr, J.E.; Halasz, S.J.; Peters, H.B.

1980-01-01

The salt within these domes has penetrated as much as 20,000 feet of Mesozoic and Cenozoic strata, and presently extends to within 120 to 800 feet of the land surface. The salt penetrates or closely underlies major freshwater and salinewater aquifers within the basin. To provide a safe repository for radioactive wastes within one or more of these domes, a thorough understanding of the geohydrology needs to be obtained, and the hydrologic stability of the domes needs to be established for the expected life of the storage facility. Dissolution may exist at all four candidate salt domes, possibly through contact with Cretaceous or Tertiary aquifers, or through fault systems in the vicinity of the domes. Strata overlying and surrounding Palestine and Keechi Salt Domes have been arched into steeply-dipping folds that are complexly faulted. Similar conditions exist at Oakwood and Mount Sylvan Domes, except that the Tertiary strata have been only moderately disturbed. Additional problems concerning the hydrologic stability of Oakwood and Palestine Salt Domes have resulted from the disposal of oil-field salinewater in the cap rock at the Oakwood Dome and previous solution mining of salt at the Palestine Dome

Some aspects of the development of NW-German salt domes

International Nuclear Information System (INIS)

Jaritz, W.

1980-01-01

Aspects of the development of salt structures that may be of some importance to the safety of a final disposal site for radioactive waste are salt ascent and salt dissolution at the surface. The geological history of the salt domes is described in terms of the dissolution of the salt at the dome surface. In many cases it can be distinguished whether dissolution was caused by the ascent of the salt into strata containing groundwater by diapirism or by epeirogenic uplift or both. The salt domes of Wesendorf, Heide, and Marne are used as examples in a discussion of the transition from dissolution to the deposition of a cover of impermeable sediments. Moreover, the development of the Gorleben salt dome is described. The author's studies show the average rate of uplift of the NW-German salt domes in the diapiric stage to have ranged from a little less than 0.1 to about 0.5 mm per year. For salt domes in later stages, the rate of uplift is several hundredths of a millimeter per year at most. (orig.) [de

Method and apparatus for nuclear heating of oil-bearing formations

International Nuclear Information System (INIS)

Alspaw, D.I.

1979-01-01

A method and apparatus are provided for using heat generated by absorption of radiation from nuclear waste materials to reduce the viscosity of petroleum products contained within a subsurface earth formation. The nuclear waste material is positioned in a salt water formation underlying the subsurface earth formation so that the radiation emitted by the material heats the salt water formation. conduction and convection transfer the heat to the subsurface earth formation, raising the temperature and thereby reducing the viscosity of the petroleum products. To prevent radioactive contamination within the salt water formation, the nuclear waste material may be encapsulated in a material selected to absorb alpha and beta radiation

Radiation effects in rock salt. A status report

International Nuclear Information System (INIS)

Gies, H.; Hild, W.; Kuehle, T.; Moenig, J.

1994-01-01

Knowledge of the irradiation defects and the accompanying energy storage in rock salt resulting from the absorption of ionizing radiation emitted by vitrified high level radioactive waste (HLW) disposed off in geological rock salt formations in an important prerequisite for a realistic assessment of possible consequences. Based on a critical review of the scientific status this report attempts to evaluate whether the available database is satisfactory and sufficiently reliable for the performance of such an assessment. Apart from a brief description of the radiation-and temperature-conditions prevailing in a HLW-repository, a detailed presentation is given of both the interaction of radiation with rock salt and the theories and models developed for their quantification

Geohydrolic studies of Gulf Coast interior salt domes

International Nuclear Information System (INIS)

Smith, C.G. Jr.

1977-01-01

Disposal of high-level radioactive wastes in Gulf Coast salt domes requires that the cavities be free from groundwater dissolution for 250,000 years. Salinity variations of groundwater near selected domes were investigated. Saline groundwater anomalies (saline plumes) in aquifers pierced or uplifted by the dome may be the result of salt solution by groundwater. In the Northeast Texas salt dome basin electric logs of oil and gas wells have been used to estimate groundwater salinities in aquifers near selected domes. Thus far, the analyses have revealed saline groundwater anomalies around 4 of the 9 domes studied. Estimates of the rate of salt dissolution from domes associated with saline groundwater plumes indicate that less than 30 meters of salt will be removed from the upper surfaces of the dome in 250,000 years. Thus, these preliminary studies show that even apparently unstable domes may be sufficiently stable to serve as waste disposal sites. 6 figures

Molten salt reactor type

International Nuclear Information System (INIS)

1977-01-01

This document is one of the three parts of a first volume devoted to the compilations of American data on the molten salt reactor concept. This part 'CIRCUITS' regroups under a condensed form - in French and using international units - the essential information contained in both basic documents of the American project for a molten-salt breeder power plant. This part is only dealing with things relating to the CEA-EDF workshop 'CIRCUITS'. It is not concerned with information on: the reactor and the moderator replacement, the primary and secondary salts, and the fuel salt reprocessing, that are dealt with in parts 'CORE' and 'CHEMISTRY' respectively. The possible evolutions in the data - and solutions - taken by the American designers for their successive projects (1970 to 1972) are shown. The MSBR power plant comprises three successive heat transfer circuits. The primary circuit (Hastelloy N), radioactive and polluted, containing the fuel salt, includes the reactor, pumps and exchangers. The secondary circuit (pipings made of modified Hastelloy N) contaminated in the exchanger, ensures the separation between the fuel and the fluid operating the turbo-alternator. The water-steam circuit feeds the turbine with steam. This steam is produced in the steam generator flowed by the secondary fluid. Some subsidiary circuits (discharge and storage of the primary and secondary salts, ventilation of the primary circuit ...) complete the three principal circuits which are briefly described. All circuits are enclosed inside the controlled-atmosphere building of the nuclear boiler. This building also ensures the biological protection and the mechanical protection against outer aggressions [fr

Alternatives for definse waste-salt disposal

International Nuclear Information System (INIS)

Benjamin, R.W.; McDonell, W.R.

1983-01-01

Alternatives for disposal of decontaminated high-level waste salt at Savannah River were reviewed to estimate costs and potential environmental impact for several processes. In this review, the reference process utilizing intermediate-depth burial of salt-concrete (saltcrete) monoliths was compared with alternatives including land application of the decontaminated salt as fertilizer for SRP pine stands, ocean disposal with and without containment, and terminal storage as saltcake in existing SRP waste tanks. Discounted total costs for the reference process and its modifications were in the same range as those for most of the alternative processes; uncontained ocean disposal with truck transport to Savannah River barges and storage as saltcake in SRP tanks had lower costs, but presented other difficulties. Environmental impacts could generally be maintained within acceptable limits for all processes except retention of saltcake in waste tanks, which could result in chemical contamination of surrounding areas on tank collapse. Land application would require additional salt decontamination to meet radioactive waste disposal standards, and ocean disposal without containment is not permitted in existing US practice. The reference process was judged to be the only salt disposal option studied which would meet all current requirements at an acceptable cost

Performance assessment of geological isolation systems for radioactive waste. Disposal in salt formations

International Nuclear Information System (INIS)

Storck, R.; Aschenbach, J.; Hirsekom, R.P.; Nies, A.; Stelte, N.

1988-01-01

In the framework of the PAGIS project of the CEC Research Programme on radioactive waste, a performance assessment of a repository of vitrified HLW in rock salt formations has been carried out. The first volume of the study is split into four tasks. Task 1 recalls the main steps that have led to the selection of the reference and the variant site. Task 2 condenses all information available on the rock formations which are planned to host the repository, the overlying geosphere and the geohistoric development of the sites. Task 3 states the technical details of repository planning, while in Task 4 conceivable release scenarios are discussed. Volume II (Tasks 5 to 10) is concerned with the modelling procedures. In Task 5 data for the waste inventory are collected and the selection of relevant nuclides for transport calculations is discussed. Task 6 gives the near-field modelling, i.e. the models for corrosion of the waste canisters, the degradation of the waste matrix and the models used for the HLW boreholes. Task 7 deals with the modelling of the repository. Its division into sections is discussed and models for physical and chemical effects taken into account in each section are presented. In Task 8 the modelling of the overburden is given. In Task 9 additional models for the subrosion scenario and a human intrusion scenario are given. Task 10 is concerned with the biosphere modelling. In Volume III results of deterministic and probabilistic calculations are presented. Task 11 gives the results for deterministic calculations with best estimate values for the parameters involved in the models. Task 12 presents the result of the uncertainty analysis, and Task 13 those of local and global sensitivity analyses followed by concluding remarks. This document is one of a set of 5 reports covering a relevant project of the European Community on a nuclear safety subject having very wide interest. The five volumes are: the summary (EUR 11775-EN), the clay (EUR 11776-EN), the

Treatment of waste salt from the advanced spent fuel conditioning process (I): characterization of Zeolite A in Molten LiCl Salt

International Nuclear Information System (INIS)

Kim, Jeong Guk; Lee, Jae Hee; Yoo, Jae Hyung; Kim, Joon Hyung

2004-01-01

The oxide fuel reduction process based on the electrochemical method (Advanced spent fuel Conditioning Process; ACP) and the long-lived radioactive nuclides partitioning process based on electro-refining process, which are being developed ay the Korea Atomic Energy Research Institute (KAERI), are to generate two types of molten salt wastes such as LiCl salt and LiCl-KCl eutectic salt, respectively. These waste salts must meet some criteria for disposal. A conditioning process for LiCl salt waste from ACP has been developed using zeolite A. This treatment process of waste salt using zeolite A was first developed by US ANL (Argonne National Laboratory) for LiCl-KCl eutectic salt waste from an electro-refining process of EBR (Experimental Breeder Reactor)-II spent fuel. This process has been developed recently, and a ceramic waste form (CWF) is produced in demonstration-scale V-mixer (50 kg/batch). However, ANL process is different from KAERI treatment process in waste salt, the former is LiCl-KCl eutectic salt and the latter is LiCl salt. Because of melting point, the immobilization of eutectic salt is carried out at about 770 K, whereas LiCl salt at around 920 K. Such difference has an effect on properties of immobilization media, zeolite A. Here, zeolite A in high-temperature (923 K) molten LiCl salt was characterized by XRD, Ion-exchange, etc., and evaluated if a promising media or not

Radioactive waste disposal in deep geologic deposits. Associated research problems

International Nuclear Information System (INIS)

Rousset, G.

1992-01-01

This paper describes the research associated problems for radioactive waste disposal in deep geologic deposits such granites, clays or salt deposits. After a brief description of the underground disposal, the author studies the rheology of sedimentary media and proposes rheological models applied to radioactive wastes repositories. Waste-rock interactions, particularly thermal effects and temperature distribution versus time. 17 refs., 14 figs

Preparation and Identification of HER2 Radioactive Ligands and Imaging Study of Breast Cancer-Bearing Nude Mice

Directory of Open Access Journals (Sweden)

Meng-zhi Zhang

2017-08-01

Full Text Available OBJECTIVE: A micro-molecule peptide TP1623 of 99mTc-human epithelial growth factor receptor 2 (HER2 was prepared and the feasibility of using it as a HER2-positive molecular imaging agent for breast cancer was evaluated. METHODS: TP1623 was chemically synthesized and labeled with 99mTc. The labeling ratio and stability were detected. HER2 expression levels of breast cancer cells (SKBR3 and MDA-MB-231 and cell binding activity were measured. Biodistribution of 99mTC-TP1623 in normal mice was detected. SKBR3/MDA-MB-231-bearing nude mice models with high/low expressions of HER2 were established. Tumor tissues were stained with hematoxylin–eosin (HE and measured by immunohistochemistry to confirm the formation of tumors and HER2 expression. SPECT imaging was conducted for HER2-overexpressing SKBR3-bearing nude mice. The T/NT ratio was calculated and compared with that of MDA-MB-231-bearing nude mice with low HER2 expression. The competitive inhibition image was used to discuss the specific binding of 99mTc- TP1623 and the tumor. RESULTS: The labeling ratio of 99mTc-TP1623, specific activity, and radiochemical purity (RCP after 6 h at room temperature were (97.39 ± 0.23%, (24.61 ± 0.06 TBq/mmol, and (93.25 ± 0.06%, respectively. HER2 of SKBR3 and MDA-MB-231 cells showed high and low expression levels by immunohistochemistry, respectively. The in vitro receptor assays indicated that specific binding of TP1623 and HER2 was retained. Radioactivity in the brain was always at the lowest level, while the clearance rate of blood and the excretion rate of the kidneys were fast. HE staining showed that tumor cells were observed in SKBR3- and MDA-MB-231-bearing nude mice, with significant heteromorphism and increased mitotic count. The imaging of mice showed that targeted images could be made of 99mTc-TP1623 in high HER2-expressing tumors, while no obvious development was shown in tumors in low HER2-expressing nude mice. No development was visible in

Geology and geomorphology of Bear Lake Valley and upper Bear River, Utah and Idaho

Science.gov (United States)

Reheis, M.C.; Laabs, B.J.C.; Kaufman, D.S.

2009-01-01

Bear Lake, on the Idaho-Utah border, lies in a fault-bounded valley through which the Bear River flows en route to the Great Salt Lake. Surficial deposits in the Bear Lake drainage basin provide a geologic context for interpretation of cores from Bear Lake deposits. In addition to groundwater discharge, Bear Lake received water and sediment from its own small drainage basin and sometimes from the Bear River and its glaciated headwaters. The lake basin interacts with the river in complex ways that are modulated by climatically induced lake-level changes, by the distribution of active Quaternary faults, and by the migration of the river across its fluvial fan north of the present lake. The upper Bear River flows northward for ???150 km from its headwaters in the northwestern Uinta Mountains, generally following the strike of regional Laramide and late Cenozoic structures. These structures likely also control the flow paths of groundwater that feeds Bear Lake, and groundwater-fed streams are the largest source of water when the lake is isolated from the Bear River. The present configuration of the Bear River with respect to Bear Lake Valley may not have been established until the late Pliocene. The absence of Uinta Range-derived quartzites in fluvial gravel on the crest of the Bear Lake Plateau east of Bear Lake suggests that the present headwaters were not part of the drainage basin in the late Tertiary. Newly mapped glacial deposits in the Bear River Range west of Bear Lake indicate several advances of valley glaciers that were probably coeval with glaciations in the Uinta Mountains. Much of the meltwater from these glaciers may have reached Bear Lake via groundwater pathways through infiltration in the karst terrain of the Bear River Range. At times during the Pleistocene, the Bear River flowed into Bear Lake and water level rose to the valley threshold at Nounan narrows. This threshold has been modified by aggradation, downcutting, and tectonics. Maximum lake

Geophysical Well-Log Measurements in Three Drill Holes at Salt Valley, Utah

OpenAIRE

Daniels, Jeffrey J.; Hite, Robert J.; Scott, James H.; U.S. Geological Survey

1980-01-01

Three exploratory drill holes were drilled at Salt Valley, Utah, to study the geologic, physical, geochemical, and hydrologic properties of the evaporite sequence in the Permian Paradox Member of the Hermosa Formation. The results of these studies will be used to help to determine the suitability of salt deposits in the Paradox basin as a storage medium for radioactive waste material.

Defense waste salt disposal at the Savannah River Plant

International Nuclear Information System (INIS)

Langton, C.A.; Dukes, M.D.

1984-01-01

A cement-based waste form, saltstone, has been designed for disposal of Savannah River Plant low-level radioactive salt waste. The disposal process includes emplacing the saltstone in engineered trenches above the water table but below grade at SRP. Design of the waste form and disposal system limits the concentration of salts and radionuclides in the groundwater so that EPA drinking water standards will not be exceeded at the perimeter of the disposal site. 10 references, 4 figures, 3 tables

Seismic-refraction survey to the top of salt in the north end of the Salt Valley Anticline, Grand County, Utah

Science.gov (United States)

Ackermann, Hans D.

1979-01-01

A seismic-refraction survey, consisting of three lines about 2700, 2760, and 5460 meters long, was made at the north end of the Salt Valley anticline of the Paradox Basin in eastern Utah. The target was the crest of a diapiric salt mass and the overlying, deformed caprock. The interpretations reveal an undulating salt surface with as much as 80 meters of relief. The minimum depth of about 165 meters is near the location of three holes drilled by the U.S. Department of Energy for the purpose of evaluating the Salt Valley anticline as a potential site for radioactive waste storages Caprock properties were difficult to estimate because the contorted nature of these beds invalidated a geologic interpretation in terms of velocity layers. However, laterally varying velocities of the critically refracted rays throughout the area suggest differences in the gross physical properties of the caprock.

Seismic-refraction survey to the top of salt in the north end of the Salt valley anticline, Grand County, Utah

International Nuclear Information System (INIS)

Achermann, H.D.

1979-01-01

A sesimic-refraction survey, consisting of three lines about 2700, 2760, and 5460 meters long, was made at the north end of the Salt valley anticline of the Paradox Basin in eastern Utah. The target was the crest of a diapiric salt mass and the overlying, deformed caprock. The interpretations reveal an undulating salt surface with as much as 80 meters of relief. The minimum depth of about 165 meters is near the location of three holes drilled by the US Department of Energy for the purpose of evaluating the Salt Valley anticline as a potential site for radioactive waste storage. Caprock properties were difficult to estimate because the contorted nature of these beds invalidated a goelogic interpretation in terms of velocity layers. However, laterally varying velocities of the critically refracted rays throughout the area suggest differences in the gross physical properties of the caprock

Radioactive methionine: determination, and distribution of radioactivity in the sulfur, methyl and 4-carbon moieties

International Nuclear Information System (INIS)

Giovanelli, J.; Mudd, S.H.

1985-01-01

A simple and inexpensive method is described for isolation and determination of [ 14 C]methionine in the non-protein fraction of tissues extensively labeled with 14 C. The effectiveness of the method was demonstrated by isolation of non-protein [ 14 C]methionine (as the carboxymethylsulfonium salt) of proven radiopurity from the plant Lemna which had been grown for a number of generations on (U- 14 C]sucrose and contained a 2000-fold excess of 14 C in undefined non-protein compounds. An advantage is that the isolated methioninecarboxymethlysulfonium salt is readily degraded to permit separate determination of radioactivity in the 4-carbon, methyl and sulfur moieties of methionine. During this work, a facile labilization of 3 H attached to the (carboxy)methylene carbon of methioninecarboxymethylsulfonium salt was observed. This labilization is ascribed to formation of a sulfur ylid. (Auth.)

Purple Salt and Tiny Drops of Water in Meteorites

Science.gov (United States)

Taylor, G. J.

1999-12-01

Some meteorites, especially those called carbonaceous chondrites, have been greatly affected by reaction with water on the asteroids in which they formed. These reactions, which took place during the first 10 million years of the Solar System's history, formed assorted water-bearing minerals, but nobody has found any of the water that caused the alteration. Nobody, that is, until now. Michael Zolensky and team of scientists from the Johnson Space Center in Houston and Virginia Tech (Blacksburg, Virginia) discovered strikingly purple sodium chloride (table salt) crystals in two meteorites. The salt contains tiny droplets of salt water (with some other elements dissolved in it). The salt is as old as the Solar System, so the water trapped inside the salt is also ancient. It might give us clues to the nature of the water that so pervasively altered carbonaceous chondrites and formed oceans on Europa and perhaps other icy satellites. However, how the salt got into the two meteorites and how it trapped the water remains a mystery - at least for now.

Hydrogeological problems in the ultimate storage of radioactive wastes

International Nuclear Information System (INIS)

Uerpmann, E.P.

1980-01-01

The following work shows how one can achieve the safe closure of ultimate-stored radioactive wastes by connecting a series of various barriers to the biosphere. The propagation of radionuclides by ground water is considered to be the most important long-term transport mechanism. Salt occurences in the Federal Republic of Germany are considered to be the best form suitable for end storage formations for known reasons. When not observing mining and hydrogeological knowledge, the danger of uncontrollable water flow in the end storage can arise from the water solubility of the salt rocks. Therefore the filling of salt mines and the subsequent procedures are dealt with in detail. The leading of radioactive nuclides is influenced by the properties of the ultimately stored wastes and by the quality of the remaining filling of the caves. These problems are dealt with in detail. A series of barriers to the closure of the underground caves are suggested and discussed. The most important barriers consist of the stability of the corresponding selected end storage structure. Possible arrangements of the storage cave are given which even after storage must maintain a high stability. Proposals are made on how the ultimately stored wastes can protect themselves against contact with free water or salt solutions. (orig.) [de

Technical issues in the geologic disposal of radioactive wastes

International Nuclear Information System (INIS)

Weart, W.D.

1980-01-01

The status of technical understanding regarding radioactive waste repositories in geologic media is improving at a rapid rate. Within a few years the knowledge regarding non-salt repositories will be on a par with that which now exists for salt. To date there is no technical reason to doubt that geologic repositories in several different geologic media can be safely implemented to provide long-term isolation of radioactive wastes. Indeed, for bedded salt, there is now sufficient knowledge to allow all the identified phenomena to be bounded with satisfactory resultant consequences. It is possible to now proceed with technical confidence in an orderly development of a bedded-salt repository at a satisfactory site. This development would call for in-situ experiments, at the earliest possible stage, to confirm or validate the predictions made for the site. These in-situ experiments will be necessary for each repository in a different rock type. If, for non-technical reasons, repository development is delayed, field test facilities should be located as soon as possible in geologic settings typical of proposed repositories. Extensive testing to resolve generic issues will allow subsequent development of repositories to proceed more rapidly with only minimal in-situ testing required to resolve site-specific concerns

Sudden venting test of an emergency bearing for the magnet bearing type compound molecular pump

International Nuclear Information System (INIS)

Hiroki, Seiji; Abe, Tetsuya; Murakami, Yoshio; Okamoto, Masatomo; Iguchi, Masashi; Nakamura, Jyunichi; Nakazeki, Tsugito.

1995-01-01

The vacuum evacuation system for nuclear fusion reactors bears the role of exhausting hydrogen isotopes in large quantity together with helium continuously for long hours, and as the high vacuum pumps for this purpose, the mechanical pumps which can do continuous evacuation and decrease the quantity of staying radioactive tritium, such as turbo molecular pumps and compound molecular pumps, are promising. Because of the compatibility with tritium, oil lubrication is not desirable, accordingly, the pumps with ceramic rotating vanes and magnetic bearings are demanded. As a part of the development of a magnetic bearing type mechanical pump which can be used for nuclear fusion reactors, the compound molecular pump, in which emergency bearings were incorporated, was made for trial, and the test of sudden air intrusion was carried out, as the results, various knowledges were obtained. The constitution of the testing setup, and the test results are reported. When air was injected at the pressure rise of 3.3x10 4 Pa/s from exhaust port side, after about 2.5 s, the maximum lift of 4.2x10 3 N arose. When air was injected at the pressure rise of 2.7x10 5 Pa/s from the suction part side, after about 0.4s, the maximum lift of 6.9x10 3 N arose. In the air injection alternately from the suction port and exhaust port sides, the emergency bearings functioned normally in 10 times of the test. (K.I.)

Molten salt destruction process for mixed wastes

International Nuclear Information System (INIS)

Upadhye, R.S.; Wilder, J.G.; Karlsen, C.E.

1993-04-01

We are developing an advanced two-stage process for the treatment of mixed wastes, which contain both hazardous and radioactive components. The wastes, together with an oxidant gas, such as air, are injected into a bed of molten salt comprising a mixture of sodium-, potassium-, and lithium-carbonates, with a melting point of about 580 degree C. The organic constituents of the mixed waste are destroyed through the combined effect of pyrolysis and oxidation. Heteroatoms. such as chlorine, in the mixed waste form stable salts, such as sodium chloride, and are retained in the melt. The radioactive actinides in the mixed waste are also retained in the melt because of the combined action of wetting and partial dissolution. The original process, consists of a one-stage unit, operated at 900--1000 degree C. The advanced two-stage process has two stages, one for pyrolysis and one for oxidation. The pyrolysis stage is designed to operate at 700 degree C. The oxidation stage can be operated at a higher temperature, if necessary

Method of solidifying liquid radioactive wastes

International Nuclear Information System (INIS)

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

1987-01-01

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

Numerical analysis of the bearing capacity of complex rock mechanical underground systems with filigree structures in the presence of imponderables. A contribution to the systematization of the investigative process with application/demonstration using the example of the salt cavern ASSE II/south flank

International Nuclear Information System (INIS)

Dyogtyev, Oleksandr

2017-01-01

The thesis dealing with the numerical analysis of the bearing capacity of complex rock mechanical underground systems with filigree structures in the presence of imponderables covers the following issues: status of science and technology, concept for the performance of numerical studies on the bearing capacity of large-volume underground systems, application example salt cavern ASSE II - application of the developed concept/development of numerical tools for the overall system/application of the global model to the given questions/realization of the modification potential.

Recoil effects of neutron-irradiated metal salts

International Nuclear Information System (INIS)

Lee, B.H.

1980-01-01

The distribution of sup(56)Mn and sup(38)Cl recoil species following radiative neutron capture permanganates, chlorates and perchlorates has been investigated by using ion-exchange chromatography method. The whole of the sup(56)Mn radioactivity in permanganates appeared in two valence states, the sup(38)Cl radioactivity in chlorates in two valence states and also the sup(38)Cl radioactivity in perchlorates in three valence states. Recoil energy was calculated. The internal conversion of sup(38m)Cl isomer transition affects the retention value. The greater the radii of the cation, the higher is the probability of the recoil atom breaking through the secondary cage. In ammonium salt, the ammonium ion behaves as a reducing agent. Crystal structures with their greater free space have shown by retention. (Author)

Possible salt mine sites for radioactive waste disposal in the northeastern states

Energy Technology Data Exchange (ETDEWEB)

Landes, K.K.

1972-06-30

The motivation for this investigation is the necessity for finding the safest possible repository for solid atomic plant wastes. It is believed that rooms mined in thick beds of salt would afford the best sanctuary. This is due especially to the impermeability of massive rock salt. This rock has enough plasticity so that it tends to give rather than fracture when disturbed by movements of the earth's crust. In addition, due to water conditions at the time of deposition, the rocks most commonly associated with salt (anhydrite and shale) are likewise relatively impervious. A number of areas have been selected for detailed discussion because of the excellence of the geological and environmental factors. The optimum requirements for a viable waste disposal prospect are described in detail and nine prospects are considered further.

Possible salt mine sites for radioactive waste disposal in the northeastern states

International Nuclear Information System (INIS)

Landes, K.K.

1972-01-01

The motivation for this investigation is the necessity for finding the safest possible repository for solid atomic plant wastes. It is believed that rooms mined in thick beds of salt would afford the best sanctuary. This is due especially to the impermeability of massive rock salt. This rock has enough plasticity so that it tends to give rather than fracture when disturbed by movements of the earth's crust. In addition, due to water conditions at the time of deposition, the rocks most commonly associated with salt (anhydrite and shale) are likewise relatively impervious. A number of areas have been selected for detailed discussion because of the excellence of the geological and environmental factors. The optimum requirements for a viable waste disposal prospect are described in detail and nine prospects are considered further

Mercury and selenium contamination in waterbird eggs and risk to avian reproduction at Great Salt Lake, Utah

Science.gov (United States)

Ackerman, Joshua T.; Herzog, Mark P.; Hartman, Christopher A.; Isanhart, John P.; Herring, Garth; Vaughn, Sharon; Cavitt, John F.; Eagles-Smith, Collin A.; Browers, Howard; Cline, Chris; Vest, Josh

2015-01-01

The wetlands of the Great Salt Lake ecosystem are recognized regionally, nationally, and hemispherically for their importance as breeding, wintering, and migratory habitat for diverse groups of waterbirds. Bear River Migratory Bird Refuge is the largest freshwater component of the Great Salt Lake ecosystem and provides critical breeding habitat for more than 60 bird species. However, the Great Salt Lake ecosystem also has a history of both mercury and selenium contamination, and this pollution could reduce the health and reproductive success of waterbirds. The overall objective of this study was to evaluate the risk of mercury and selenium contamination to birds breeding within Great Salt Lake, especially at Bear River Migratory Bird Refuge, and to identify the waterbird species and areas at greatest risk to contamination. We sampled eggs from 33 species of birds breeding within wetlands of Great Salt Lake during 2010 ̶ 2012 and focused on American avocets (Recurvirostra americana), black-necked stilts (Himantopus mexicanus), Forster’s terns (Sterna forsteri), white-faced ibis (Plegadis chihi), and marsh wrens (Cistothorus palustris) for additional studies of the effects of contaminants on reproduction.

Provenance of radioactive placers, Big Meadow area, Valley and Boise Counties, Idaho

International Nuclear Information System (INIS)

Truesdell, D.; Wegrzyn, R.; Dixon, M.

1977-02-01

For many years, radioactive black-sand placers have been known to be present in the Bear Valley area of west-central Idaho. The largest of these is in Big Meadow, near the head of Bear Valley Creek. Presence of these placers suggests that low-grade uranium deposits might occur in rocks of the Idaho Batholith, adjacent to Bear Valley. This study was undertaken to locate the provenance of the radioactive minerals and to identify problems that need to be solved before undertaking further investigations. The principal radioactive minerals in these placers are monazite and euxenite. Other minerals include columbite, samarskite, fergusonite, xenotime, zircon, allanite, sphene, and brannerite. Only brannerite is a uranium mineral; the others contain uranium as an impurity in crystal lattices. Radiometric determinations of the concentration of uranium in stream sediments strongly indicate that the radioactive materials originate in an area drained by Casner and Howard Creeks. Equivalent uranium levels in bedrock are highest on the divide between Casner and Howard Creeks. However, this area is not known to contain low-grade uranium occurrences. Euxenite, brannerite, columbite-tantalite, samarskite, and allanite are the principal radioactive minerals that were identified in rock samples. These minerals were found in granite pegmatites, granites, and quartz monzonites. Appreciably higher equivalent uranium concentrations were also found within these rock types. The major problem encountered in this study was the difficulty in mapping bedrock because of extensive soil and glacial mantle. A partial solution to this problem might be the application of radon emanometry so that radiometric measurements would not be limited to the sparse bedrock samples

Possibility of Radioactive and Toxic WasteDisposal in a Rock Ssalt Deposits in Slovakia Combining Wells and Cavities

Directory of Open Access Journals (Sweden)

Škvareková Erika

2004-09-01

Full Text Available Disposal of radioactive and toxic waste in rock salt can be performed in two ways – disposal in the salt mine repository or disposal in the deep wells connected with salt cavity. Presented article deals with the option of the disposal in a salt cavity at medium depths. The article also cover partially salt deposits in Slovakia and their potential suitability for waste disposal..

Synthesis of a metabolically stable modified long-chain fatty acid salt and its photolabile derivative

Energy Technology Data Exchange (ETDEWEB)

Stoll, G.H.; Voges, R.; Gerok, W.; Kurz, G. (Institut fuer Organische Chemie and Biochemie, Universitaet Freiburg (Germany))

1991-05-01

An analogue of the long-chain fatty acid salt, sodium stearate, was synthesized in which the hydrogen atoms at carbons 2, 3, and 18 were replaced by fluorine. The key step in the synthesis was the addition of 3-iodo-2,2,3,3-tetrafluoropropanoic acid amide to 15,15,15-trifluoro-1-pentadecene. Radioactivity was introduced by catalytic reduction of 2,2,3,3,18,18,18-heptafluoro-4-octadecenoic acid amide with carrier-free tritium gas yielding a product with the specific radioactivity of 2.63 TBq/mmol. The resulting 2,2,3,3,18,18,18-heptafluoro-4-octadecenoic acid has a pKa of about 0.5 and is completely dissociated under normal physiological conditions. The fluorinated fatty acid salt analogue is readily taken up into hepatocytes and proved to be metabolically inert. In an approach to the identification of proteins involved in long-chain fatty acid salt transport across membranes and intracellular compartments, the photolabile derivative 11,11-azo-2,2,3,3,18,18,18-heptafluoro(G-3H)octadecanoic acid sodium salt was synthesized with a specific radioactivity of 2.63 TBq/mmol. Photolysis of the photolabile derivative, using a light source with a maximum emission at 350 nm, occurred with a half-life of 1.5 min. The generated carbene reacted with 14C-labeled methanol and acetonitrile with covalent bond formation of 6-13%. Its efficacy for photoaffinity labeling was demonstrated by incorporation into serum albumin, the extracellular fatty acid salt-binding protein, as well as into the intracellular fatty acid salt-binding protein (FABP) of rat liver with the molecular weight of 14,000.

Comparison of folylderivative biosynthesis in Ehrlich ascites carcinoma cells and in some organs of healthy and tumor-bearing mice

Energy Technology Data Exchange (ETDEWEB)

Sikora, E; Grzelakowska-Sztabert, B [Polska Akademia Nauk, Warsaw. Inst. Biologii Doswiadczelnej

1984-01-01

Biosynthesis of folyl derivatives derived from subcutaneously injected 2-(/sup 14/C)folate was studied in Ehrlich ascites carcinoma (EAC) cells and in mouse liver and kidneys. Retention of exogenous folate was followed by measurements of the total radioactivity of folyl derivatives present in the EAC cells and organs examined. Identification of unconjugated and conjugated folyl derivatives was done by means of column chromatography on Sephadex G-25, G-15 and cellulose sheets. The level of retained radioactivity in folyl derivatives, being 5% in the liver and 1% in the kidneys of the radioactivity administered to mice, was similar in healthy and tumor-bearing animals. Moreover, no quantitative and qualitative differences were found in folyl mono- and polyglutamates originating from the organs of healthy or tumor-bearing mice although the content of folyl polyglutamates rose faster in liver and kidneys of EAC cells-bearing mice as well as in the tumor cells, than in the organs of healthy mice.

Treatment technology of low concentration uranium-bearing wastewater and its research progress

International Nuclear Information System (INIS)

Wei Guangzhi; Xu Lechang

2007-01-01

With growth of the discharged uranium-bearing wastewater capacity, a low cost and effective treatment technology is required to avoid transferring and diffusion of the radioactive nuclides. On the basis of analyses of the source and characteristics of the low-concentration uranium-bearing wastewater, the conventional treatment technologies, such as, flocculating settling, ion exchange, concentration, adsorption, and some innovatory technologies, such as, membrane, microorganism, phytoremediation and zero-valent iron technology are introduced. (authors)

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)

Constitutive modeling of salt behavior: State of the technology

International Nuclear Information System (INIS)

Munson, D.E.; Wawersik, W.R.

1992-01-01

The modern investigation of the thermomechanical behavior of salt started in the mid-1930's and, for what appears to be a very narrow discipline, ''salt mechanics'' has acquired considerable technical depth and sophistication. The last three decades have been especially productive in constitutive model development and laboratory investigations of time-dependent creep behavior. This has been largely due ot anticipated use of domal or bedded salt deposits as sites for radioactive waste repositories and to expanded need for hydrocarbon and feedback storage caverns. Salt is an interesting material, in that it is ''metal-like''; and, therefore, constitutive modeling can draw upon a large body of metal deformation information to arrive at appropriate models of behavior. Testing apparatus and methods have centered on either uniaxial or triaxial compression to obtain steady state and transient creep responses. Flow and fracture potentials have been defined. Validation attempts of the models against field data, although limited, have proved promising. The objective here is to summarize the state-of-the-technology of the constitutive modeling of salt behavior or ''salt mechanics.''

Organic Matter in Extraterrestrial Water-Bearing Salt Crystals

Science.gov (United States)

Chan, Q. H. S.; Zolensky, M. E.; Kebukwa, Y.; Fries, M.; Steele, A.

2017-01-01

Introduction: Direct samples of early Solar System fluids are present in two thermally-metamorphosed ordinary chondrite regolith breccias (Monahans (1998) [H5] and Zag [H3-6]), which were found to contain brine-bearing halite (NaCl) crystals that have been added to the regolith of an S-type asteroid following asteroidal metamorphism [1, 2]. The brine-bearing halite grains were proposed to be formed on an icy C-type asteroids (possibly Ceres), and transferred to an S-type asteroid via cryovolcanic event(s) [3]. A unique aspect of these halites is that they contain abundant organic rich solid inclusions hosted within the halites alongside the water inclusions. Methods: We analyzed in detail the compositions of the organic solids and the amino acid content of the halite crystals with two-step laser desorption/laser ionization mass spectrometry (L(sup 2) MS), Raman spectroscopy, X-ray absorption near edge structure (XANES), nanoscale secondary ion mass spectrometry (NanoSIMS), and ultra-performance liquid chromatography fluorescence detection and quadrupole time of flight hybrid mass spectrometry (UPLC-FD/QToF-MS). Results and Discussion: The L(sup 2) MS results show signatures of low-mass polyaromatic hydro-carbons (PAHs) indicated by sequences of peaks separated by 14 atomic mass units (amu) due to successive addition of methylene (CH2) groups to the PAH skeletons [4]. Raman spectra of the micron-sized solid inclusions of the halites indicate the presence of abundant and highly variable organic matter that include a mixture of short-chain aliphatic compounds and macromolecular carbon. C-XANES analysis identified C-rich areas with peaks at 285.0 eV (aromatic C=C) and 286.6 eV (vinyl-keto C=O). However, there is no 1s-sigma* exciton peak (291.7 eV) that is indicative of the development of graphene structure [5], which suggests the organics were synthesized cold. Na-noSIMS analyses show C-rich and N-rich areas that exhibit similar isotopic values with that of the IOM in

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

Impact of solid second phases on deformation mechanisms of naturally deformed salt rocks (Kuh-e-Namak, Dashti, Iran) and rheological stratification of the Hormuz Salt Formation

Science.gov (United States)

Závada, P.; Desbois, G.; Urai, J. L.; Schulmann, K.; Rahmati, M.; Lexa, O.; Wollenberg, U.

2015-05-01

Viscosity contrasts displayed in flow structures of a mountain namakier (Kuh-e-Namak - Dashti), between 'weak' second phase bearing rock salt and 'strong' pure rock salt types are studied for deformation mechanisms using detailed quantitative microstructural study. While the solid inclusions rich ("dirty") rock salts contain disaggregated siltstone and dolomite interlayers, "clean" salts reveal microscopic hematite and remnants of abundant fluid inclusions in non-recrystallized cores of porphyroclasts. Although the flow in both, the recrystallized "dirty" and "clean" salt types is accommodated by combined mechanisms of pressure-solution creep (PS), grain boundary sliding (GBS), transgranular microcracking and dislocation creep accommodated grain boundary migration (GBM), their viscosity contrasts observed in the field outcrops are explained by: 1) enhanced ductility of "dirty" salts due to increased diffusion rates along the solid inclusion-halite contacts than along halite-halite contacts, and 2) slow rates of intergranular diffusion due to dissolved iron and inhibited dislocation creep due to hematite inclusions for "clean" salt types Rheological contrasts inferred by microstructural analysis between both salt rock classes apply in general for the "dirty" salt forming Lower Hormuz and the "clean" salt forming the Upper Hormuz of the Hormuz Formation and imply strain rate gradients or decoupling along horizons of mobilized salt types of different composition and microstructure.

Salt geologic evaluation of the impact of cryogenic fissures and halokinetic deformation processes on the integrity of the geological barrier of the salt dome Gorleben

International Nuclear Information System (INIS)

Hammer, Joerg; Fleig, Stephanie; Mingerzahn, Gerhard

2012-07-01

In several salt domes of the area close to Hannover fissures were observed that might be caused by thermally induced fissure formation due to cold periods (cryogenic fissures). Comprehensive substantial-structural analyses are performed as an example for the salt dome Bokeloh with respect to genesis and transferability to the salt dome Gorleben. Based on recent structure-geological, mineralogical-geochemical and micro-paleontological studies and thermo-mechanical modeling a solely thermally induced fissure formation due to cold periods is unlikely for the salt dome Bokeloh. There is a direct relation between the genesis of the salt dome Bokeloh, its regional tectonic site and the fissure formation. Due to the completely different genesis and another regional-tectonic situation the existence of cryogenic fissures is excluded for the salt dome Gorleben. The salt-geologic and experimental studies on the deformation of anhydrite layers in salt domes are summarized and evaluated with respect to the long-term consequences for a potential final repository for high-level heat-generating radioactive waste in the salt dome Gorleben. The studies confirm the older BGR studies that anhydrite layers do not represent hydraulic potential ling-distance liquid paths.

Studies on radioactivity distribution and radioactive mineral identification in uranium ores from Espinharas (PB), Brazil

International Nuclear Information System (INIS)

Oliveira, G.N.M. de.

1979-01-01

Studies about the identification of radioactive minerals in uranium bearing rocks from Espinharas (PB), Brazil are presented. Autoradiography with α-sensitive nuclear emulsions was utilized for determining radioctivity distributions and for localizing radioactive minerals, in combination with microscopy, X-ray diffractometry, PIXE and eletron microprobe analysis for its identification. Mineralized gneisse and feldspatic rock, the two principal samples studied, show distinct differences in radioactive distribution patterns, however the main carriers for U and Th seem to be the same. Microanalysis shows that elements are associated with Si, Ca, Fe and Al an some trace elements like Y, Zr, Ti, etc. U and Th are distributed uniformly in feldspatic rock and inhomogeneously in mineralized gneisse, indicating that the zonary structure of the radioactive cristals, frequently observed in gneisse, could be due to variable U:Th ratios. Chemical analysis, X-ray diffraction datas and microscopic studies indicates that the principal carrier for radioactivity in the rocks of Espinharas is a silicate mineral of U and Th, probably situaded in the series of transition: Coffinite -> uraninite, thorogummite -> thorianite. Some additional experiments about leachability of uranium with diluted sulfuric acid are reported, which confirm the different nature of radioactivity distribution in feldspatic and gneissic rocks. (author) [pt

Lanthanide bearing radioactive particles for cancer therapy and multimodality imaging

NARCIS (Netherlands)

Zielhuis, S.W.

2006-01-01

Local radionuclide therapy using radioactive microspheres is a promising therapy for patients suffering from liver malignancies. In contrast to normal liver tissue, which receives most of its blood flow from the portal vein, liver malignancies are almost exclusively dependent on arterial blood

Advance in the study of removal of cesium from radioactive wastewater by inorganic ion exchangers

International Nuclear Information System (INIS)

Wang Songping; Wang Xiaowei; Du Zhihui

2014-01-01

The excellent performance in the removal of cesium from radioactive wastewater by inorganic ion exchangers has received extensive attention due to their characteristic physico-chemical features. The paper summarized research progress of removal of cesium by different inorganic ion exchangers such as silicoaluminate, salts of hetero polyacid, hexacyanoferrate, insoluble salts of acid with multivalent metals, insoluble hydrous oxides of multivalent metals and silicotitanate and reviewed several removal systems of cesium by inorganic ion exchangers which might offer China some reference in treatment and disposal of radioactive wastewater. (authors)

Evaluation of gross radioactivity in foodstuffs

International Nuclear Information System (INIS)

Zorer, Oezlem Selcuk; Oeter, Cigdem

2015-01-01

The paper presents the results of radiological investigations of food products sampled in the summer and fall of 2011 and 2012 in different parts of Van, Turkey. Gross radioactivity measurements in food products were evaluated. Food items were divided into eight groups: (1) water, (2) fish, (3) cheese products, (4) fruits, (5) vegetables, (6) herbs, (7) walnut and (8) rock salt. The levels of the gross alpha and gross beta radioactivity in all food samples varied widely ranging from 0.070 to 10.885 Bq/g and from 0.132 to 48.285 Bq/g on dry mass basis, respectively. In one sample, gross alpha and gross beta activity concentrations were found to be relatively high according to the other samples and in all samples, the gross alpha radioactivity was measured lower than the gross beta radioactivity. The gross α and gross β activities were measured by using α/β counter of the multi-detector low background system (PIC MPC-9604).

Evaluation of gross radioactivity in foodstuffs

Energy Technology Data Exchange (ETDEWEB)

Zorer, Oezlem Selcuk; Oeter, Cigdem [Yuzuncu Yil Univ., Van (Turkey). Dept. of Chemistry

2015-05-15

The paper presents the results of radiological investigations of food products sampled in the summer and fall of 2011 and 2012 in different parts of Van, Turkey. Gross radioactivity measurements in food products were evaluated. Food items were divided into eight groups: (1) water, (2) fish, (3) cheese products, (4) fruits, (5) vegetables, (6) herbs, (7) walnut and (8) rock salt. The levels of the gross alpha and gross beta radioactivity in all food samples varied widely ranging from 0.070 to 10.885 Bq/g and from 0.132 to 48.285 Bq/g on dry mass basis, respectively. In one sample, gross alpha and gross beta activity concentrations were found to be relatively high according to the other samples and in all samples, the gross alpha radioactivity was measured lower than the gross beta radioactivity. The gross α and gross β activities were measured by using α/β counter of the multi-detector low background system (PIC MPC-9604).

Behavior of crushed salt under heat source in boreholes in a salt mine (Amelie Mine, Alsace Potash Mines, France)

International Nuclear Information System (INIS)

Ghoreychi, M.

1991-01-01

The study of thermomechanical interaction between rock salt and crushed salt, used as a backfilling material at the final stage of radioactive waste disposal in salt formations, led to perform an in situ test at the Amelie Mine(The Alsace Potash Mines in France). The field tests site is located at a depth of 520m and the tests were performed in six parallel boreholes. Five boreholes were backfilled using three types of crushed salt, changing by their grain size (fine = 0.4 mm; natural = 1 mm; coarse = 2 mm). The sixth borehole was not backfilled in order to witness for rock salt behavior without backfilling confinement. Except the first borehole used as a pilot test, the four backfilled boreholes were heated during four months with two levels of heat output (1.6 kW, then 2.2 kW). Cooling was also followed during four months after heating interruption. The maximum of temperature obtained on the wall of the backfilled boreholes was about 100 0 C during the first field test and 130 0 C during the second. The thermal diffusivity of rock mass and the coefficient of heat exchange by convection are studied. In spite of the case that the crushed salt thermal conductivity is initially ten times less than of rock salt, no excessive temperature concentration was obtained on the heat sources

Mixed Waste Salt Encapsulation Using Polysiloxane - Final Report

International Nuclear Information System (INIS)

Miller, C.M.; Loomis, G.G.; Prewett, S.W.

1997-01-01

A proof-of-concept experimental study was performed to investigate the use of Orbit Technologies polysiloxane grouting material for encapsulation of U.S. Department of Energy mixed waste salts leading to a final waste form for disposal. Evaporator pond salt residues and other salt-like material contaminated with both radioactive isotopes and hazardous components are ubiquitous in the DOE complex and may exceed 250,000,000 kg of material. Current treatment involves mixing low waste percentages (less than 10% by mass salt) with cement or costly thermal treatment followed by cementation to the ash residue. The proposed technology involves simple mixing of the granular salt material (with relatively high waste loadings-greater than 50%) in a polysiloxane-based system that polymerizes to form a silicon-based polymer material. This study involved a mixing study to determine optimum waste loadings and compressive strengths of the resultant monoliths. Following the mixing study, durability testing was performed on promising waste forms. Leaching studies including the accelerated leach test and the toxicity characteristic leaching procedure were also performed on a high nitrate salt waste form. In addition to this testing, the waste form was examined by scanning electron microscope. Preliminary cost estimates for applying this technology to the DOE complex mixed waste salt problem is also given

Gas release during salt well pumping: model predictions and comparisons to laboratory experiments

International Nuclear Information System (INIS)

Peurrung, L.M.; Caley, S.M.; Bian, E.Y.; Gauglitz, P.A.

1996-09-01

The Hanford Site has 149 single-shell tanks (SSTs) containing radioactive wastes that are complex mixes of radioactive and chemical products. Some of these wastes are known to generate mixtures of flammable gases, including hydrogen, nitrous oxide, and ammonia. Nineteen of these SSTs have been placed on the Flammable Gas Watch List (FGWL) because they are known or suspected, in all but one case, to retain these flammable gases. Salt well pumping to remove the interstitial liquid from SSTs is expected to cause the release of much of the retained gas, posing a number of safety concerns. Research at the Pacific Northwest National Laboratory (PNNL) has sought to quantify the release of flammable gases during salt well pumping operations. This study is being conducted for Westinghouse Hanford Company as part of the PNNL Flammable Gas Project. Understanding and quantifying the physical mechanisms and waste properties that govern gas release during salt well pumping will help to resolve the associated safety issues

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.)

Ni-Ti Next Generation Bearings for Space Applications

Science.gov (United States)

DellaCorte, Christopher

2018-01-01

NASA applications challenge traditional bearing materials. The rigors of launch often include heavy shock loads and exposure to corrosive environments (e.g., salt spray). Unfortunately, ball and roller bearings made from hardened steels are vulnerable to Brinell denting and rust which can limit performance and life. Ceramic materials can eliminate corrosion concerns but their high stiffness and extreme hardness actually makes denting problems worse. In this presentation, an emerging superelastic alloy, NiTi, is introduced for rolling element bearing applications. Through a decade of RD, NiTi alloy bearings have been put through a comprehensive series of life and performance tests. Hardness, corrosion, strength, stiffness, and rolling contact fatigue tests have been conducted and reported. Ball bearings ranging in size from 12 to 50mm bore have been successfully engineered and operated over a wide range of speeds and test conditions including being submerged in water. The combination of high hardness, moderate elastic modulus, low density, and intrinsic corrosion immunity provide new possibilities for mechanisms that operate under extreme conditions. Recent preliminary tests indicate that bearings can be made from NiTi alloys that are easily lubricated by conventional oils and greases and exhibit acceptable rolling contact fatigue resistance. This presentation introduces the NiTi materials systems and shows how NASA is using it to alleviate several specific problems encountered in advanced space applications.

Ultimate storage of thorium-bearing waste

International Nuclear Information System (INIS)

Ganser, B.

1986-01-01

The goal of this R and D project was to experimentally determine the release of the radioactive noble gas radon from thorium-bearing waste. For the experiments, three 200 litre waste forms have been prepared: One package consisting of inactive cement (for blank value determination), the second of cemented, radioactive sludge precipitate (for reference value determination), and the third of untreated sludge precipitate in a drum. The release rate measured on the reference package at room temperature is 3.1x10 10 Bq/a for Rn-220, and 2.4x10 6 Bq/a for Rn-222. The release rate from a drum under equal conditions is 4.1x10 8 Bq/a for Rn-220, and 2.1x10 6 Bq/a for Rn-222. (orig./RB) [de

Sealing considerations for repository shafts in bedded and dome salt

International Nuclear Information System (INIS)

1981-12-01

The report reviews the geologic and hydrologic data base for penetration seal designs referenced to the Los Medanos bedded salt site in New Mexico and to four candidate salt domes in the Gulf Interior. Experience with existing shafts highlights the importance, for shaft decommissioning as well as operation, of achieving an adequate seal at and immediately below the top of salt. Possible construction procedures for repository shafts are reviewed, noting advantages and disadvantages with respect to repository sealing. At this stage, there does not appear to be a clear preference for excavation by drill and blast or by drilling. If conventional drill and blast methods are used, it may be necessary to grout in permeable zones above the salt. An important consideration with respect to sealing is that grouting operations (or freezing should it be used) should not establish connections between the top of salt and water-bearing zones higher in the stratigraphic section. Generally, it is concluded that Los Medanos and the dome salt sites are favorable candidate repository sites from the point of view of sealing

Final storage of radioactive waste in Germany. Are administrative structures in need of modification?

International Nuclear Information System (INIS)

Schneider, Horst

2011-01-01

Delays in commissioning the Konrad Mine as a repository for radioactive waste not generating heat, and in exploring the Gorleben salt dome for suitability as a repository for high-level waste generating heat, invite the question whether the legal regulations in place, especially administration and funding of the repository, are suitable for solving current problems or whether they are in need of improvement. The key principles of the back end of the nuclear fuel cycle, final storage included, were laid down as rules in 1976. Execution of the necessary waste management steps, from radioactive waste arisings to their final disposal, was split between private responsibilities and government competences. Final storage, to this day, has been of prime importance. Pursuant to the Atomic Energy Act, the federal government is required to set up facilities for final storage of radioactive waste. The waste management duties incumbent upon private parties, from radioactive waste arisings to delivery, are mainly subject to safety criteria under the Atomic Energy Act and the Radiation Protection Ordinance. As far as administration is concerned, the private parties are free in the way they comply with regulatory requirements. They are required to bear the cost in accordance with the polluter-pays-principle. In the light of the sluggish execution of government tasks from 1976 to this day, the question of improvements has become more acute than ever. This is where assignment offers an approach towards better administration which can be taken at short notice, as assignment implies a reduction in the number of interfaces and clearer responsibilities. However, even the best administration is unable to lead to the repositories required by law if those responsible in government fail to act in accordance with the spirit and letter of the law. (orig.)

Status and future developments of risk analysis for repositories of radioactive wastes in salt formations in the Federal Republic of Germany

International Nuclear Information System (INIS)

Proske, R.

1977-01-01

In the Federal Republic of Germany a hypothetical repository for High-Level Radioactive Waste in a salt formation was taken as basis for a first attempt to use the methodology of risk analysis in order to get statements on the safety of such a geologic and mining system. Several institutions were engaged in drawing up fault trees, development of release models and calculation of the risk. A lot of R+D-work is scheduled to be carried out in future which includes optimization of the application of risk analysis methodology to geologic and mining systems, further development of release models, development of a model describing the migration of radionuclides in typical geologic strata and soils of the Federal Republic of Germany and application of risk analysis methodology to different repositories and disposal technologies

Origins of the Salado, Seven Rivers, and San Andres salt margins in Texas and New Mexico: Revision 1: Topical report

International Nuclear Information System (INIS)

Boyd, S.D.; Murphy, P.J.

1987-02-01

The present boundaries of the San Andres, Seven Rivers, and Salado salts generally lie along the periphery of the Palo Duro and Tucumcari Basins. Various geologic mechanisms occurring singularly or in combination determined the positions of the salt margins. These mechanisms include nondeposition of salt and syndepositional and postdepositional dissolution. In New Mexico, San Andres units pinch out against the Pedernal and Sierra Grande Uplifts, indicating that nondeposition established the original salt margins there. Syndepositional dissolution of exposed Upper San Andres salts occurred in response to Guadalupian upwarp of the basin margins. Triassic erosion differentially removed Permian salt-bearing formations along the uplifts. Late Tertiary dissolution is indicated by fill of north-south trending collapse valleys. In Texas, Guadalupian upwarp along the Amarillo Uplift caused pinchout of Units 2 and 3 in the Lower San Andres and influenced the deposition of subsequent salt-bearing strata. The discontinuity of Upper San Andres evaporites across the Amarillo Uplift suggests syndepositional dissolution. Along the eastern and northeastern basin margin, dissolution may have accompanied Triassic erosion of locally uplifted Upper Permian strata. Tertiary dissolution is recognized beneath anomalously thick Ogallala Formation sections that overlie collasped Permian strata. 49 refs., 31 figs., 2 tabs

Effect of potassium-salt muds on gamma ray, and spontaneous potential measurements

International Nuclear Information System (INIS)

Cox, J.W.; Raymer, L.L.

1976-01-01

Interpretations of the gamma ray and Spontaneous Potential curves generally assume the presence of sodium chloride as the dominant salt in both the formation water and the mud filtrate. However, potassium-salt muds are increasingly being used by the oil industry. The potassium cation is significantly different from the sodium cation in its radioactive and electrochemical properties. Natural potassium contains a radioactive isotope which emits gamma rays. Thus, the presence of potassium salts in the mud system may contribute to Gamma-Ray tool response. Since the Gamma Ray is used quantitatively in many geological sequences as an indicator of clay content, a way to correct for the effect of potassium in the mud column is desirable. Correction methods and charts based on laboratory measurements and field observations are presented. The effect of temperature on the resistivity of potassium muds is also briefly discussed. From data available, it appears to be similar to that for NaCl muds. On the bases of field observations and laboratory work, the electrochemical properties of potassium-chloride and potassium-carbonate muds and mud filtrates are discussed. Activity relationships are proposed, and the influence of these salts on the SP component potentials--namely, the liquid-junction, membrane, and bi-ionic potentials--is described. Several field examples are presented

Polar bears at risk

Energy Technology Data Exchange (ETDEWEB)

Norris, S.; Rosentrater, L.; Eid, P.M. [WWF International Arctic Programme, Oslo (Norway)

2002-05-01

rains also destroy the denning habitat of ringed seals, the polar bears' primary prey. Declines in the ringed seal population would mean a loss of food for polar bears. A trend toward stronger winds and increasing ice drift observed in some parts of the Arctic over the last five decades will likely increase energy expenditures and stress levels in polar bears that spend most of their lives on drifting sea ice. Polar bears face other limiting factors as well. Historically, the main threat to polar bears has been hunting. Satisfactory monitoring information has been obtained for most polar bear populations in recent years, however there is concern about hunting in areas without formal quota systems, such as Greenland. A range of toxic pollutants, including heavy metals, radioactivity, and persistent organic pollutants (POPs) are found throughout the Arctic. Of greatest concern are the effects of POPs on polar bears, which include a general weakening of the immune system, reduced reproductive success and physical deformities. The expansion of oil development in the Arctic poses additional threats; for example, disturbances to denning females in the Arctic National Wildlife Refuge in Alaska could undermine recruitment of the Beaufort Sea polar bear population. These threats, along with other effects of human activity in the Arctic, combine to pressure polar bears and their habitat. Large carnivores are sensitive indicators of ecosystem health and can be used to define the minimum area necessary to preserve intact ecosystems. WWF has identified the polar bear as a unique symbol of the complexities and interdependencies of the arctic marine ecosystem as it works toward its goal of preserving biodiversity for future generations.

Radioactive waste generated from JAERI partitioning-transmutation cycle system

Energy Technology Data Exchange (ETDEWEB)

Shinichi, Nakayama; Yasuji, Morita; Kenji, Nishihara [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan)

2001-07-01

Production of lower-level radioactive wastes, as well as the reduction in radioactivity of HLW, is an important performance indicator in assessing the viability of a partitioning-transmutation system. We have begun to identify the chemical compositions and to quantify the amounts of radioactive wastes that may be generated by JAERI processes. Long-lived radionuclides such as {sup 14}C and {sup 59}Ni and spallation products of Pb-Bi coolants are added to the existing inventory of these nuclides that are generated in the current fuel cycle. Spent salts of KCl-LiCl, which is not generated from the current fuel cycle, will be introduced as a waste. (author)

Radioactive Wastes Generated From JAERI Partitioning-Transmutation Fuel Cycle

International Nuclear Information System (INIS)

Nakayama, Shinichi; Morita, Yasuji; Nishihara, Kenji

2003-01-01

Production of lower-level radioactive wastes, as well as the reduction in radioactivity of HLW, is an important performance indicator in assessing the viability of a partitioning-transmutation system. We have begun to identify the chemical compositions and to quantify the amounts of radioactive wastes that may be generated by JAERI's processes. Long-lived radionuclides such as 14 C and 59 Ni and spallation products of Pb-Bi coolants are added to the existing inventory of these nuclides that are generated in the current fuel cycle. Spent salts of KCl-LiCl, which is not generated from the current fuel cycle, will be introduced as a waste. (authors)

Radioactive waste in Federal Germany

International Nuclear Information System (INIS)

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

1988-01-01

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

EVIDENCE OF CORROSIVE GAS FORMED BY RADIOLYSIS OF CHLORIDE SALTS IN PLUTONIUM-BEARING MATERIALS

Energy Technology Data Exchange (ETDEWEB)

Dunn, K.; Louthan, M.

2010-02-01

Corrosion and pitting have been observed in headspace regions of stainless steel containers enclosing plutonium oxide/salt mixtures. These observations are consistent with the formation of a corrosive gas, probably HCl, and transport of that gas to the headspace regions of sealed containers. The NH{sub 4}Cl films found on the walls of the sealed containers is also indicative of the presence of HCl gas. Radiolysis of hydrated alkaline earth salts is the probable source of HCl.

Testing thin-skinned inversion of a prerift salt-bearing passive margin (Eastern Prebetic Zone, SE Iberia)

Science.gov (United States)

Escosa, Frederic O.; Roca, Eduard; Ferrer, Oriol

2018-04-01

Detailed geologic mapping combined with well and seismic data from the Eastern Prebetic Zone (SE Iberia) reveal extensional and contractional structures that permit characterization of passive margin development and its incorporation into a thin-skinned fold-and-thrust belt. The study area is represented by NW-directed, ENE-trending folds and thrusts faults locally disrupted by the NW-trending Matamoros Basin and the active Jumilla and La Rosa diapirs. These structures resulted from the thin-skinned inversion of the proximal part of the Eastern South Iberian passive margin containing prerift salt. Here, Upper Jurassic to Santonian thick-skinned extension controlled the accumulation of sediment over mobile prerift salt. This in turn defined the style of salt tectonics characterized by monoclinal drape folds, suprasalt extensional faults and diapirs. The structural and sedimentological analysis suggests that during extension, salt localizes strain thus decoupling sub- and suprasalt deformation. Thick-skinned extension controls suprasalt deformation as well as its location and distribution which changes over time. Salt also localizes strain during inversion. The preexisting salt structures, weaker than adjacent areas, preferentially absorb the contractional deformation. In addition, the stepped subsalt geometry that results from thick-skinned extension also controls the shortening propagation. Therefore, the degree of strain localization depends on the thickness of the suprasalt cover and on the dip of subsalt faults relative to the thin-skinned transport direction.

USGS studies of physical--chemical relationships in salt repositories

International Nuclear Information System (INIS)

Stewart, D.B.

1977-01-01

The amount and physical properties of brine that can occur in salt repositories at elevated temperatures and pressures adjacent to waste-bearing cannisters will have considerable impact on the mechanical strength and stability of the repository. Brine will form readily from H 2 O absorbed on surfaces, diffusion along grain boundaries, movement of fluid inclusions, dehydration of hydrous minerals such as gypsum, polyhalite or clays, or even from leakage through failed shaft openings. A T-P diagram for NaCl--H 2 O shows the limits for coexisting solid-liquid-gas assemblages in salt repositories. Isobaric T-X diagrams are included to show compositional details at pressures below the critical point and above it. Properties of the fluid phases such as volume, density, heat capacity, enthalpy, viscosity, surface tension, osmotic coefficients, etc. can be well described (>0.01% to 2 O. Because aggregates of solids are mechanically weakened by interstitial liquid, determination of mechanical properties of brine-bearing aggregates is needed. A maximum of one third by weight of brine (not H 2 O), and probably much less, will destroy rock strength

Treatment of radioactive waste salt by using synthetic silica-based phosphate composite for de-chlorination and solidification

Science.gov (United States)

Cho, In-Hak; Park, Hwan-Seo; Lee, Ki-Rak; Choi, Jung-Hun; Kim, In-Tae; Hur, Jin Mok; Lee, Young-Seak

2017-09-01

In the radioactive waste management, waste salts as metal chloride generated from a pyrochemical process to recover uranium and transuranic elements are one of problematic wastes due to their intrinsic properties such as high volatility and low compatibility with conventional glasses. This study reports a method to stabilize and solidify LiCl waste via de-chlorination using a synthetic composite, U-SAP (SiO2-Al2O3-B2O3-Fe2O3-P2O5) prepared by a sol-gel process. The composite was reacted with alkali metal elements to produce some metal aluminosilicates, aluminophosphates or orthophosphate as a crystalline or amorphous compound. Different from the original SAP (SiO2-Al2O3-P2O5), the reaction product of U-SAP could be successfully fabricated as a monolithic wasteform without a glassy binder at a proper reaction/consolidation condition. From the results of the FE-SEM, FT-IR and MAS-NMR analysis, it could be inferred that the Si-rich phase and P-rich phase as a glassy grains would be distributed in tens of nm scale, where alkali metal elements would be chemically interacted with Si-rich or P-rich region in the virgin U-SAP composite and its products was vitrified into a silicate or phosphate glass after a heat-treatment at 1150 °C. The PCT-A (Product Consistency Test, ASTM-1208) revealed that the mass loss of Cs and Sr in the U-SAP wasteform had a range of 10-3∼10-1 g/m2 and the leach-resistance of the U-SAP wasteform was comparable to other conventional wasteforms. From the U-SAP method, LiCl waste salt was effectively stabilized and solidified with high waste loading and good leach-resistance.

Low- and intermediate level radioactive waste from Risoe, Denmark. Location studies for potential disposal areas. Report no. 11. Description of areas. Danish and English summary; Low- and intermediate level radioactive waste from Risoe, Denmark. Location studies for potential disposal areas. Report no. 11. Omraadebeskrivelser - Description of areas. Dansk og engelsk resume

Energy Technology Data Exchange (ETDEWEB)

Gravesen, P.; Nilsson, B.; Schack Pedersen, S.A.; Binderup, M.

2011-07-01

The low - and intermediate level radioactive waste from Risoe: the nuclear reactor buildings, different types of material from the research periods and waste from hospitals and research institutes have to be stored in a final disposal in Denmark for at least 300 years. The task is to locate and recognize sediments or rocks with low permeability which can isolate the radioactive waste from the surrounding deposits, the groundwater resources, the recipients and from human activities. The sediments or rocks shall also act as a protection if the waste disposal leaks radioactive material to the surroundings. This goal can be reached by choosing deposits with low water flow and high sorption potential of the sediments or rocks. The investigation of geological deposits as potential waste disposals for high radioactive waste from nuclear power plants has earlier focused on deep seated salt deposits and basement rocks but the Tertiary clays were also mapped. The salt diapirs, salt pillows and salt deposits and deep basement rocks are not included in the present study. These rocks and deposits are situated too deep for the present study and salt deposits seem to be unstable for a disposal (e.g. German salt mines). The regional geologic survey based on existing data was concluded by selecting 22 areas in Denmark. There remains now to reduce the number of potential areas to 1-3 where detailed field studies will be performed in order to select the final location. (LN)

Radioactive waste isolation in salt: special advisory report on the status of the Office of Nuclear Waste Isolation's plans for repository performance assessment

International Nuclear Information System (INIS)

Ditmars, J.D.; Walbridge, E.W.; Rote, D.M.; Harrison, W.; Herzenberg, C.L.

1983-10-01

Repository performance assessment is analysis that identifies events and processes that might affect a repository system for isolation of radioactive waste, examines their effects on barriers to waste migration, and estimates the probabilities of their occurrence and their consequences. In 1983 Battelle Memorial Institute's Office of Nuclear Waste Isolation (ONWI) prepared two plans - one for performance assessment for a waste repository in salt and one for verification and validation of performance assessment technology. At the request of the US Department of Energy's Salt Repository Project Office (SRPO), Argonne National Laboratory reviewed those plans and prepared this report to advise SRPO of specific areas where ONWI's plans for performance assessment might be improved. This report presents a framework for repository performance assessment that clearly identifies the relationships among the disposal problems, the processes underlying the problems, the tools for assessment (computer codes), and the data. In particular, the relationships among important processes and 26 model codes available to ONWI are indicated. A common suggestion for computer code verification and validation is the need for specific and unambiguous documentation of the results of performance assessment activities. A major portion of this report consists of status summaries of 27 model codes indicated as potentially useful by ONWI. The code summaries focus on three main areas: (1) the code's purpose, capabilities, and limitations; (2) status of the elements of documentation and review essential for code verification and validation; and (3) proposed application of the code for performance assessment of salt repository systems. 15 references, 6 figures, 4 tables

Admissible thermal loading in geological formations. Consequences on radioactive waste disposal methods

International Nuclear Information System (INIS)

1982-01-01

The study of the ''Admissible thermal loading in geological formations and its consequence on radioactive waste disposal methods'' comprises four volumes: Volume 1. ''Synthesis report'' (English/French text). Volume 2. Granite formations (French text). Volume 3. Salt formations (German text). Volume 4. Clay formations (French text). The present ''synthesis report'' brings together the formation produced by the three specific studies dealing with granite, salt and clay

Ultimate disposal of radioactive wastes

International Nuclear Information System (INIS)

Roethemeyer, H.

1991-01-01

The activities developed by the Federal Institution of Physical Engineering PTB and by the Federal Office for Radiation Protection (BfS) concentrated, among others, on work to implement ultimate storage facilities for radioactive wastes. The book illuminates this development from site designation to the preliminary evaluation of the Gorleben salt dome, to the preparation of planning documents proving that the Konrad ore mine is suitable for a repository. The paper shows the legal provisions involved; research and development tasks; collection of radioactive wastes ready for ultimate disposal; safety analysis in the commissioning and post-operational stages, and product control. The historical development of waste management in the Federal Republic of Germany and international cooperation in this area are outlined. (DG) [de

Thermal conductivity measurements in relation to the geothermal exploration of the Gorleben salt dome

International Nuclear Information System (INIS)

Kopietz, J.

1985-01-01

The results of thermal conductivity measurements on rock salt and associated structures are presented in this paper. Thermal conductivity data obtained from laboratory measurements on the core material are compared with high-precision temperature gradient logs from the exploration boreholes. This work is part of an extensive investigation into the suitability of the Gorleben salt done in northern Germany as a radioactive waste disposal site

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

Molten salt processing of mixed wastes with offgas condensation

International Nuclear Information System (INIS)

Cooper, J.F.; Brummond, W.; Celeste, J.; Farmer, J.; Hoenig, C.; Krikorian, O.H.; Upadhye, R.; Gay, R.L.; Stewart, A.; Yosim, S.

1991-01-01

We are developing an advanced process for treatment of mixed wastes in molten salt media at temperatures of 700--1000 degrees C. Waste destruction has been demonstrated in a single stage oxidation process, with destruction efficiencies above 99.9999% for many waste categories. The molten salt provides a heat transfer medium, prevents thermal surges, and functions as an in situ scrubber to transform the acid-gas forming components of the waste into neutral salts and immobilizes potentially fugitive materials by a combination of particle wetting, encapsulation and chemical dissolution and solvation. Because the offgas is collected and assayed before release, and wastes containing toxic and radioactive materials are treated while immobilized in a condensed phase, the process avoids the problems sometimes associated with incineration processes. We are studying a potentially improved modification of this process, which treats oxidizable wastes in two stages: pyrolysis followed by catalyzed molten salt oxidation of the pyrolysis gases at ca. 700 degrees C. 15 refs., 5 figs., 1 tab

Modeling of waste/near field interactions for a waste repository in bedded salt: the Dynamic Network (DNET) model

International Nuclear Information System (INIS)

Cranwell, R.M.

1983-01-01

The Fuel Cycle Risk Analysis Division of Sandia National Laboratories has been funded by the US Nuclear Regulatory Commission to develop a methodology for use in assessing the long-term risk from the disposal of radioactive wastes in deep geologic formations. As part of this program, the Dynamic Network (DNET) model was developed to investigate waste/near field interactions associated with the disposal of radioactive wastes in bedded salt formations. The model is a quasi-multi-dimensional network model with capabilities for simulating processes such as fluid flow, heat transport, salt dissolution, salt creep, and the effects of thermal expansion and subsedence on the rock units surrounding the repository. The use of DNET has been demonstrated in the analysis of a hypothetical disposal site containing a bedded salt formation as the host medium for the repository. An example of this demonstration analysis is discussed. Furthermore, the outcome of sensitivity analyses performed on the DNET model are presented

Influence of deformation on the fluid transport properties of salt rocks

NARCIS (Netherlands)

Peach, C.J.

1991-01-01

While the fluid transport properties of rocks are well understood under hydrostatic conditions, little is known regarding these properties in rocks undergoing crystal plastic deformation. However, such data are needed as input in the field of radioactive waste disposal in salt formations. They

Exothermic potential of sodium nitrate salt cake

International Nuclear Information System (INIS)

Beitel, G.A.

1977-06-01

High-Level radioactive liquid waste is being reduced to a liquid slurry by an evaporation and crystallization process and stored in the existing single-shell tanks. Continuous pumping of the waste storage tank will reduce the present 30 to 50% moisture to the minimum possible. The reduced waste is a relatively immobile salt cake consisting predominantly of sodium nitrate (NaNO 3 ) with lesser amounts of sodium nitrite (NaNO 2 ), sodium metaaluminate (NaAlO 2 ), and sodium hydroxide (NaOH). Trace amounts of fission products, transuranics, and a broad spectrum of organic materials in small but unknown amounts are also present. A program was initiated in 1973 to determine whether or not conditions exist which could lead to an exothermic reaction in the salt cake. Results of the latest series of tests conducted to determine the effects of mass and pressure are summarized. Hanford salt cake, as stored, cannot support combustion, and does not ignite when covered with a burning volatile hydrocarbon

Mobile calcination and cementation unit for solidification of concentrated radioactive wastes

International Nuclear Information System (INIS)

Napravnik, J.; Sazavsky, P.; Skaba, V.; Skvarenina, R.; Ditl, P.

1985-01-01

Mobile experimental unit MESA-1 was developed and manufactured for processing radioactive concentrates by direct cementation. The unit is mainly designed for processing low-level liquid wastes from nuclear power plants and other nuclear installations, in which the level of radioactivity does not exceed 10 10 Bq/m 3 , the salt content of liquid solutions does not exceed 500 kg/m 3 and the maximum amount of boric acid is 130 kg/m 3 . The equipment is built into three modules which may be assembled and dismantled in a short time and transported separately. The unit without the calciner module was tested in non-radioactive mode and in operation with actual radioactive wastes from the V-1 nuclear power plant. The course and results of the tests are described in detail. All project design values were achieved, a total of 18 dm 3 model solutions were processed and 1 m 3 of actual wastes with a salt content of 450 kg/m 3 . The test showed that with regard to the radiation level reached it will be necessary in the process of calcination to increase the shielding of certain exposed points. The calciner module is being assembled for completion. (Z.M.)

Brine migration in salt and its implications in the geologic disposal of nuclear waste

International Nuclear Information System (INIS)

Jenks, G.H.; Claiborne, H.C.

1981-12-01

This report respresents a comprehensive review and analysis of available information relating to brine migration in salt surrounding radioactive waste in a salt repository. The topics covered relate to (1) the characteristics of salt formations and waste packages pertinent to considerations of rates, amounts, and effects of brine migration, (2) experimental and theoretical information on brine migration, and (3) means of designing to minimize any adverse effects of brine migration. Flooding, brine pockets, and other topics were not considered, since these features will presumably be eliminated by appropriate site selection and repository design. 115 references

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

Size measurement of radioactive aerosol particles in intense radiation fields using wire screens and imaging plates

Energy Technology Data Exchange (ETDEWEB)

Oki, Yuichi; Tanaka, Toru; Takamiya, Koichi; Ishi, Yoshihiro; UesugI, Tomonori; Kuriyama, Yasutoshi; Sakamoto, Masaaki; Ohtsuki, Tsutomu [Kyoto University Research Reactor Institute, Osaka (Japan); Nitta, Shinnosuke [Graduate School of Engineering, Kyoto University, Kyoto (Japan); Osada, Naoyuki [Advanced Science Research Center, Okayama University, Okayama (Japan)

2016-09-15

Very fine radiation-induced aerosol particles are produced in intense radiation fields, such as high-intensity accelerator rooms and containment vessels such as those in the Fukushima Daiichi nuclear power plant (FDNPP). Size measurement of the aerosol particles is very important for understanding the behavior of radioactive aerosols released in the FDNPP accident and radiation safety in high-energy accelerators. A combined technique using wire screens and imaging plates was developed for size measurement of fine radioactive aerosol particles smaller than 100 nm in diameter. This technique was applied to the radiation field of a proton accelerator room, in which radioactive atoms produced in air during machine operation are incorporated into radiation-induced aerosol particles. The size of 11C-bearing aerosol particles was analyzed using the wire screen technique in distinction from other positron emitters in combination with a radioactive decay analysis. The size distribution for 11C-bearing aerosol particles was found to be ca. 70 μm in geometric mean diameter. The size was similar to that for 7Be-bearing particles obtained by a Ge detector measurement, and was slightly larger than the number-based size distribution measured with a scanning mobility particle sizer. The particle size measuring method using wire screens and imaging plates was successfully applied to the fine aerosol particles produced in an intense radiation field of a proton accelerator. This technique is applicable to size measurement of radioactive aerosol particles produced in the intense radiation fields of radiation facilities.

Influence of deformation on the fluid transport properties of salt rocks

NARCIS (Netherlands)

Peach, C.J.

1991-01-01

While the fluid transport properties of rocks are well understood under hydrostatic conditions, little is known regarding these properties in rocks undergoing crystal plastic deformation. However, such data are needed as input in the field of radioactive waste disposal in salt formations. They are

Screening specifications for Gulf Coast salt domes

International Nuclear Information System (INIS)

Brunton, G.D.; Laughon, R.B.; McClain, W.C.

1978-01-01

A reconnaissance survey of the salt domes of Mississippi, Louisiana, and east Texas is being planned to identify study areas for potential sites for radioactive waste disposal. Preliminary screening specifications were derived for each of the geological evaluation criteria by application of the significant factors that will have an impact on the reconnaissance survey. The procedure for the derivation of each screening specification is discussed. The screening specifications are the official OWI values to be used for the first-cut acceptance for salt dome study areas along the Gulf Coast. The derivation of the screening specifications is illustrated by (1) a statement of the geological evaluation criterion, (2) a discussion of the pertinent factors affecting the criterion, and (3) the evaluation of the value of the specification

Salt deposits of Los Medanos Area, Eddy and Lea counties, New Mexico

International Nuclear Information System (INIS)

Jones, C.L.; Cooley, M.E.; Bachman, G.O.

1973-01-01

The salt deposits of Los Medanos area, in Eddy and Lea Counties, southeastern New Mexico, are being considered for possible use as a receptacle for radioactive wastes in a pilot-plant repository. The salt deposits of the area are in three evaporite formations: the Castile, Salado, and Rustler formations, in ascending order. The three formations are dominantly anhydrite and rock salt; but some gypsum, potassium ores, carbonate rock, and fine-grained clastic rocks are present. They have combined thicknesses of slightly more than 4000 feet, of which roughly one-half belongs to the Salado. Both the Castile and the Rustler are richer in anhydrite and poorer in rock salt than the Salado, and they provide this salt-rich formation with considerable protection from any fluids which might be present in underlying or overlying rocks. The Salado Formation contains many thick seams of rock salt at moderate depths below the surface. The rock salt has a substantial cover of well-consolidated rocks, and it is very little deformed structurally. 37 refs., 48 figs., 4 tabs

Management of radioactive wastes - an overview of the Indian programme

International Nuclear Information System (INIS)

Thomas, K.T.; Sunder Rajan, N.S.; Balu, K.; Khan, A.A.

1977-01-01

An overview of the management of radioactive wastes with particular reference to the Indian Nuclear Programme is presented. The initial design philosophy of the radwaste management system is discussed in relation to accepting a calculated, minimum discharge of radioactivity to the environment. A brief report of the operational experience with the low and intermediate level radwaste systems is given. Factors that influence the review of the present philosophy for future adoption are presented. Some methods being developed for decreasing release of the radioactivity to the environment are discussed. Among techniques considered are solar evaporation, delay and decay of fission rare gases from power reactors and concentration and storage of Kr 85 from fuel reprocessing plants. Problems in the management of high level and alph-bearing wastes are discussed with particular reference to the nature of the waste generated and the policy under implementation for their management. The matrices, solidification processes, modes of interim storage and criteria for selection of site for ultimate dispensation of the solidified high level wastes in geological formations are described. An approach towards the solution of the probelm of management of alpha-bearing waste is also presented

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

Low- and intermediate level radioactive waste from Risoe, Denmark. Location studies for potential disposal areas. Report no. 11. Description of areas. Danish and English summary

International Nuclear Information System (INIS)

Gravesen, P.; Nilsson, B.; Schack Pedersen, S.A.; Binderup, M.

2011-01-01

The low - and intermediate level radioactive waste from Risoe: the nuclear reactor buildings, different types of material from the research periods and waste from hospitals and research institutes have to be stored in a final disposal in Denmark for at least 300 years. The task is to locate and recognize sediments or rocks with low permeability which can isolate the radioactive waste from the surrounding deposits, the groundwater resources, the recipients and from human activities. The sediments or rocks shall also act as a protection if the waste disposal leaks radioactive material to the surroundings. This goal can be reached by choosing deposits with low water flow and high sorption potential of the sediments or rocks. The investigation of geological deposits as potential waste disposals for high radioactive waste from nuclear power plants has earlier focused on deep seated salt deposits and basement rocks but the Tertiary clays were also mapped. The salt diapirs, salt pillows and salt deposits and deep basement rocks are not included in the present study. These rocks and deposits are situated too deep for the present study and salt deposits seem to be unstable for a disposal (e.g. German salt mines). The regional geologic survey based on existing data was concluded by selecting 22 areas in Denmark. There remains now to reduce the number of potential areas to 1-3 where detailed field studies will be performed in order to select the final location. (LN)

Environmental aspects of produced-water salt releases in onshore and coastal petroleum-producing areas of the conterminous U.S. - a bibliography

Science.gov (United States)

Otton, James K.

2006-01-01

Environmental effects associated with the production of oil and gas have been reported since the first oil wells were drilled in the Appalachian Basin in Pennsylvania and Kentucky in the early to mid-1800s. The most significant of these effects are the degradation of soils, ground water, surface water, and ecosystems they support by releases of suspended and dissolved hydrocarbons and co-produced saline water. Produced water salts are less likely than hydrocarbons to be adsorbed by mineral phases in the soil and sediment and are not subject to degradation by biologic processes. Sodium is a major dissolved constituent in most produced waters and it causes substantial degradation of soils through altering of clays and soil textures and subsequent erosion. Produced water salts seem to have the most wide-ranging effects on soils, water quality, and ecosystems. Trace elements, including boron, lithium, bromine, fluorine, and radium, also occur in elevated concentrations in some produced waters. Many trace elements are phytotoxic and are adsorbed and may remain in soils after the saline water has been flushed away. Radium-bearing scale and sludge found in oilfield equipment and discarded on soils pose additional hazards to human health and ecosystems. This bibliography includes studies from across the oil- and natural-gas-producing areas of the conterminous United States that were published in the last 80 yrs. The studies describe the effects of produced water salts on soils, water quality, and ecosystems. Also included are reports that describe (1) the inorganic chemistry of produced waters included in studies of formation waters for various purposes, (2) other sources of salt affecting water quality that may be mistaken for produced water effects, (3) geochemical and geophysical techniques that allow discrimination of salt sources, (4) remediation technologies designed to repair damage caused to soils and ground water by produced water salts, and (5) contamination by

Small zeolite column tests for removal of cesium from high radioactive contaminated water in Fukushima Daiichi Nuclear Power Station

International Nuclear Information System (INIS)

Hijikata, Takatoshi; Uozumi, Koichi; Tukada, Takeshi; Koyama, Tadafumi; Ishikawa, Keiji; Ono, Shoichi; Suzuki, Shunichi; Denton, Mark; Raymont, John

2011-01-01

After the earthquake on March 11th 2011, a large amount (more than 0.12 million m 3 ) of highly radioactive contaminated water had pooled in Fukushima Daiichi nuclear power station. As an urgent issue, highly radioactive nuclides should be removed from this contaminated water to reduce radioactivity in the turbine buildings and nuclear reactor buildings. Removal of Cs from this contaminated water is a key issue, because 134 Cs and 137 Cs are highly radioactive γ-emitting nuclides. The zeolite column system was used for Cs and Sr removal from the radioactive water of Three-Mile Island Unit 2, and modified columns were then developed as a Cs removal method for high-level radioactive water in US national laboratories (WRSC, ORNL, PNNL, Hanford, etc.). In order to treat Fukushima's highly contaminated water with a similar system, it was necessary to understand the properties of zeolite to remove Cs from sea salt as well as the applicability of the column system to a high throughput of around 1200 m 3 /d. The kinetic characteristics of the column were another property to be understood before actual operation. Hence, a functional small-scale zeolite column system was installed in CRIEPI for conducting the experiments to understand decontamination behaviors. Each column has a 2- or 3-cm inner diameter and a 12-cm height, and 12 g of zeolite-type media was packed into the column. The column experiments were carried out with Kurion-zeolite, Herschelite, at different feed rates of simulated water with different concentrations of Cs and sea salt. As for the water with 4 ppm Cs and 0 ppm sea salt, only a 10% Cs concentration was observed in the effluent after 20,000 bed volumes were fed at a rate of 33 cm/min, which corresponds to the actual system. On the other hand, a 40% Cs concentration was observed in the effluent after only 50 bed volumes were passed for water with 2 ppm Cs and 3.4 wt.% sea salt at a feed rate of 34 cm/min. As the absorption of Cs is hampered by the

Chemical decontamination method for radioactive metal waste

International Nuclear Information System (INIS)

Onuma, Tsutomu; Akimoto, Hidetoshi

1991-01-01

The invention relates to a decontamination method for radioactive metal waste products derived from equipment that handles radioactive materials whose surfaces have been contaminated; in particular it concerns a decontamination method that reduces the amount of radioactive waste by decontaminating radioactive waste substances to a level of radioactivity in line with normal waste products. In order to apply chemical decontamination to metal waste products whose surfaces are divided into carbon steel waste and stainless steel waste; the carbon steel waste is treated using only a primary process in which the waste is immersed in a sulfuric acid solution, while the stainless steel waste must be treated with both the primary process and then electrolytically reduces it for a specific length of time and a secondary process that uses a solution of sulfuric acid mixed with oxidizing metal salts. The method used to categorize metal waste into carbon steel waste and stainless steel waste involves determining the presence, or absence, of magnetism. Voltage is applied for a fixed duration; once that has stopped, electrolytic reduction repeats the operative cycle of applying, then stopping voltage until the potential of the radioactive metal waste is retained in the active region. 1 fig. 2 tabs

Water-bearing explosives thickened with a partially hydrolyzed acrylamide polymer

Energy Technology Data Exchange (ETDEWEB)

Lyerly, W.M.

1971-11-23

Thickened water-bearing explosives are provided which do not segregate and are water-resistant over a wide range of viscosities. Preferred compositions have a unique combination of pourability and fluidity coupled with resistance to water and segregation which makes them particularly suitable in small diameter holes and in holes partially filled with water. Accordingly, water-bearing explosive compositions also are provided which consist of inorganic oxidizing salt, fuel, and water, which improvement consists of thickening the compositions with the combination of polyacrylamide and cross-linked galactomannan. The weight ratio of the polyacrylamide to galactomanan is from about ratio 0.1:1 to 10:1, and preferably 1:1 to 5:1. (1 claim)

Schematic designs for penetration seals for a reference repository in bedded salt

International Nuclear Information System (INIS)

Kelsall, P.C.; Case, J.B.; Meyer, D.; Coons, W.E.

1982-11-01

The isolation of radioactive wastes in geologic repositories requires that man-made penetrations such as shafts, tunnels, or boreholes are adequately sealed. This report describes schematic seal designs for a repository in bedded salt referenced to the straitigraphy of southeastern New Mexico. The designs are presented for extensive peer review and will be updated as site-specific conceptual designs when a site for a repository in salt has been selected. The principal material used in the seal system is crushed salt obtained from excavating the repository. It is anticipated that crushed salt will consolidate as the repository rooms creep close to the degree that mechanical and hydrologic properties will eventually match those of undisturbed, intact salt. For southeastern New Mexico salt, analyses indicate that this process will require approximately 1000 years for a seal located at the base of one of the repository shafts (where there is little increase in temperature due to waste emplacement) and approximately 400 years for a seal located in an access tunnel within the repository. Bulkheads composed of contrete or salt bricks are also included in the seal system as components which will have low permeability during the period required for salt consolidation

The HAW project: demonstration facility for the disposal of high-level waste in salt

International Nuclear Information System (INIS)

Rothfuchs, T.

1991-01-01

This publication is the interim report 1988-89 of the international HAW project performed in the 800 m level of the Asse salt mine in the Federal Republic of Germany. The major objective of this project is the pilot testing and demonstration of safe methods for the final disposal of high-level radioactive waste in geological salt deposits. The HAW-project is carried out by the GSF-Institut fuer Tieflagerung (IFT) in cooperation with the French Agence Nationale pour la Gestion des Dechets Radioactifs (ANDRA); the Spanish Empresa Nacional de Residuos Radiactivos S.A. (ENRESA) and the Netherlands Energy Research Foundation (ECN). After some delays in the licensing procedure the emplacement of 30 vitrified highly radioactive canisters (containers) is now envisaged for early 1991. 20 refs.; 92 figs.; 14 tabs

Distribution of various water soluble radioactive metalloporphyrins in tumor bearing mice

International Nuclear Information System (INIS)

Hambright, P.; Fawwaz, R.; Valk, P.; McRae, J.; Bearden, A.J.

1975-01-01

The distribution of a variety of water soluble 109 Pd and 64 Cu porphyrins were studied in mice bearing three types of tumors. While the metalloporphyrins are found to have an affinity for neoplastic tissue, substantial extra-tumor concentrations are also noted. Although this limits their value as specific tumor imaging agents, their use in localized therapy is discussed

Savannah River Site - Salt-stone Disposal Facility Performance Assessment Update

International Nuclear Information System (INIS)

Newman, J.L.

2009-01-01

The Savannah River Site (SRS) Salt-stone Facility is currently in the midst of a Performance Assessment revision to estimate the effect on human health and the environment of adding new disposal units to the current Salt-stone Disposal Facility (SDF). These disposal units continue the ability to safely process the salt component of the radioactive liquid waste stored in the underground storage tanks at SRS, and is a crucial prerequisite for completion of the overall SRS waste disposition plan. Removal and disposal of low activity salt waste from the SRS liquid waste system is required in order to empty tanks for future tank waste processing and closure operations. The Salt-stone Production Facility (SPF) solidifies a low-activity salt stream into a grout matrix, known as salt-stone, suitable for disposal at the SDF. The ability to dispose of the low-activity salt stream in the SDF required a waste determination pursuant to Section 3116 of the Ronald Reagan National Defense Authorization Act of 2005 and was approved in January 2006. One of the requirements of Section 3116 of the NDAA is to demonstrate compliance with the performance objectives set out in Subpart C of Part 61 of Title 10, Code of Federal Regulations. The PA is the document that is used to ensure ongoing compliance. (authors)

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

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

International Nuclear Information System (INIS)

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

1978-01-01

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

Salt mine Asse II. Status of the retrieval activities

International Nuclear Information System (INIS)

2017-02-01

The booklet on the status of retrieval activities in the salt mine Asse II includes information on the background of medium-level radioactive waste disposal during 1967 to 1978 on behalf of the Federal government. Since 2009 the former mine is operated by the BfS with the assignment of decommissioning. The potential risk for stability and safety due to problems of water ingress were known before beginning of the disposals. The retrieval of the radioactive waste will require many decades; the costs are financed by tax money. The planning of the retrieval is currently on the way, details of the concept are described.

Concentration and solidification of liquid radioactive wastes. Laboratory studies

International Nuclear Information System (INIS)

Nuche Vazquez, F.; Lora Soria, F. de

1969-01-01

Bench scale runs on concentration of intermediate level radioactive wastes, and incorporation of the concentrates in asphalt, are described. The feasibility of the process has been demonstrated, with a maximum incorporation of 60 percent of salts into the asphaltic matrix and a volume reduction factor of 10. (Author) 14 refs

The use of ceramic membranes for radioactive solutions purification

International Nuclear Information System (INIS)

Zakrzewska-Trznadel, G.

2002-01-01

Membrane permeation combined with complexation was tested for radioactive wastes processing purpose. The results of experiments with MEMBRALOX and CeRAM INSIDE filtering elements are presented in the paper. The pore size of ceramic membranes was in 1kD-100 nm range. The experiments were performed with non-active and with radioactive model solutions and original radioactive waste samples. To achieve high decontamination factors the process was enhanced by chemical complexation. Such complexants as poly(acrylic) acid and polyacrylic)acid salts of different crosslinking, polyethylenimine and cyanoferrates were tested. The experiments showed the significant increase of retention and decontamination factors while before ultrafiltration macromolecular ligands were added. The effectiveness of complexation by each ligand is strongly dependent on pH and alkali metals concentration. (author)

Salt decontamination demonstration test results

International Nuclear Information System (INIS)

Snell, E.B.; Heng, C.J.

1983-06-01

The Salt Decontamination Demonstration confirmed that the precipitation process could be used for large-scale decontamination of radioactive waste sale solution. Although a number of refinements are necessary to safely process the long-term requirement of 5 million gallons of waste salt solution per year, there were no observations to suggest that any fundamentals of the process require re-evaluation. Major accomplishments were: (1) 518,000 gallons of decontaminated filtrate were produced from 427,000 gallons of waste salt solution from tank 24H. The demonstration goal was to produce a minimum of 200,000 gallons of decontaminated salt solution; (2) cesium activity in the filtrate was reduced by a factor of 43,000 below the cesium activity in the tank 24 solution. This decontamination factor (DF) exceeded the demonstration goal of a DF greater than 10,000; (3) average strontium-90 activity in the filtrate was reduced by a factor of 26 to less than 10 3 d/m/ml versus a goal of less than 10 4 d/m/ml; and (4) the concentrated precipitate was washed to a final sodium ion concentration of 0.15 M, well below the 0.225 M upper limit for DWPF feed. These accomplishments were achieved on schedule and without incident. Total radiation exposure to personnel was less than 350 mrem and resulted primarily from sampling precipitate slurry inside tank 48. 3 references, 6 figures, 2 tables

Crystallization of sodium nitrate from radioactive waste

International Nuclear Information System (INIS)

Krapukhin, V.B.; Krasavina, E.P.; Pikaev, A.K.

1997-07-01

From the 1940s to the 1980s, the Institute of Physical Chemistry of the Russian Academy of Sciences (IPC/RAS) conducted research and development on processes to separate acetate and nitrate salts and acetic acid from radioactive wastes by crystallization. The research objective was to decrease waste volumes and produce the separated decontaminated materials for recycle. This report presents an account of the IPC/RAS experience in this field. Details on operating conditions, waste and product compositions, decontamination factors, and process equipment are described. The research and development was generally related to the management of intermediate-level radioactive wastes. The waste solutions resulted from recovery and processing of uranium, plutonium, and other products from irradiated nuclear fuel, neutralization of nuclear process solutions after extractant recovery, regeneration of process nitric acid, equipment decontamination, and other radiochemical processes. Waste components include nitric acid, metal nitrate and acetate salts, organic impurities, and surfactants. Waste management operations generally consist of two stages: volume reduction and processing of the concentrates for storage, solidification, and disposal. Filtration, coprecipitation, coagulation, evaporation, and sorption were used to reduce waste volume. 28 figs., 40 tabs

Re-evaluation of salt deposits. BGR investigates subhorizontally-bedded salt layers; Salzvorkommen neu bewertet. BGR untersucht flach lagernde salinare Schichten

Energy Technology Data Exchange (ETDEWEB)

Hammer, Joerg [Bundesanstalt fuer Geowissenschaften und Rohstoffe, Hannover (Germany). Fachbereich ' ' Geologisch-geotechnische Erkundung' ' ; Fahland, Sandra [Bundesanstalt fuer Geowissenschaften und Rohstoffe, Hannover (Germany). Fachberech ' ' Geotechnische Sicherheitsnachweise' '

2016-05-15

The search for a site for a repository for high-level radioactive waste was restarted in 2013. All of the potential host rocks existing in Germany must be re-evaluated and compared as a result. The list now also includes so-called ''subhorizontally-bedded evaporite formations''. BGR is analysing today's knowledge base on these salt deposits as part of the BASAL project.

Test Plan for Radioactive Testing of a Vertical Direct Denitration Calciner

International Nuclear Information System (INIS)

COMPTON, J.A.

1999-01-01

A vertical calciner will be used to demonstrate the direct denitration process for converting plutonium-bearing liquors to stable plutonium rich solids. The calciner and some of its associated equipment were previously tested with non-radioactive chemicals to demonstrate operability

Test Plan for Radioactive Testing of a Vertical Direct Denitration Calciner

Energy Technology Data Exchange (ETDEWEB)

COMPTON, J.A.

1999-10-05

A vertical calciner will be used to demonstrate the direct denitration process for converting plutonium-bearing liquors to stable plutonium rich solids. The calciner and some of its associated equipment were previously tested with non-radioactive chemicals to demonstrate operability.

Molten salt oxidation as a technique for decommissioning: selection of low melting point salt mixtures

International Nuclear Information System (INIS)

Lainetti, Paulo E.O.; Garcia, Vitor F.; Benvegnu, Guilherme

2013-01-01

During the 70 and 80 years, IPEN built several facilities in pilot scale, destined to the technological domain of the Nuclear Fuel Cycle. In the nineties, radical changes in the Brazilian nuclear policy determined the interruption of the activities and the shut-down of pilot plants. Nowadays, IPEN has been facing the problem of the dismantling and decommissioning of its Nuclear Fuel Cycle old facilities. The facility CELESTE-I of the IPEN is a laboratory where reprocessing studies were accomplished during the decade of 80 and in the beginning of the 90s. The last operations occurred in 92-93. The research activities generated radioactive wastes in the form of organic and aqueous solutions of different compositions and concentrations. For the treatment of these liquid wastes it was proposed a study of waste thermal decomposition based on the molten salt oxidation process.Decomposition tests of different organic wastes have been performed in laboratory equipment developed at IPEN, in the range of temperatures of 900 to 1020 deg C, demonstrating the complete oxidation of the compounds. The reduction of the process temperatures would be of crucial importance. Besides this, the selection of lower melting point salt mixtures would have an important impact in the reduction of equipment costs. Several experiments were performed to determine the most suitable salt mixtures, optimizing costs and melting temperatures as low as possible. This paper describes the main characteristics of the molten salt oxidation process, besides the selection of salt mixtures of binary and ternary compositions, respectively Na 2 CO 3 - NaOH and Na 2 CO 3 - K 2 CO 3 -Li 2 CO 3 . (author)

Analysis of 99Tc in the radioactive liquid waste after extraction into suitable solvent

International Nuclear Information System (INIS)

Sonar, N.L.; Vaishali De; Pardeshi, V.; Raghvendra, Y.; Valsala, T.P.; Sonavane, M.S.; Kulkarni, Y.; Raj Kanwar

2012-01-01

99 Tc is one of the long lived fission product with high fission yield. >From radioactive waste management point of view it is very much essential to evaluate the concentration of technetium in the radioactive liquid waste in order to finalise the treatment process to extract/isolate it from the stream which is discharged to the environment. For the estimation of 99 Tc in the radioactive liquid waste stream, extraction of the stable complex of technetium-tetraphenyl arsonium chloride (TPAC) into chloroform followed by beta counting was studied. Various parameters like pH, time of equilibration, concentration of TPAC in chloroform, use of other solvent for extraction as well as interference of various other radionuclides present in the waste were also studied. The radioactive liquid waste being handled in plant contains high concentrations of salts in the form of sodium nitrate. Hence effect of salt concentration on the percentage extraction was also evaluated. The extraction behavior does not dependent on change in the pH of the solution. Almost 99.5% extraction was observed in the pH range of 1-13.0. High concentration of salt is affecting the extraction. However, this can be taken care by diluting the radioactive waste. It takes almost 90 min time for maximum extraction. Presence of radionuclides like 137 Cs, 90 Sr are not interfering the extraction of 99 Tc. However, 106 Ru is getting slightly extracted along with 99 Tc. The error due to 106 Ru can be eliminated by taking gamma spectrum and deducting the activity from the total beta activity to get 99 Tc activity. Nitrobenzene can be used for extraction of Tc-TPAC complex in place of chloroform. (author)

Overview of insoluble radioactive cesium particles emitted from the Fukushima Dai-ichi Nuclear Power Station

Science.gov (United States)

Satou, Yukihiko

2017-04-01

In the early stage of the Fukushima Dai-ichi Nuclear Power Station (F1NPS) accident, number of spot type contamination has been observed in computed autoradiography (Kashimura 2013, Shibata 2013, Satou 2014). It's means presence of radioactive particles, however, insoluble cesium particle was overlooked because cesium, which is dominant radioactive element in the accident, becomes ionized in the environment. Adachi et al. (2013) showed presence of cesium (Cs)-bearing particles within air dust sample collected at Tsukuba, 170 km south from the Fukushima site, in midnight of 14 to morning of 15 March 2011. These particles were micrometer order small particles and Cs was could be detectable as element using an energy dispersive X-ray spectroscopy (EDX). However, other radioactive elements such as Co-60, Ru-103 and uranium, which were dominant element of radioactive particles delivered from Chernobyl accident, could not detected. Abe et al. (2014) employed a synchrotron radiation (SR)-micro(μ)-X-ray analysis to the Cs-bearing particles, and they were concluded that (1) contained elements derived from nuclear fission processes and from nuclear reactor and fuel materials; (2) were amorphous; (3) were highly oxidized; and (4) consisted of glassy spherules formed from a molten mixture of nuclear fuel and reactor material. In addition, Satou et al. (2016) and Yamaguchi et al. (2016) disclosed that silicate is main component of Cs-bearing particles. Satou et al. (2015) discovered two types of radioactive particles from soil samples collected in the vicinity of the F1NPS. These particles were remained in the natural environment more than four years, silicate is main component in common of each group particles. Group A particles were very similar to Cs-bearing particles reported by Adachi et al. except particle shape. On the other hand, group B is big particles found in north area from the F1NPS, and the strongest particles contained 20 kBq of Cs-137 within a particle

COSA II Further benchmark exercises to compare geomechanical computer codes for salt

International Nuclear Information System (INIS)

Lowe, M.J.S.; Knowles, N.C.

1989-01-01

Project COSA (COmputer COdes COmparison for SAlt) was a benchmarking exercise involving the numerical modelling of the geomechanical behaviour of heated rock salt. Its main objective was to assess the current European capability to predict the geomechanical behaviour of salt, in the context of the disposal of heat-producing radioactive waste in salt formations. Twelve organisations participated in the exercise in which their solutions to a number of benchmark problems were compared. The project was organised in two distinct phases: The first, from 1984-1986, concentrated on the verification of the computer codes. The second, from 1986-1988 progressed to validation, using three in-situ experiments at the Asse research facility in West Germany as a basis for comparison. This document reports the activities of the second phase of the project and presents the results, assessments and conclusions

Comparison of the salt domes Asse and Gorleben with regard to their suitability for the final storage of radoactive wastes

International Nuclear Information System (INIS)

Deisenroth, Norbert; Kokorsch, Rudolf

2012-01-01

In Germany, the search for a proper solution to the issue of final disposal of radioactive wastes is complicated by political leaders. The Gorleben moratorium from October 2000 delayed the proper solution unnecessary to ten years. Asse proves that salt domes such as Gorleben do not offer a permanent partitioning of the waste over the biosphere. With this in mind, the authors of the contribution under consideration compare the two salt domes Gorleben and Asse from a mining and geological point of view based on publicly available data with regard to their suitability for the disposal of radioactive waste.

Radioactivity in bottled waters sold in Mexico

International Nuclear Information System (INIS)

Davila Rangel, J.I.; Lopez del Rio, H.; Mireles Garcia, F.; Quirino Torres, L.L.; Villalba, M.L.; Colmenero Sujo, L.; Montero Cabrera, M.E.

2002-01-01

Measurements of gross alpha and beta activities were made on 21 domestic and international brands of bottled (purified and mineral) water sold in the Mexican market to assess its radiological quality. Alpha and beta activities were determined using a liquid-scintillation detector with pulse-shape analysis feature. All the purified water had values of beta activity lower than the limit for potable drinking water (1.0 Bq/l), while three brands surpassed the limit of alpha activity (0.1 Bq/l). The limit for alpha radioactivity content was exceed by three mineral waters; the results show a correlation between radioactivity content and mineral salts, which are related with the origin and treatment of the waters

Method of burying vessel containing radioactive waste

International Nuclear Information System (INIS)

Koga, Yoshihito.

1989-01-01

A float having an inert gas sealed therein is attached to a tightly closed vessel containing radioactive wastes. The vessel is inserted and kept in a small hole for burying the tightly closed vessel in an excavated shaft in rocks such as of granite or rock salts, while filling bentonite as shielding material therearound. In this case, the float is so adjusted that the apparent specific gravity is made equal or nearer between the tightly closed vessel and the bentonite, so that the rightly closed vessel does not sink and cause direct contact with the rocks even if bentonite flows due to earthquakes, etc. This can prevent radioactivity contamination through water in the rocks. (S.K.)

To the issue about negative consequences of underground nuclear explosions in the salt domes

International Nuclear Information System (INIS)

Belyashov, D.N.; Mokhov, V.A.; Murzadilov, T.D.

1998-01-01

I. From 1970 to 1984, 26 underground explosions were conducted at Azgir test site salt domes and Karachaganak gas-condensate deposit (KGKD) of Kazakhstan. Consequence, 9 and 6, relatively, underground cavities were created. At Azgir test site 5 cavities were filled by water and brines. Some of them were destroyed with surface spotting formation. It is noticed the spreading of radionuclides out of cavities bounds. At the KGKD gas-condensate is loaded into 4 cavities, another 2 cavities are in the accident condition, the last one (5TK) was filled by brine. There are characters of radioecological situation degradation above the last cavity. Radioactive logging in the cavity shown that the γ-activity of rock was increased more then 8 times in the distance of depths 0-64 m for 3 years. Apparently, outbreak of radioactive brines takes place along the zones of fissuring on the bound of casing tubes into the 5TK borehole and along enclosing rocks with sorption of radioactive isotopes in clay rocks. 2. There are examples of negative evolution of events at the Astrakhan gas-condensate deposit, where 15 nuclear cavities were created from 1980 to 1984 years. In 1986 year, 13 of them stopped to exist because of tectonic shearing, triggering by underground nuclear explosion in the salt dome. Many of them are flooded and they throw out the radioactive brines, reaching the surface. 3. Negative development of radioecological situation is occurred because of depressurization of cavities, their flooding, displacement of radionuclides with salt into the brines, destroying of cavities, extrusion of radioactive brines along the permeable zones, more often along the militant and observation boreholes. It is possible to spread of radioactive contamination along horizontal at the distance for l,5-3 km. In 2 years after the underground nuclear explosion at the Grachev oil deposit of Bashkiria radioactive tritium was detected in underground water and in the ground more then 3 km far from

Radioactive liquid water processing method

International Nuclear Information System (INIS)

Yasumura, Keijiro; Noda, Tetsuya; Kobayashi, Fumio.

1993-01-01

Alkaline earth metals and heavy metals are added to radioactive liquid wastes containing a surface active agent comprising alkali metal salts of higher fatty acids. These metals form metal soaps with the surface active agent dissolved in the liquid wastes and crystallized. The crystallized metal soaps are introduced to a filtering column filled with a burnable polymeric fibrous filtering material. The filtering material is burnt. This can remove the surface active agent to remove COD without using an active carbon. (T.M.)

Geothermal in situ experiments in the Asse salt-mine

International Nuclear Information System (INIS)

Kopietz, J.; Jung, R.

1978-01-01

The paper presents design and results of in situ experiments carried out by the Bundesanstat fuer Geowissenschaften und Rohstoffe (Federal Institute for Geosciences and Natural Resources, F.R. of Germany) in the Asse salt-mine. With reference to model calculations of the temperature field which is produced in salt formations by radioactive waste, temperature measurements in the area of electrical heating elements and in situ measurements of thermal conductivity have been performed. The measured temperatures are in good accordance with the theoretical prediction. Preliminary results of the thermal conductivity measurements correspond with the data of single NaCl crystals published by Birch and Clark. At present a heating experiment is being conducted in the Asse mine to investigate thermo-mechanical effects of a cylindrical heat source upon the surrounding rock salt. Possible thermal induced fractures monitored by permeability changes and seismoacoustical phenomena are the main objects of this experiment

Waste salt recovery, recycle, and destruction

International Nuclear Information System (INIS)

Hickman, R.G.

1992-12-01

Starting in 1943 and continuing into the 1970s, radioactive wastes resulting from plutonium processing at Hanford were stored underground in 149 single shell tanks. Of these tanks, 66 are known or believedto be leaking, and over a period are believed to have leaked about 750,000 gal into the surrounding soil. The bulk of the aqueous solution has been removed and transferred to double shell tanks, none of which are leaking. The waste consists of 37 million gallons of salt cake and sludge. Most of the salt cake is sodium nitrate and other sodium salts. A substantial fraction of the sludge is sodium nitrate. Small amounts of the radionuclides are present in the sludge as oxides or hydroxides. In addition, some of the tanks contain organic compounds and ferrocyanide complexes, many of which have undergone radiolytic induced chemical changes during the years of storage. As part of the Hanford site remediation effort, the tank wastes must be removed, treated, and the residuals must be immobilized and disposed of in an environmentally acceptable manner. Removal methods of the waste from the tanks fall generally into three approaches: dry removal, slurry removal, and solution removed. The latter two methods are likely to result in some additional leakage to the surrounding soil, but that may be acceptable if the tank can be emptied and remediated before the leaked material permeates deeply into the soil. This effort includes three parts: salt splitting, acid separation, and destruction, with initial emphasis on salt splitting

Implications of sedimentological and hydrological processes on the distribution of radionuclides: the example of a salt marsh near Ravenglass, Cumbria

International Nuclear Information System (INIS)

Carr, A.P.; Blackley, M.W.L.

1986-01-01

This paper summarizes sedimentological and hydrological studies at a salt marsh site on the north bank of the River Esk near Ravenglass which have a bearing on the fate of the low-level radioactive effluent from the reprocessing facility at Sellafield, Cumbria. A range of techniques has been used including electromagnetic distance measurement (EDM) and pore water pressure studies. The results show that: (a) Over a two-year period there were no significant net changes in salt marsh creek level, although shorter-term (probably seasonal) fluctuations, of the order of 2 cm, occurred. These were attributed to expansion of clay particles during the winter months. Nearby, however, there were vertical changes of the order of 1 m due to erosion. (b) Pore water pressures indicated a dynamic situation with very rapid responses both to tidal fluctuations and to rainfall. During neap tides there was clear evidence for water seeping upwards from the underlying clay/sand interface. Shortlived radionuclides ( 95 Zr/ 95 Nb and 106 Ru) were detected in this zone. (c) soil polygons, once initiated by desiccation, thereafter provide preferential routes for water (and radionuclides) to the sub-surface sediment. These, and other results, are discussed in the context of previous studies. It is concluded that the complexity of the estuarine environment results in most data being site specific. (author)

Digestion of Alumina from Non-Magnetic Material Obtained from Magnetic Separation of Reduced Iron-Rich Diasporic Bauxite with Sodium Salts

Directory of Open Access Journals (Sweden)

Guanghui Li

2016-11-01

Full Text Available Recovery of iron from iron-rich diasporic bauxite ore via reductive roasting followed by magnetic separation has been explored recently. However, the efficiency of alumina extraction in the non-magnetic materials is absent. In this paper, a further study on the digestion of alumina by the Bayer process from non-magnetic material obtained after magnetic separation of reduced iron-rich diasporic bauxite with sodium salts was investigated. The results indicate that the addition of sodium salts can destroy the original occurrences of iron-, aluminum- and silicon-containing minerals of bauxite ore during reductive roasting. Meanwhile, the reactions of sodium salts with complex aluminum- and silicon-bearing phases generate diaoyudaoite and sodium aluminosilicate. The separation of iron via reductive roasting of bauxite ore with sodium salts followed by magnetic separation improves alumina digestion in the Bayer process. When the alumina-bearing material in bauxite ore is converted into non-magnetic material, the digestion temperature decreases significantly from 280 °C to 240 °C with a nearly 99% relative digestion ratio of alumina.

Health physics aspects of incineration of low level radioactive solvent at the Savannah River Plant

International Nuclear Information System (INIS)

Strain, C.D.

1987-01-01

This document contains the lecture notes and illustrations used in a presentation at the 1987 Health Physics Society Annual Meeting in Salt Lake City, Utah. Included is a description of the radioactive waste disposal facilities at the Savannah River Plant, South Carolina, and of the current use of this facility in incinerating thousands of gallons of radioactive waste. 12 figs

Deep underground exploration in the Asse salt mine

International Nuclear Information System (INIS)

Steinberg, S.; Schmidt, M.W.

1992-01-01

The activities reported here under the project task entitled ''Deep underground exploration up to the 925 m level'' opened up depths and salt formations in the Asse salt mine which are intended sites for R and D work for investigating and determining the conditions of radioactive waste disposal in a repository of the Gorleben type. The newly developed experimental levels will thus allow to directly apply research results obtained in the Asse mine to the Gorleben project. The activities reported included among other tasks work for increasing the depth of exploration in the Asse mine 2 down to 950 m, using a newly developed cutting method. The work was performed in cooperation with a mining corporation specializing in this sort of tasks. (orig.) With 18 maps [de

Calculation of density and permeability of compacted crushed salt within an engineered shaft sealing system

International Nuclear Information System (INIS)

Loken, M.; Statham, W.

1997-01-01

Crushed salt from the host Salado Formation is proposed as a sealing material in one component of a multicomponent seal system design for the shafts of the Waste Isolation Pilot Plant (WIPP), a mined geological repository for storage and disposal of transuranic radioactive wastes located near Carlsbad, New Mexico. The crushed salt will be compacted and placed at a density approaching 90% of the intact density of the host Salado salt. Creep closure of the shaft will further compact the crushed salt over time, thereby reducing the crushed-salt permeability from the initial state and creating an effective long-term seal. A structural model and a fluid flow model have been developed to provide an estimate of crushed-salt reconsolidation rate as a function of depth, time, and pore pressure. Model results are obtained in terms of crushed-salt permeability as a function of time and depth within the salt column. Model results indicate that average salt column permeability will be reduced to 3.3 x 10 -20 m 2 in about 100 years, which provides for an acceptable long-term seal component

Liquid radioactive waste concentration by the method of evaporation from porous plates

International Nuclear Information System (INIS)

Dmitriev, S.A.; Karlin, Yu.V.; Maryakhin, M.A.; Myasnikov, Yu.G.; Slastennikov, Yu.T.

2009-01-01

As it is shown by bench-scale experiments radioactive effluents are concentrated to salt content 319 g/l at temperature lower, than evaporation temperature of water, and specific power inputs lower, than specific evaporation heat of water by 20 times. Results of tests at pilot plant (productivity to 43 kg/h by evaporation water) that is placed in mobile water purification unit ECO are described. This unit is used for radioactive water treatment from different organizations at SPU Radon

Conceptual model for regional radionuclide transport from a salt dome repository: a technical memorandum

International Nuclear Information System (INIS)

Kier, R.S.; Showalter, P.A.; Dettinger, M.D.

1980-01-01

Disposal of high-level radioactive wastes is a major environmental problem influencing further development of nuclear energy in this country. Salt domes in the Gulf Coast Basin are being investigated as repository sites. A major concern is geologic and hydrologic stability of candidate domes and potential transport of radionuclides by groundwater to the biosphere prior to their degradation to harmless levels of activity. This report conceptualizes a regional geohydrologic model for transport of radionuclides from a salt dome repository. The model considers transport pathways and the physical and chemical changes that would occur through time prior to the radionuclides reaching the biosphere. Necessary, but unknown inputs to the regional model involve entry and movement of fluids through the repository dome and across the dome-country rock interface and the effect on the dome and surrounding strata of heat generated by the radioactive wastes

Hydrogeochemical radioactive features and prospecting in granopegmatite type uranium ore district in Danfeng area

International Nuclear Information System (INIS)

Feng Zhangsheng

2011-01-01

Hydrochemical radioactive prospecting plays an important role in the all stages of grano-pegmatite type uranium deposit exploration in Danfeng area dut to its fast, simple, economic and high effective advantage. Radioactive anomalous halo in the shallow underground water has identical distribution scopes with the ore-bearing biotite granite-pegmatite, which can be used to delineate uranium ore-forming prospective area, reconnaissance area and detailed prospecting area. Deep underground water close to the ore is characterized by hydrogeochemical radioactive features with high uranium and radon content. Through prospecting engineering of radioactive hydrogeochemical, the situation of blind ore bodies can be used to guide the layout. (authors)

Research and development action of the Commission of the European Communities (CEC) in the field of radioactive waste management

International Nuclear Information System (INIS)

Orlowski, S.; Bresesti, M.

1983-01-01

The CEC R and D action, started in 1973, is carried out within the framework of cost-sharing contracts with Community organizations and in the laboratories of the Joint Research Centre, Ispra. About 350 research workers from 30 organizations within the Community are taking part. The R and D activities cover processing, conditioning, characterization, intermediate storage and final disposal of the radioactive wastes generated in reactors and in fuel reprocessing and fuel fabrication plants. In the Community, spent fuels are not considered as radioactive waste. About one half of the total effort has been devoted to the disposal of high-level and long-lived radioactive wastes in geological formations (granite, clay, salt) and to related studies. The sub-seabed disposal option is also being investigated with a more limited effort. The R and D activities on waste treatment cover low-level, alpha-bearing and gaseous wastes. An important activity has been developed on the characterization of vitrified HLW. A similar activity for the characterization of other types of conditioned wastes has been started. The R and D activity of the CEC is supported by the existence of a Community Plan of Action (1980-1992) which entrusts to the Commission a wider role in the development of waste management policies. The Plan assures in particular the continuity of the R and D work up to 1992. International co-operation is considered important; international symposia have been co-sponsored with the IAEA; co-operative agreements with non-Community countries are in force (such as with Canada) or in preparation (such as with the USA). (author)

Environments with elevated radiation levels from natural radioactive substances

International Nuclear Information System (INIS)

Sohrabi, M.

2000-01-01

Some areas in the world have elevated levels of radioactive substances in the environment forming elevated radiation areas (ERAs) where public potential annual effective doses can exceed even the dose limit of radiation workers. Such radioactive substances are either terrestrial natural radioactivity added naturally in the soil or natural and/or man-made radioactivity from human activities added into the environment. If radioactivity is added naturally, elevated natural radiation areas (ENRAs) are formed. Based on the classification criteria introduced by the author, such regions are divided into static and dynamic areas. They are also classified in accordance with their level of potential effective dose to the public. Some main ENRAs are classified. Highlights are presented of the results of activity studies carried out in selected areas. The concepts discussed can also be applied to areas formed by human activities. The author suggests some guidelines for future studies, regulatory control and decision making, bearing in mind the need for harmonization of policies for regulatory control and remedial actions at sites to protect the public from environmental chronic exposures. (author)

Radioactive waste - a select list of material

International Nuclear Information System (INIS)

Lambert, C.M.

1982-01-01

A chronological bibliography is presented of literature relating to radioactive waste management in the United Kingdom concentrating on material published since 1978. The main sections include Dept. of Environ. and Official publications, administrative and environmental concerns, technological and scientific considerations, including publications on geological aspects, deep-sea bed and ocean-dumping and salt domes, with general background material and further sources of information listed at the end. (U.K.)

PHYSICAL, CHEMICAL AND STRUCTURAL EVOLUTIION OF ZEOLITE-CONTAINING WASTE FORMS PRODUCED FROM METAKAOLINITE AND CALCINED SODUIM BEARING WASTE (HLW AND/OR LLW)

International Nuclear Information System (INIS)

Grutzeck, Michael W.

2003-01-01

chemistries are entirely different. In addition to being vastly superior to conventional Portland cement grouts with respect to salt retention, standard radwaste leach protocols (PCT, TCLP) have shown that hydroceramics also do a better job of immobilizing the RCRA-toxic and radioactive components of liquid sodium bearing waste (SBW) now in storage at DOE's Hanford, Savannah River and Idaho sites

Electrodialysis-ion exchange for the separation of dissolved salts

Energy Technology Data Exchange (ETDEWEB)

Baroch, C.J. [Wastren, Inc., Westminster, CO (United States); Grant, P.J. [Wastren, Inc., Hummelstown, PA (United States)

1995-10-01

The Department of Energy generates and stores a significant quantity of low level, high level, and mixed wastes. As some of the DOE facilities are decontaminated and decommissioned, additional and possibly different forms of wastes will be generated. A significant portion of these wastes are aqueous streams containing acids, bases, and salts, or are wet solids containing inorganic salts. Some of these wastes are quite dilute solutions, whereas others contain large quantities of nitrates either in the form of dissolved salts or acids. Many of the wastes are also contaminated with heavy metals, radioactive products, or organics. Some of these wastes are in storage because a satisfactory treatment and disposal processes have not been developed. There is considerable interest in developing processes that remove or destroy the nitrate wastes. Electrodialysis-Ion Exchange (EDIX) is a possible process that should be more cost effective in treating aqueous waste steams. This report describes the EDIX process.

Thorium cycle and molten salt reactors: field parameters and field constraints investigations toward 'thorium molten salt reactor' definition

International Nuclear Information System (INIS)

Mathieu, L.

2005-09-01

Producing nuclear energy in order to reduce the anthropic CO 2 emission requires major technological advances. Nuclear plants of 4. generation have to respond to several constraints, as safety improvements, fuel breeding and radioactive waste minimization. For this purpose, it seems promising to use Thorium Cycle in Molten Salt Reactors. Studies on this domain have already been carried out. However, the final concept suffered from serious issues and was discontinued. A new reflection on this topic is being led in order to find acceptable solutions, and to design the Thorium Molten Salt Reactor concept. A nuclear reactor is simulated by the coupling of a neutron transport code with a materials evolution code. This allows us to reproduce the reactor behavior and its evolution all along its operation. Thanks to this method, we have studied a large number of reactor configurations. We have evaluated their efficiency through a group of constraints they have to satisfy. This work leads us to a better understanding of many physical phenomena controlling the reactor behavior. As a consequence, several efficient configurations have been discovered, allowing the emergence of new points of view in the research of Molten Salt Reactors. (author)

Summary of tank information relating salt well pumping to flammable gas safety issues

International Nuclear Information System (INIS)

Caley, S.M.; Mahoney, L.A.; Gauglitz, P.A.

1996-09-01

The Hanford Site has 149 single-shell tanks (SSTs) containing radioactive wastes that are complex mixes of radioactive and chemical products. Active use of these SSTs was phased out completely by November 1980, and the first step toward final disposal of the waste in the SSTs is interim stabilization, which involves removing essentially all of the drainable liquid from the tank. Stabilization can be achieved administratively, by jet pumping to remove drainable interstitial liquid, or by supernatant pumping. To date, 116 tanks have been declared interim stabilized; 44 SSTs have had drainable liquid removed by salt well jet pumping. Of the 149 SSTs, 19 are on the Flammable Gas Watch List (FGWL) because the waste in these tanks is known or suspected, in all but one case, to generate and retain mixtures of flammable gases, including; hydrogen, nitrous oxide, and ammonia. Salt well pumping to remove the drainable interstitial liquid from these SSTs is expected to cause the release of much of the retained gas, posing a number of safety concerns. The scope of this work is to collect and summarize information, primarily tank data and observations, that relate salt well pumping to flammable gas safety issues. While the waste within FGWL SSTs is suspected offering flammable gases, the effect of salt well pumping on the waste behavior is not well understood. This study is being conducted for the Westinghouse Hanford Company as part of the Flammable Gas Project at the Pacific Northwest National Laboratory (PNNL). Understanding the historical tank behavior during and following salt well pumping will help to resolve the associated safety issues

Identification of tissue sites for increased albumin degradation in sarcoma-bearing mice

International Nuclear Information System (INIS)

Andersson, C.; Iresjoe, B.M.L.; Lundholm, K.

1991-01-01

Plasma albumin concentration declines in both experimental and clinical cancer. Previous investigations have demonstrated that this is partly explained by increased breakdown of albumin. The present study has identified the tissue sites for increased albumin degradation in a nonmetastasizing sarcoma mouse (C57/BL6J) model. Results have been compared to nontumor-bearing animals either freely fed or food restricted (pair-weighed) so that their body composition was similar to tumor-bearing animals. Tumor-bearing mice had increased albumin degradation (0.13 +/- 0.02 mg/hr/g bw) compared to both freely fed (0.09 +/- 0.007) and pair-weighed control animals (0.05 +/- 0.008). Radioactivity from circulating [3H]raffine aldehyde labeled albumin appeared with maximum peak values in lysosomes isolated from both tumor and nontumor tissues at 48 hr following iv injection. The intralysosomal accumulation of radioactivity was two- to threefold higher in tumor tissue compared to liver tissue, although the specific activity of protease(s) for albumin degradation measured in vitro was not higher in tumor tissue (30.4 +/- 3.6 mg/hr/g tissue) compared to normal liver tissue (36.9 +/- 1.7). Accounting for the entire tumor the proteolytic capacity for albumin breakdown was however much larger in the tumor (161.6 +/- 32.6 mg/organ) compared to both normal liver (37.5 +/- 2.3) and tumor-host liver (56.4 +/- 2.8). Pepstatin inhibited 78 +/- 6% of the proteolytic activity in the tumor measured by 125I-labeled undenatured mouse albumin as the substrate. Leupeptin inhibited 49 +/- 6%. There was a significantly decreased breakdown of albumin in both skeletal muscles and the gastrointestinal tract from tumor-bearing animals

Textural and fluid phase analysis of rock salt subjected to the combined effects of pressure, heat and gamma radiation

International Nuclear Information System (INIS)

Huertas, F.; Major, J.C.; Del Olmo, C.

1992-01-01

The formation of colloidal sodium by radiolytic processes is a main concern with respect to the safety of disposal of high-level radioactive waste in salt formations. The research work seeks to assess the irradiation damage in natural rock salt when exposed to a different dose, dose rate, temperature and time of gamma irradiation. The work encompasses four major tasks: (i) detailed characterization of both solid and fluid phases of natural rock salt; (ii) gamma irradiation of salt samples; (iii) determination of the amount of colloidal sodium present in irradiated samples; (iv) calculation of radiation damage. 40 refs., 36 figs., 34 tabs

Steel corrosion in radioactive waste storage tanks

International Nuclear Information System (INIS)

Carranza, Ricardo M.; Giordano, Celia M.; Saenz, E.; Weier, Dennis R.

2004-01-01

A collaborative study is being conducted by CNEA and USDOE (Department of Energy of the United States of America) to investigate the effects of tank waste chemistry on radioactive waste storage tank corrosion. Radioactive waste is stored in underground storage tanks that contain a combination of salts, consisting primarily of sodium nitrate, sodium nitrite and sodium hydroxide. The USDOE, Office of River Protection at the Hanford Site, has identified a need to conduct a laboratory study to better understand the effects of radioactive waste chemistry on the corrosion of waste storage tanks at the Hanford Site. The USDOE science need (RL-WT079-S Double-Shell Tanks Corrosion Chemistry) called for a multi year effort to identify waste chemistries and temperatures within the double-shell tank (DST) operating limits for corrosion control and operating temperature range that may not provide the expected corrosion protection and to evaluate future operations for the conditions outside the existing corrosion database. Assessment of corrosion damage using simulated (non-radioactive) waste is being made of the double-shell tank wall carbon steel alloy. Evaluation of the influence of exposure time, and electrolyte composition and/or concentration is being also conducted. (author) [es

The treatment of radioactive waste in Institute of Nuclear Physics of Uzbekistan

International Nuclear Information System (INIS)

Radyuk, R.I.

2001-01-01

Full text: The main purpose of radioactive waste treatment is security of humanity and environment for future. The formation of radioactive waste in Institute of Nuclear Physics connects with scientific and research works on reactor and cyclotron. There are works in the field of radiochemistry, activation analysis, research of material. It is connected with some different materials used in practical work: mountain rock, food-stuffs, biological materials and other. The Institute of Nuclear Physics has enterprise, making radioactive isotopes. In consequence of this work radioactive wastes form. Average annual volume of liquid radioactive waste is 2000 m 3 in year. During normal work of nuclear reactor and enterprise of radioactive isotope small part of radionuclides with gaseous waste gets in environment. The content of inert gas does not exceed 2% of permissible level . Value of radionuclides fall out in area from 0.5 Km to 10 Km does not differ global fall out and changes from 1.1.10 6 Bq/km 2 to 1.6.10 7 Bq/km 2 month (permissible doze - 5.6.10 8 Bq/km 2 .month). The solid radioactive waste of medium and low activity are burying on Republic point of radioactive waste storage. Annual volume of solid radioactive waste is 60 m 3 in year and total radioactivity is 10 11 Bk. The solid radioactive waste of high activity are going to of Chelyabinsk. The liquid radioactive waste belong to second and third group of radioactive waste (classification of IAEA). The decontamination of liquid radioactive waste are made on the station of liquid radioactive waste treatment by method of sedimentation and distillation. The productivity of this plant is 15m 3 in day. Before treatment liquid radioactive waste is analyzed to determine chemical and radiochemical composition. It is solution with content of salt from 0.8 g/l to 15 g/l, salt Ca 2+ and Mg 2+ - 20 mg-eqv/l, oxygen - 100 mg O 2 /l , activity from 10 2 Bq/l to 10 4 Bq/l. The radionuclides composition of liquid radioactive

The HAW project: demonstration facility for the disposal of high-level waste in salt

International Nuclear Information System (INIS)

Rothfuchs, T.

1991-01-01

This report is the so-called Synthesis report 1985-1989 of the international HAW project performed in the 800 m level of the ASSE salt mine in the Federal Republic of Germany. The major objective of this project is the pilot testing and demonstration of safe methods for the final disposal of high-level radioactive waste in geological salt-deposits. The HAW-project is carried out by the GSF-Institut fuer Tieflagerung (IFT) in cooperation with the French Agence Nationale pour la Gestion des Dechets Radioactifs (ANDRA); the Spanish Empresa Nacional de Residuos Radioactivos S.A (ENRESA) and the Netherlands Energy Research Foundation (ECN). During the years 1985 to 1989 the underground test field was excavated and after some delays in the licensing procedure, the emplacement of 30 vitrified highly radioactive canisters (containers) is now envisaged for early 1991. 32 refs; 76 figs., 11 tabs

ISDP salt batch #2 supernate qualification

Energy Technology Data Exchange (ETDEWEB)

Peters, T. B. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Nash, C. A. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Fink, S. D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2009-01-05

This report covers the laboratory testing and analyses of the second Integrated Salt Disposition Project (ISDP) salt supernate samples, performed in support of initial radioactive operations of Actinide Removal Process (ARP) and Modular Caustic-Side Solvent Extraction Unit (MCU). Major goals of this work include characterizing Tank 22H supernate, characterizing Tank 41H supernate, verifying actinide and strontium adsorption with a standard laboratory-scale test using monosodium titanate (MST) and filtration, and checking cesium mass transfer behavior for the MCU solvent performance when contacted with the liquid produced from MST contact. This study also includes characterization of a post-blend Tank 49H sample as part of the Nuclear Criticality Safety Evaluation (NCSE). This work was specified by Task Technical Request and by Task Technical and Quality Assurance Plan (TTQAP). In addition, a sampling plan will be written to guide analytical future work. Safety and environmental aspects of the work were documented in a Hazard Assessment Package.

Improvements of Spiers model for compaction creep of crushed rock salt

International Nuclear Information System (INIS)

Poley, A.D.

1996-10-01

This report describes a number of improvements to the existing model for the process of compaction creep of rock salt developed by Spiers and co-workers. The process of compaction creep determines the behaviour of the seals of crushed rock salt, the last engineered barriers of a repository in rock salt for (radioactive) wastes. In Chapter 2 the derivation of the original model of Spiers and co-workers is followed except for some simplifying approximations. A comparison of the model results is made with experimental data and a number of model adjustments are suggested. In Chapter 3 one of these suggested model adjustments is explored, and an alternative model is developed. The results obtained with this model compare favourably with the experimental data without the use of adjustable shape functions as for the original model. Preliminary investigations of the impact of the new model on estimated releases to the geosphere of radionuclides form a repository in rock salt revealed striking differences: with the new model the compaction of the rock salt seals was so rapid that no releases could occur. The striking differences between the results - in terms of releases form a rock salt repository to the geosphere after groundwater intrusion - obtained using the two models clearly indicate the need for further experimental research into the end-compaction behaviour of rock salt backfill. (orig.)

Glass Formulation Development for INEEL Sodium-Bearing Waste

International Nuclear Information System (INIS)

Vienna, J.D.; Schweiger, M.J.; Smith, D.E.; Smith, H.D.; Crum, J.V.; Peeler, D.K.; Reamer, I.A.; Musick, C.A.; Tillotson, R.D.

1999-01-01

For about four decades, radioactive wastes have been collected and calcined from nuclear fuels reprocessing at the Idaho Nuclear Technology and Engineering Center (INTEC), formerly Idaho Chemical Processing Plant (ICPP). Over this time span, secondary radioactive wastes have also been collected and stored as liquid from decontamination, laboratory activities, and fuel-storage activities. These liquid wastes are collectively called sodium-bearing wastes (SBW). About 5.7 million liters of these wastes are temporarily stored in stainless steel tanks at the Idaho National Engineering and Environmental Laboratory (INEEL). Vitrification is being considered as an immobilization step for SBW with a number of treatment and disposal options. A systematic study was undertaken to develop a glass composition to demonstrate direct vitrification of INEEL's SBW. The objectives of this study were to show the feasibility of SBW vitrification, not a development of an optimum formulation. The waste composition is relatively high in sodium, aluminum, and sulfur. A specific composition and glass property restrictions, discussed in Section 2, were used as a basis for the development. Calculations based on first-order expansions of selected glass properties in composition and some general tenets of glass chemistry led to an additive (fit) composition (68.69 mass % SiO 2 , 14.26 mass% B 2 O 3 , 11.31 mass% Fe 2 O 3 , 3.08 mass% TiO 2 , and 2.67 mass % Li 2 O) that meets all property restrictions when melted with 35 mass % of SBW on an oxide basis, The glass was prepared using oxides, carbonates, and boric acid and tested to confirm the acceptability of its properties. Glass was then made using waste simulant at three facilities, and limited testing was performed to test and optimize processing-related properties and confirm results of glass property testing. The measured glass properties are given in Section 4. The viscosity at 1150 C, 5 Pa·s, is nearly ideal for waste-glass processing in

Salt Repository Project transportation program plan

International Nuclear Information System (INIS)

Fisher, R.L.; Greenberg, A.H.; Anderson, T.L.; Yates, K.R.

1987-01-01

The Salt Repository Project (SRP) has the responsibility to develop a comprehensive transportation program plan (TrPP) that treats the transportation of workers, supplies, and high-level radioactive waste to the site and the transportation of salt, low-level, and transuranic wastes from the site. The TrPP has developed a systematic approach to transportation which is directed towards satisfying statutes, regulations, and directives and is guided by a hierarchy of specific functional requirements, strategies, plans, and reports. The TrPP identifies and develops the planning process for transportation-related studies and provides guidance to staff in performing and documenting these activities. The TrPP also includes an explanation of the responsibilities of the organizational elements involved in these transportation studies. Several of the report chapters relate to identifying routes for transporting nuclear waste to the site. These include a chapter on identifying an access corridor for a new rail route leading to the site, identifying and evaluating emergency-response preparedness capabilities along candidate routes in the state, and identifying alternative routes from the state border, ports, or in-state reactors to the site. The TrPP also includes plans for identifying salt disposal routes and a discussion of repository/transportation interface requirements. 89 refs., 6 figs

Preliminary analyses of scenarios for potential human interference for repositories in three salt formations

International Nuclear Information System (INIS)

1985-10-01

Preliminary analyses of scenarios for human interference with the performance of a radioactive waste repository in a deep salt formation are presented. The following scenarios are analyzed: (1) the U-Tube Connection Scenario involving multiple connections between the repository and the overlying aquifer system; (2) the Single Borehole Intrusion Scenario involving penetration of the repository by an exploratory borehole that simultaneously connects the repository with overlying and underlying aquifers; and (3) the Pressure Release Scenario involving inflow of water to saturate any void space in the repository prior to creep closure with subsequent release under near lithostatic pressures following creep closure. The methodology to evaluate repository performance in these scenarios is described and this methodology is applied to reference systems in three candidate formations: bedded salt in the Palo Duro Basin, Texas; bedded salt in the Paradox Basin, Utah; and the Richton Salt Dome, Mississippi, of the Gulf Coast Salt Dome Basin

Biodistribution and receptor imaging studies of insulin labelled with radioiodine in mice bearing H22 hepatocellular cacinoma

International Nuclear Information System (INIS)

Tang Gongshun; Kuang Anren; Liang Zenlu

2004-01-01

Objectives: It has been demonstrated that insulin receptor of hepatocellular carcinoma cells is overexpression. The biodistribution of 125I-insulin and receptor imaging studies of 131I-insulin in mice bearing solid liver tumor comprised of hepatic carcinoma H22 cells were performed to develop insulin as a carder of radioiodine. Methods: 1 )Insulin was radiolabeled with iodine-125 or iodine-131 using a Chloramines T method. Twenty mice bearing tumor were divided into 4 groups (n = 5 each) randomly. They were killed at 5, 15, 30, 60 min after 125I-insulin administered intravenously. The percentage of injected dose of 125I-insulin per gram of tissue(%ID/gdis) in mice bearing tumor were determined. 2) Another ten mice bearing tumor were selected to be as a inhibition group. They received cold insulin 2 mg intravenously 2 min ahead of administration of 125I-insulin and they were killed at 30 min (n=5) and 60 rain (n=5) randomly post 125I-insulin injection. The %ID/ginh and the inhibited rates[(%ID/gdis-%iD/ginh) %ID/gdis 100%] were obtained. 3) One tumor-mouse received 7.4 Mbq 13II-insulin intravenously, another received cold insulin 2 mg injection before 13II-insulin injection. Whole body images were carded out and the radioactivity ratios of tumor/normal were accounted at 60 min. Results: 1) The radiochemical purities of 125I-insulin and 13II-insulin were 96.7%-98.9%. The tumors uptake of the 125I-insulin increased gradually, its peak (%ID/gdis) was 3.44% 0.42% at 30 min, when the normal tissues uptake decreased sharply post-injection. The radioactivity ratio of the tumor/blood and tumor/muscle reached to 1.44 and 3.62 respectively at 60 min. 2)The tumor-inhibition rate was 32.07% at 30 min and 37.42% at 60 min. 3) A high radioactivity accumulation in tumor region could be seen in the mouse at 60 min post 131I-insulin injection. The radioactivity ratio of the tumor/normal tissue was 2.13 and it declined to 1.37 after received insulin 2 mg intervention. Conclusions

Hydrogeological research at the site of the Asse salt mine

International Nuclear Information System (INIS)

Batsche, H.; Rauert, W.; Klarr, K.

1980-01-01

In connection with the storage of radioactive wastes in the abandoned Asse salt mine near Brunswick (Federal Republic of Germany), the hydrogeology of the ridge of hills of Asse has been investigated. In order to obtain as detailed information as possible on the hydrogeological conditions, a long-term investigation programme has been set up and many methods of investigation have been used. Hydrogeological boring operations resulted in important scientific findings regarding, for example, the course of the salt table and the main anhydrite which towers up above the salt table into the overlying collapsed rocks. Hydrochemical data showed the hydraulic effect of transverse faults. Isotopic hydrological measurements permitted distinction between the flow behaviour of the groundwater in different aquifers. The origin of the salt springs at the northwest end of the structure can be explained. Some additional pumping and labelling tests are expected to yield quantitative results concerning hydraulic interrelationships recognized to date. The very complex hydrogeological structure of the ridge of hills of Asse is the result of the multiple succession of permeable and impermeable layers on the flanks of the structure, and, furthermore, is possibly due to the fact that in some individual faults groundwater may seep through normally impermeable layers as well as via waterways at the salt table. (author)

Qualification of flow barriers in salt formations; SVV 2. Qualifizierung von Stroemungsbarrieren in Salzformationen

Energy Technology Data Exchange (ETDEWEB)

Herbert, Horst-Juergen; Hertes, Uwe; Meyer, Lothar; Hellwald, Karsten; Dittrich, Juergen

2011-09-15

The GRS report covers the technical concept of flow barriers in salt formations of self-healing salt backfilling (SVV) for the plugging of boreholes in underground radioactive waste repositories. Laboratory measurements in large dimensions and in-situ experiments were performed in the salt formations carnallitite (Asse mine) and tachhydrite (mine Teutschenthal) and showed the technical feasibility of the concept. The report includes the following chapters following the introduction: State-of-the-art of science and technology before the beginning of the project. Aims of the project. SVV plugging principle. Raw material and salt solutions. Laboratory measurements. Experiments concerning the efficiency assessment of SVV plugging elements. Results on SVV sealing properties - laboratory experiments. Results on SVV sealing properties - in-situ experiments. Results concerning the geomechanical properties. SVV-long-term behavior. Technical requirements to the practicability of SVV plugging elements.

Location-independent study concerning the construction, operation and closure of possible facilities for the final storage of radioactive waste in rock-salt formations in the Netherlands

International Nuclear Information System (INIS)

1986-04-01

Final storage of radioactive waste has been studied on the base of two main concepts: in deep boreholes and caverns from the mowing-field, and for a, for this purpose to be developed, underground ore. Storage supplies have been designed, including the closing constructions after finishing the storage activities, with a, much longer than usually, technical lifetime. Herein use has been made of in general known materials whose properties and behaviour were assumed to remain unaltered over long periods and also will not be influenced by the rock-salt environment. The possible storage concepts described are location independent and based upon the geological and geomechanical information which have been provided with the task and which are indicative for the rock-salt formations occurring in the Netherlands. In first instance the authors have started from the thermodynamical, chemical and fysical properties of the storage rock-formations as are mentioned in the apendices of the task. It particularly concerns properties of the storage rock-formations and the construction materials needed for a qualitatively good and reliable closing of the storage. The construction and operation of the in this report described storage concepts is based upon the storage scenario's as indicated in the task circumscriptions

Further design work on a repository in a salt dome

International Nuclear Information System (INIS)

Hamstra, J.; Janssen, L.G.L.

1985-01-01

The report presents the cost estimate and the work plan for the construction of a repository, to be mined in a medium-size salt dome for the simultaneous disposal of different categories of solid radioactive wastes. The repository is designed for all categories of waste from 40 years of operation of 25 nuclear power stations of 1000 MWe each, including the decommissioning waste from these stations as well as all the radioactive wastes from the hospitals and laboratories during a hundred-year period. The cost estimate includes preparation of a site, the construction, operation and abandonment of that repository. Moreover, an outline has been presented for a future updating and optimization study of the concept

Sorption-reagent treatment of brines produced by reverse osmosis unit for liquid radioactive waste management

International Nuclear Information System (INIS)

Avramenko, V. A.; Zheleznov, V. V.; Sergienko, V. I.; Chizhevsky, I. Yu

2003-01-01

The results of the pilot plant tests (2002-2003) of the sorption-reagent decontamination of high salinity radioactive waste (brines) remaining after the low-salinity liquid radioactive waste (LRW) treatment in the reverse-osmosis unit from long-lived radionuclides are presented. The sorption-reagent materials used in this work were developed in the Institute of Chemistry FEDRAS. They enable one to decontaminate brines with total salt content up to 50 g/l from long-lived radionuclides of Cs, Sr and Co. At joint application of the reverse-osmosis and sorption-reagent technologies total volume of solid radioactive waste (SRW) decreases up to 100-fold as compared to the technology of cementation of reverse osmosis brines. In this case total cost of LRW treatment and SRW disposal decreases more than 10-fold. Brines decontaminated from radionuclides are then diluted down to the ecologically safe total salts content in water to be disposed of. Tests were performed to compare the efficiency of technologies including evaporation of brines remaining after reverse osmosis process and their decontamination by means of the sorption-reagent method. It was shown that, as compared to evaporation, the sorption-reagent technology provides substantial advantages as in regard to radioactive waste total volume reduction as in view of total cost of the waste management

The catabolism of radioiodinated anti-lung-cancer monoclonal antibodies in tumor-bearing nude mice

International Nuclear Information System (INIS)

Shi Xubao

1991-01-01

Nude mice bearing humor lung cancer xenografts were injected intravenously or intraperitoneally with a mixture of radioiodinated anti-lung-cancer monoclonal antibodies, 2E3 and 6D1. The blood radioactivity versus time curve was fitted to a two-compartment open model with a 3.4 day blood radioactivity clearance half-life and a 636 ml/kg apparent distribution volume. Radioiodinated 2E3 and 6D1 given intraperitoneally were rapidly absorbed, with a 2.08 absorption half-life and 89% bioavailability. The highest radioactivity levels were found in the tumor, blood, liver and spleen 1-3 days after injection; next came the lung, kidney, stomach and intestine. The relative radioactivity increased in the tumor as levels in blood and normal tissues decreased. The in vivo deiodination of radioiodinated 2E3 and 6D1 was about 18.6% and free radioiodine was excreted in the urine

The land disposal of organic materials in radioactive wastes: international practice and regulation

International Nuclear Information System (INIS)

Hooper, A.J.

1988-01-01

World-wide practice and regulation with regard to organic materials in radioactive wastes for land disposal have been examined with a view to establishing, where possible, their scientific justification and their relevance to disposal of organic-bearing wastes in the UK. (author)

1974 conceptual design description of a bedded salt pilot plant in southeast New Mexico

International Nuclear Information System (INIS)

1977-06-01

The policy of the United States Atomic Energy Commission is to take custody of all commercial high-level radioactive wastes and maintain control of them in perpetuity. This policy (Title 10, Code of Federal Regulations, Part 50, Appendix F) requires that the high-level wastes from nuclear fuels reprocessing plants be solidified within five years after reprocessing and then shipped to a federal repository within ten years after reprocessing. Ultimate disposal sites and/or methods have not yet been selected and are not expected to be ready when waste deliveries begin about 1983. Therefore, the AEC plans to build an interim storage facility, called Retrievable Surface Storage Facility (RSSF), to store and isolate the waste from man and his environment until the suitability of the permanent repository is demonstrated and public acceptance has been established. Meantime, the AEC is proceeding with the study and development of an ultimate disposal method. Bedded salt is being considered for ultimate waste disposal, and work is in progress to develop a Bedded Salt Pilot Plant to demonstrate its acceptability. The pilot plant will permit in situ verification of laboratory work on the interaction of heat and radioactivity of the waste with the salt and surroundings. One concept of such a pilot facility is described

Annual report 1975

International Nuclear Information System (INIS)

Krause, H.

1976-12-01

In 1975 R+D-work of the Waste Management Research Department (ABRA) essentially covered the following fields: 1) Solidification of high-level fission product solutions and liquid α-wastes into borosilicate glasses. 2) Treatment of medium-level aqueous and organic waste solutions from reprocessing plants. 3) Incorporation of low and medium level waste concentrates into bitumen. 4) Wet combustion of solid plutonium-bearing wastes. 5) Decontamination of equipment. 6) Development of nuclear equipment for the transport and for final disposal of radioactive wastes in the Asse salt mine. 7) Investigations relating to the nuclear safety of final disposal of radioactive wastes in the Asse salt mine. 8) Actinide migration in soils. This report gives a survey of the main results obtained. (orig.) [de

The geological and material investigation programme

International Nuclear Information System (INIS)

Joshi, A.V.

1982-01-01

The radioactive waste disposal problem is an interdisciplinary problem. The geological formation cannot be considered on its own, but must also be considered in connection with the engineering design of the disposal facility. Engineering design including the encapsulation of the glass in a 15 cm thick steel cylinder and a minimum 40 year cooling time ensures low temperatures in the salt-steel interface. Even if large quantities of carnallite were found 3.5 m away from the sides of the borehole, the temperature at 2500 m depth after taking into account temperature increase from radioactive waste will not release crystal water from the carnallite. Anhydrite layers, which may be found in the neighbourhood of Erslev 2 and at the depths contemplated for radioactive waste disposal, will not be continous, but only in the form of blocks of limited lengths. They cannot therefore form a passage to a water bearing aquifer. The volume of salt necessary for waste disposal - including a 200 m safety barrier - is less than 2 km 3 . The Mors dome with a salt volume of about 264 km 3 provides a very substantial safety margin. The geological investigations have fulfilled the purpose of the present phase of investigations and show the Mors salt dome to be a suitable dome for disposal of high-level radioactive waste. (EG)

Static fuel molten salt reactors - simpler, cheaper and safer

International Nuclear Information System (INIS)

Scott, Ian

2015-01-01

The many conceptual designs for Molten Salt Reactors (MSR's) today are all evolutions from the prototype MSR that went critical at Oak Ridge 50 years ago. Critically, they are based on pumping the molten fuel salt from a reaction chamber where the fuel achieves critical mass through a heat exchanger where the resulting heat is transferred to another working fluid. This basic concept was not the first idea that the Oak Ridge scientists considered. Their initial preference was to put the molten salt fuel into tubes, just like solid fuel pellets in their cladding, and circulate a coolant past the tubes. They concluded however that the low thermal conductivity of the salt meant that the tubes could be no wider than 2mm which would be entirely impractical. In this analysis they ignored the contribution of convection to heat transfer in fluids, probably because they were designing an aircraft engine where varying g forces would make convection unreliable. Moltex Energy has re-examined this decision using the modern tools of computational fluid dynamics to simulate convective flow in the molten salt and discovered that in fact tubes of similar diameter to those used for solid fuels are entirely practical. Power densities of 250kW/litre of fuel salt are readily attainable providing a higher overall power density than a PWR reactor. This discovery permits MSR's to be built without any of the complex pumping, passively safe drain systems, on line degassing, filtration and chemical processing needed in pumped MSR's. Their design is very simple and they have many intrinsic safety factors including low pressure operation, chemically unreactive fluids and strongly negative fuel thermal and coolant voiding reactivity coefficients. Most importantly, the highly radioactive fission products are retained in non-volatile form within the fuel tubes in the reactor core. Radioactive fuel salt never leaves the reactor vessel except in an immobile frozen form during

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.)

Treatment of low radioactive liquid waste by electrodialysis. Principles and experimental model

International Nuclear Information System (INIS)

Dogaru, D.

1998-01-01

Electrodialysis is a membrane separation process achieved by the use of differential driving force due to an electric potential across the membrane. It can be considered as a process in which salts are transferred under the impetus of an electrical potential from one solution to another, usually from a dilute to a concentrated solution, through a membrane barrier. In water, salts dissolve producing positively charged cations and negatively charged anions. If an electrical field is placed across a solution of salt by inserting a pair of electrodes into the solution, the cations migrate toward the negatively charged cathode, while anions migrate toward the positively charged anode. This contribution presents principles and experimental model for removed radionuclides from low radioactive liquid wastes. A typical electrodialysis cell arrangement consists of a series of anionic- and cationic-exchange membranes arranged in an alternating pattern between an anode and a cathode, to form an individual cell. The laboratory experimental apparatus consisted of an electrodialysis unit, two recirculating pumps, a voltage stabilizer, connecting pipes and recirculating tanks. The unit had 10 cell pairs. The cell geometry was a flat-plate and frame configuration with anode, cathode, charge selective membranes, gaskets and spacers. The anode material was nickel and the cathode material was TiO 2 , with an electrode area of 90 cm 2 . For Radioactive Waste Treatment Plant, the ability to separate equivalent ions is very attractive and opens the possibility of applying electrodialysis to a wide variety of systems with appropriate choice of operating conditions and ion-selective membranes. The technique creates minimal secondary waste. However, before electrodialysis can be implemented, a chemical pre-treatment for radioactive wastes is necessary. (author)

Bearing system

Science.gov (United States)

Kapich, Davorin D.

1987-01-01

A bearing system includes backup bearings for supporting a rotating shaft upon failure of primary bearings. In the preferred embodiment, the backup bearings are rolling element bearings having their rolling elements disposed out of contact with their associated respective inner races during normal functioning of the primary bearings. Displacement detection sensors are provided for detecting displacement of the shaft upon failure of the primary bearings. Upon detection of the failure of the primary bearings, the rolling elements and inner races of the backup bearings are brought into mutual contact by axial displacement of the shaft.

Study of solid phase kinetics during cyanidation using the 198 Au radioactive tracer

International Nuclear Information System (INIS)

Barbus, A.; Pop, I.I.; Gaspar, E.

1995-01-01

During cyanidation, the various gold bearing pyrite sorts exhibit different behaviour, that sometimes cause increased cyanidation times influencing the reagent and power consumption, in the same time generating fluctuations in the recovery efficiencies. The introduction of the 198 Au radioactive tracer into the cyanidation circuit enabled us to follow several parameters of the cyanidation kinetics: the average residence time of the gold bearing pyrite in the technological equipment, information about the homogenization process, dispersion of solids and gold dissolution efficiency on each technological stage. (author)

Container materials for isolation of radioactive waste in salt

International Nuclear Information System (INIS)

Streicher, M.A.; Andrews, A.

1987-10-01

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

Formation of filtration fields close to near-surface radioactive waste storages

International Nuclear Information System (INIS)

Mart'yanov, V.V.

2008-01-01

Data on the formation of filtration fields in the location of near-surface radioactive waste storages for the conditions of uniformly isotropic properties of bearing strata are demonstrated. The possibility for changing parameters of mean-caused ground flow depending on water permeability of the storages and their dimensions in plan is noted. Comparison of different filtration fields permits to determine a state of its isolating properties. Assessment criteria of the storage engineering barriers integrity are given. Conditions for uniformly isotropic properties of bearing strata by three scenarios, when engineering barriers of the storage are waterproof, distracted or lost protective properties in full, have been determined. Changing filtration field, geochemical and radiochemical situations in bearing strata are noted to represent one of basic characteristics of the integrity of the storage [ru

Experiments in a 600m borehole in the Asse II salt mine

International Nuclear Information System (INIS)

Heijdra, J.J.

1992-07-01

In the design and fabrication of underground disposal sites for radio-active waste in salt formations and the assessment of the safety of such disposal facilities, the thermo-mechanical behaviour of rock salt plays an important role. In previous research programmes models have been developed which need to be verified by in-situ experiments. It has been proven during the COSA project that computations based on laboratory scale experiments do not agree with in-situ measurements. Based on the experiments performed already and on the associated validation work, two items were considered to be of special concern, viz. the consecutive behaviour of rock salt and the rock pressure in the Asse salt mine. A particular problem in the constitutive relations is the elastic or apparent elastic behaviour of rock salt. It appeared that the salt around openings is weaker than could be expected on the basis of laboratory experiments. Possible explanations are primary creep and the weakening effect of micro cracks. In the research programme discussed here, in-situ experiments will be carried out in the Asse II salt mine in the Federal Republic of Germany. The measurements will be carried out in dry drilled boreholes. The development of the drilling technique was part of a related programme carried out under supervision of GSF-Forschungszentrum fuer Umwelt und Gesundheit (Research Centre for Environment and Health). (author). 3 refs

Removal of uranium from spent salt from the moltensalt oxidation process

International Nuclear Information System (INIS)

Summers, L.; Hsu, P.C.; Holtz, E.V.; Hipple, D.; Wang, F.; Adamson, M.

1997-03-01

Molten salt oxidation (MSO) is a thermal process that has the capability of destroying organic constituents of mixed wastes, hazardous wastes, and energetic materials. In this process, combustible waste and air are introduced into the molten sodium carbonate salt. The organic constituents of the waste materials are oxidized to carbon dioxide and water, while most of the inorganic constituents, including toxic metals, minerals, and radioisotopes, are retained in the molten salt bath. As these impurities accumulate in the salt, the process efficiency drops and the salt must be replaced. An efficient process is needed to separate these toxic metals, minerals, and radioisotopes from the spent carbonate to avoid generating a large volume of secondary waste. Toxic metals such as cadmium, chromium, lead, and zinc etc. are removed by a method described elsewhere. This paper describes a separation strategy developed for radioisotope removal from the mixed spent salt, as well as experimental results, as part of the spent salt cleanup. As the MSO system operates, inorganic products resulting from the reaction of halides, sulfides, phosphates, metals and radionuclides with carbonate accumulate in the salt bath. These must be removed to prevent complete conversion of the sodium carbonate, which would result in eventual losses of destruction efficiency and acid scrubbing capability. There are two operational modes for salt removal: (1) during reactor operation a slip-stream of molten salt is continuously withdrawn with continuous replacement by carbonate, or (2) the spent salt melt is discharged completely and the reactor then refilled with carbonate in batch mode. Because many of the metals and/or radionuclides captured in the salt are hazardous and/or radioactive, spent salt removed from the reactor would create a large secondary waste stream without further treatment. A spent salt clean up/recovery system is necessary to segregate these materials and minimize the amount of

Offsite testing in support of the Salt Repository Project

International Nuclear Information System (INIS)

Kalia, H.N.

1987-04-01

This report presents a rationale and recommendation to perform an offsite testing program in support of the Salt Repository Project. The investigation to be performed primarily consists of qualifying test methods and procedures, qualifying personnel-training procedures, evaluating test instruments and selected equipment, and obtaining mining and production equipment performance-related information. The key objective of these activities is to develop capabilities to be used at the exploratory shaft facility (ESF). The ESF is to be excavated at the Deaf Smith County site to characterize the salt site for the construction of a repository used to isolate radioactive waste from the biosphere. The bulk of the offsite testing work will be performed at Avery Island Salt Mine at New Iberia, Lousiana. Additional knowledge will be obtained by exchanging technical information either as participants or as observers at the Waste Isolation Pilot Plant (WIPP) site and the Asse Mine in the Federal Republic of Germany (FRG). It is estimated that the offsite testing program will cost approximately $9.3 million over 4 fiscal years. 14 refs., 1 fig., 8 tabs

Legal aspects of sub-seabed disposal of radioactive waste

International Nuclear Information System (INIS)

Reyners, P.

1981-10-01

In connection with methods for disposal of highly radioactive waste, that consisting of burying such waste in the sub-seabed arouses an increasingly marked interest among specialists. Apart from the technical difficulties still to be overcome and current safety assessments, this method gives rise to quite considerable legal and political problems. Their solution will undoubtedly have a bearing on its chances of being implemented. (NEA) [fr

Slurry explosives containing the combination of nitrogen-base salt and hard solid particles as sensitizer

Energy Technology Data Exchange (ETDEWEB)

Lyerly, W.M.

1971-11-02

In recent years, blasting agents, particularly those of the type known as water gels or slurry explosives have gained considerable commercial acceptance. Generally, the slurry explosives are comprised of an inorganic oxidizing salt, predominantly ammonium nitrate, a thickening agent for the liquid, water, and fuel. The density, velocity of detonation, and ability to sustain detonation are increased so that the compositions propagate in small diameter boreholes. A water-bearing slurry explosive is described containing inorganic oxidizing salt, fuel, water and thickener together with nitrogen- base salt and solid particles having a hardness of at least 4 on the Mohs scale and that have an acoustic impedance at least 2 times that of the matrix of the slurry explosive. (15 claims)

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 thermo-mechanical behaviour of a salt dome with a heat-generating waste repository

International Nuclear Information System (INIS)

Janssen, L.G.J.; Prij, J.; Kevenaar, J.W.A.M.; Jong, C.J.T.; Klok, J.; Beemsterboer, C.

1984-01-01

This report reviews the analytical work on the disposal of radioactive waste in salt domes performed at ECN in the period 1 January 1980 to 31 December 1982. Chapter 4 in the main report covers the global temperature and deformation analyses of the salt dome and the surrounding rocks. The attached three topical reports cover self-contained parts of the study. The computer program TASTE developed to analyse, at acceptable cost and with, for engineering purposes, sufficient accuracies, the temperature rises in the salt dome due to the stored heat-generating waste is described in Annex 1. Annex 2 gives a description of the extended finite element program GOLIA. The program has been extended to make it suitable for the creep analysis of salt domes with repositories of heat-generating waste. The study on the closing and sealing of boreholes wit heat-generating waste is reported in Annex 3

Safety Assessment of Radioactive waste Repositories

International Nuclear Information System (INIS)

1991-01-01

It is planned to dispose of high-level radioactive wastes in deep geological formations. To access the long-term safety of radioactive waste disposal systems, mathematical models are used to describe groundwater flow, chemistry and potential radionuclide migration through these formations. Establishing the validity of such models is important in order to obtain the necessary confidence in the safety of the disposal method. The papers in these proceedings of the GEOVAL'90 Symposium describe the current state of knowledge on the validation of geosphere flow and transport models. This symposium, divided into five sessions, contains 65 technical papers: session 1 - Necessity of validation. Session 2 - Progress in validation of flow and transport models in orystalline rock, unsaturated media, salt media or clay. Session 3 - Progress in validation of geochemical models. Session 4 - Progress in validation of coupled thermo-hydro-mechanical effects. Session 5 - Validation strategy

Proceedings of the 7th US/German Workshop on Salt Repository Research, Design, and Operation.

Energy Technology Data Exchange (ETDEWEB)

Hansen, Francis D. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Steininger, Walter [Karisruhe Inst. of Technology (Germany); Bollingerfehr, Willhelm [DBE TECHNOLOGY GmbH (Germany)

2017-01-01

The 7th US/German Workshop on Salt Repository Research, Design, and Operation was held in Washington, DC on September 7-9, 2016. Over fifty participants representing governmental agencies, internationally recognized salt research groups, universities, and private companies helped advance the technical basis for salt disposal of radioactive waste. Representatives from several United States federal agencies were able to attend, including the Department of Energy´s Office of Environmental Management and Office of Nuclear Energy, the Environmental Protection Agency, the Nuclear Regulatory Commission, and the Nuclear Waste Technical Review Board. A similar representation from the German ministries showcased the covenant established in a Memorandum of Understanding executed between the United States and Germany in 2011. The US/German workshops´ results and activities also contribute significantly to the Nuclear Energy Agency Salt Club repository research agenda.

Policy of radioactive waste disposal in the Netherlands

International Nuclear Information System (INIS)

Selling, H.A.

2002-01-01

Earlier this year the final report of the CORA Commission on retrievable disposal of radioactive waste was published. It confirmed the technical feasibility of retrievable repository concepts in the deep underground. Rock salt and sedimentary clay were considered as potential host rocks for such a repository. It is recommended, among other things, that subsequent research programmes should focus on stakeholder identification and involvement in a stepwise decision-making process of waste disposal. (author)

The research on biodistribution of bearing sarcoma mice and rabbit SPECT imaging of 177Lu-DOTMP

International Nuclear Information System (INIS)

Deng Xinrong; Xiang Xueqin; Li Fenglin; Fan Caiyun; Liu Zihua; Luo Zhifu; Chen Yang

2012-01-01

Cyclen (1, 4, 7, 10-tetraazacyclododecane) and H 3 PO 3 were used to synthesis DOTMP (1, 4, 7, 10-tetraazacyclododecane-1, 4, 7, 10-Tetraaminomethylenephosphonate), and then DOTMP was labelled with 177 Lu. The research of biodistribution of 177 Lu-DOTMP in model mice bearing S180 sarcoma and SPECT imaging in Japanese white rabbit were also carried out. The results of biodistribution of bearing S180 mice indicated that 177 Lu-DOTMP cleared rapidly from blood and was selectively delivered to target bone. The radioactivity uptake was mainly in bone and less in other organs and tissues. The results of SPECT imaging of Japanese white rabbit showed that the radioactivity was accumulated in bladder. 177 Lu-DOTMP was mainly excreted by kidney. The uptake of the activity in the skeleton was observed significantly within 22 h post-injection and it became quite significant at 46 h post-injection. It indicated that 177 Lu-DOTMP has good bone targeting and is worthy of further study. (authors)

The backfilling and sealing of radioactive waste repositories. V. 2. Figure - Tables - Appendices

International Nuclear Information System (INIS)

1984-01-01

The two volumes of this report present a review study about backfilling and sealing of radioactive waste repositories in granites, argillaceous and salt formations. Volume 2 contains all the figures, table and appendices A detailed account of candidate backfill materials is given in a standardized format

Large-scale demonstration of disposal of decontaminated salt as saltstone. Part I. Construction, loading, and capping of lysimeters

International Nuclear Information System (INIS)

Wolf, H.C.

1984-06-01

The installation phase of a large-scale demonstration of the disposal concept for decontaminated, low-level radioactive salt waste at the Savannah River Plant was completed in December 1983 and January 1984. The installation entailed immobilizing 7500 gallons of decontaminated salt solution with a blended cement formulation and pouring the resulting grout, saltstone, into three specially designed lysimeters for extended in-field leaching tests under natural conditions. 4 references, 35 figures, 4 tables

Geologic characterization report for the Paradox Basin Study Region, Utah Study Areas. Volume 6: Salt Valley

Science.gov (United States)

1984-12-01

Surface landforms in the Salt Valley Area are generally a function of the Salt Valley anticline and are characterized by parallel and subparallel cuestaform ridges and hogbacks and flat valley floors. The most prominent structure in the Area is the Salt Valley anticline. Erosion resulting from the Tertiary uplift of the Colorado Plateau led to salt dissolution and subsequent collapse along the crest of the anticline. Continued erosion removed the collapse material, forming an axial valley along the crest of the anticline. Paleozoic rocks beneath the salt bearing Paradox Formation consist of limestone, dolomite, sandstone, siltstone and shale. The salt beds of the Paradox formation occur in distinct cycles separated by an interbed sequence of anhydrite, carbonate, and clastic rocks. The Paradox Formation is overlain by Pennsylvanian limestone; Permian sandstone; and Mesozoic sandstone, mudstone, conglomerate and shale. No earthquakes have been reported in the area during the period of the historic record and contemporary seismicity appears to be diffusely distributed, of low level and small magnitude. The upper unit includes the Permian strata and upper Honaker trail formation.

Volume reduction of waste contaminated by fission product elements and plutonium using molten salt combustion

International Nuclear Information System (INIS)

McKenzie, D.E.; Grantham, L.F.; Paulson, R.B.

1979-01-01

In the Molten Salt Combustion Process, transuranic or β-γ organic waste and air are continuously introduced beneath the surface of a sodium carbonate-containing melt at a temperature of about 800 0 C. Complete combustion of the organic material to carbon dioxide and steam occurs without the conversion of nitrogen to nitrogen oxides. The noxious gases formed by combustion of the chloride, sulfur or phosphorus content of the waste instantly react with the melt to form the corresponding sodium compounds. These compounds as well as the ash and radionuclides are retained in the molten salt. The spent salt is either fused cast into an engineered disposal container or processed to recover salt and plutonium. Molten salt combustion reduces the waste to about 2% of its original volume. Many reactor or reprocessing wastes which cannot be incinerated without difficulty are readily combusted in the molten salt. A 50 kg/hr molten salt combustion system is being designed for the Radioactive Waste Management Complex at the Idaho National Engineering Laboratory. Construction of the combustor started during 1977, and combustor startup was scheduled for the spring of 1978

A natural analogue for near-field behaviour in a high level radioactive waste repository in salt: the Salton Sea geothermal field, California, USA

International Nuclear Information System (INIS)

Elders, W.A.

1987-01-01

In the Salton Sea Geothermal Field (SSGF), in the sediments of the delta of the Colorado River, we are developing a three-dimensional picture of active water/rock reactions at temperatures of 0 C and salinities of 7 to 25 weight percent to produce quantitative data on mineral stabilities and mobilities of naturally-occurring radio-nuclides. The aim is to produce data to validate geochemical computer codes being developed to assess the performance of a Commercial High-Level Waste (CHLW) repository in salt. Among the findings to date are: (1) greenschist facies metamorphism is occurring; (2) brine compositions are fairly similar to those expected in candidate salt repository sites; (3) U and Th concentrations in the rocks are typical for sedimentary rocks; (4) the brines are enriched in Na, Mn, Zn, Sr, Ra Po and strongly depleted in U and Th relative to the rocks; (5) significant radioactive disequilibria exist in brines and solid phases of the SSGF. The disequilibria in the actinide series allow estimation of the rates of brine-rock interaction and understanding of hydrologic processes and radionuclide behaviour. Work is continuing emphasizing the reactions of authigenic clay minerals, epidotes, feldspars, chlorites and sulphates. So far, adapting geochemical codes to the necessary combination of high salinity and high temperature has lagged behind the natural analogue study of the SSGF so that validation is still in progress. In the future our data can be also used in validating performance assessment codes which couple geochemistry and transport processes, and in design of waste packages and back fill compositions. (author)

Remaining porosity and permeability of compacted crushed rock salt backfill in a HLW repository. Final report

Energy Technology Data Exchange (ETDEWEB)

Jobmann, M.; Mueller, C.; Schirmer, S.

2015-11-15

The safe containment of radioactive waste is to be ensured by the geotechnical barriers in combination with the containment-providing rock zone (CRZ). The latter is a key element of the recently developed concept of demonstrating the integrity of the geologic barrier (Krone et al., 2013). As stipulated in the safety requirements of the regulating body the CRZ has to have strong barrier properties, and evidence needs to be provided that it retains its integrity throughout the reference period (BMU, 2010). The underground openings excavated in the rock salt will close over time due to the creep properties of the rock salt. This process causes deformations in the surrounding rock salt, which leads to a change in stress state in the virgin rock and may impair the integrity of the containment-providing rock zone. In order to limit the effects of these processes, all underground openings will be backfilled with crushed salt. Immediately after backfilling, the crushed salt will have an initial porosity of approx. 35%, which - over time - will be reduced to very low values due to the creep properties of the rock salt. The supporting pressure that builds up in the crushed salt with increasing compaction slows down the creeping of the salt. Major influencing factors are the temperature (with higher temperatures accelerating the salt creeping) and the moisture of the salt, which - due to the related decrease in the resistance of the crushed salt - facilitates its compaction. The phenomenology of these processes and dependencies is understood to a wide extent. This project investigated the duration until compaction is completed and when and under what circumstances the crushed salt will have the sealing properties necessary to ensure safe containment. Thermo-hydro-mechanical (THM) processes play a crucial role in determining whether solutions which might enter the mine could reach the radioactive waste. This includes changes in material behaviour due to a partial or complete

Comparison of the rift and post-rift architecture of conjugated salt and salt-free basins offshore Brazil and Angola/Namibia, South Atlantic

Science.gov (United States)

Strozyk, Frank; Back, Stefan; Kukla, Peter A.

2017-10-01

This study presents a regional comparison between selected 2D seismic transects from large, conjugated salt and salt-free basins offshore southern Brazil (Campos Basin, Santos Basin, Pelotas Basin) and southwest Africa (Kwanza Basin, northern and southern Namibe Basin, Walvis Basin). Tectonic-stratigraphic interpretation of the main rift and post-rift units, free-air gravity data and flexural isostatic backstripping were used for a comprehensive basin-to-basin documentation of key mechanisms controlling the present-day differences in conjugated and neighbouring South Atlantic basins. A significant variation in the tectonic-sedimentary architecture along-strike at each margin and between the conjugated basins across the South Atlantic reflects major differences in (1) the structural configuration of each margin segment at transitional phase between rifting and breakup, as emphasized in the highly asymmetric settings of the large Santos salt basin and the conjugated, salt-free southern Namibe Basin, (2) the post-breakup subsidence and uplift history of the respective margin segment, which caused major differences for example between the Campos and Espirito Santo basins and the conjugated northern Namibe and Kwanza basins, (3) variations in the quantity and distribution of post-breakup margin sediments, which led to major differences in the subsidence history and the related present-day basin architecture, for example in the initially rather symmetric, siliciclastic Pelotas and Walvis basins, and (4) the deposition of Aptian evaporites in the large rift and sag basin provinces north of the Rio Grande Rise and Walvis Ridge, highly contrasting the siliciclastic basins along the margin segments south of the ridges. The resulting present-day architecture of the basins can be generally classified as (i) moderately symmetric, salt-free, and magma-rich in the northern part of the southern segment, (i) highly asymmetric, salt-bearing and magma-poor vs. salt-free and magma

Boron removal in radioactive liquid waste by forward osmosis membrane

Energy Technology Data Exchange (ETDEWEB)

Doo Seong Hwang; Hei Min Choi; Kune Woo Lee; Jei Kwon Moon [KAERI, Daejeon (Korea, Republic of)

2013-07-01

This study investigated the treatment of boric acid contained in liquid radioactive waste using a forward osmosis membrane. The boron permeation through the membrane depends on the type of membrane, membrane orientation, pH of the feed solution, salt and boron concentration in the feed solution, and osmotic pressure of the draw solution. The boron flux begins to decline from pH 7 and increases with an increase of the osmotic driving force. The boron flux decreases slightly with the salt concentration, but is not heavily influenced by a low salt concentration. The boron flux increases linearly with the concentration of boron. No element except for boron was permeated through the FO membrane in the multi-component system. The maximum boron flux is obtained in an active layer facing a draw solution orientation of the CTA-ES membrane under conditions of less than pH 7 and high osmotic pressure. (authors)

Geologic feasibility of selected chalk-bearing sequences within the conterminous United States with regard to siting of radioactive-waste repositories

International Nuclear Information System (INIS)

Gonzales, S.

1975-11-01

Various geologic and hydrologic parameters are evaluated in relation to assessing the potential for repository storage of high-level radioactive wastes within several stratigraphic sequences dominated by chalks and chalky limestones. The former lithology is defined as a carbonate rock consisting mainly of very fine-grained particles of micritic calcite. Although chalks also contain coarser-grained particles such as shells of fossil foraminifera and non-calcitic minerals like quartz, most contain more than 90 percent micritic material. The latter represents broken fossil coccolith plates. The chalk-dominated formations discussed are exposed and underlie two different physiographic provinces which nevertheless display a general similarity in both being regions of extensive plains. The Niobrara Formation occurs mainly within the Great Plains province, while the Austin Chalk of Texas and the Selma Group of Alabama and Mississippi are located in the western and eastern Gulf Coastal Plain, respectively. The preliminary assessment is that chalk-bearing sequences show some promise and are deserving of added consideration and evaluation. Containment for hundreds of thousands of years would seem possible given certain assumptions. The most promising units from the three studied are the Niobrara Formation and Selma Group. Regional and local conditions make the Austin more suspect

Salts of the iodine oxyacids in the impregnation of adsorbent charcoal for trapping radioactive methyliodide

International Nuclear Information System (INIS)

Deitz, V.R.; Blachly, C.H.

1977-01-01

Radioactive iodine and radioactive methyliodide can be more than 99.7 percent removed from the air stream of a nuclear reactor by passing the air stream through a 2-inch thick filter which is made up of impregnated charcoal prepared by contacting the charcoal with a solution containing KOH, iodine or an iodide, and an oxyacid, followed by contacting with a solution containing a tertiary amine. 3 claims

Research on rejection performance of reverse osmosis to nickel in simulated radioactive wastewater

International Nuclear Information System (INIS)

Kong Jinsong; Wang Xiaowei

2013-01-01

In order to reveal the rejection performance of the reverse osmosis applied in the radioactive wastewater treatment, treatment experiments were carried out on a pilot reverse osmosis equipment using wastewater containing nickel nuclide. Results showed that the rejection ratio of reverse osmosis to nickel was almost not affected by the operation pressure and the ratio of reclaiming, and had no direct relation with the salt rejection ratio. The ratio of nickel rejection reached 95% in the experiment condition and could meet the requirement on the disposal of radioactive wastewater produced by nuclear powered installations. (authors)

Research on rejection performance of reverse osmosis to cobalt in simulated radioactive wastewater

International Nuclear Information System (INIS)

Kong Jinsong; Tian Yanjie

2012-01-01

In order to reveal the rejection performance of the reverse osmosis applied in the radioactive wastewater treatment, treatment experiments were carried out on a pilot reverse osmosis equipment using wastewater containing cobalt nuclide. Results showed that the rejection ratio of reverse osmosis to cobalt was almost not affected by the operation pressure and the ratio of reclaiming, and had no direct relation with the salt rejection ratio. The ratio of cobalt rejection reached 90% in the experiment condition and could meet the requirement on the disposal of radioactive wastewater produced by nuclear powered installations. (authors)

Research on rejection performance of reverse osmosis to manganese in simulated radioactive wastewater

International Nuclear Information System (INIS)

Kong Jinsong; Wang Xiaowei

2012-01-01

In order to reveal the performance of reverse osmosis applied in the radioactive wastewater treatment, treatment experiments are carried out on a pilot RO equipment using wastewater containing manganese nuclide. Results show that the rejection ratio of RO to manganese is almost not influenced by the operation pressure and the ration of reclaiming, and has no direct relation with the salt rejection ratio. The ratio of manganese rejection is more than 95% and can meet the requirement on the disposal of radioactive wastewater produced by pressurized water reactors. (authors)

Safe disposal of radionuclides in low-level radioactive-waste repository sites; Low-level radioactive-waste disposal workshop, U.S. Geological Survey, July 11-16, 1987, Big Bear Lake, Calif., Proceedings

Science.gov (United States)

Bedinger, Marion S.; Stevens, Peter R.

1990-01-01

In the United States, low-level radioactive waste is disposed by shallow-land burial. Low-level radioactive waste generated by non-Federal facilities has been buried at six commercially operated sites; low-level radioactive waste generated by Federal facilities has been buried at eight major and several minor Federally operated sites (fig. 1). Generally, low-level radioactive waste is somewhat imprecisely defined as waste that does not fit the definition of high-level radioactive waste and does not exceed 100 nCi/g in the concentration of transuranic elements. Most low-level radioactive waste generated by non-Federal facilities is generated at nuclear powerplants; the remainder is generated primarily at research laboratories, hospitals, industrial facilities, and universities. On the basis of half lives and concentrations of radionuclides in low-level radioactive waste, the hazard associated with burial of such waste generally lasts for about 500 years. Studies made at several of the commercially and Federally operated low-level radioactive-waste repository sites indicate that some of these sites have not provided containment of waste nor the expected protection of the environment.

National environmental radioactivity networks-1993; Reti nazionali si sorveglianza della radioattivita` ambientale in Italia-1993

Energy Technology Data Exchange (ETDEWEB)

Belli, M; Notaro, M.; Rosamilia, S.; Sansone, U; Tommasi, R.

1998-12-31

This report contains the environmental radioactivity data collected in Italy during 1993, by the National Environmental Radioactivity Networks. The data contained in this report have been provided by the institutions participating in the National Environmental Radioactivity Networks. The National Environmental Protection Agency (ANPA) is law-fully responsible for publishing the report. The results of the measurements of radioactivity, are generally reported by only one significant figure. An arithmetical average of a series of figures, some of which are preceded by the sign `less than` (<), is given with this sign only when the figures bearing < affect remarkably (more then 50%) the value resulting from the average. Reproduction of the data contained in this report is authorized, provided the source is acknowledged.

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

International Nuclear Information System (INIS)

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

2011-12-01

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

Salts of the iodine oxyacids in the impregnation of adsorbent charcoal for trapping radioactive methyliodide

International Nuclear Information System (INIS)

1980-01-01

A method of removing methyliodide 131 gas from the effluent of a reactor, comprises passing the effluent gas through a charcoal sorbent formed by first contacting charcoal with a liquid containing a hypoiodite obtained when an aqueous mixture of a first component comprising a salt of an iodine oxyacid selected from periodate, iodate and hypoiodite and a second component selected from iodine and/or an iodide salt is adjusted to a pH of about 10 by the addition of an inorganic base, and then contacting the resulting impregnated charcoal with a tertiary amine. (author)

Policies on radioactive waste disposal in the Netherlands

International Nuclear Information System (INIS)

Selling, H.A.

1999-01-01

An outline is given of the policy in the Netherlands on radioactive waste management, with an emphasis on the preferred disposal strategies. A description is given of the siting and licensing process for the waste treatment and storage facility of COVRA, which is in many respects expected to be comparable with that for a disposal site in due course. Immediate disposal of radioactive waste is not envisaged. Instead, the government has opted for long term interim storage in an engineered facility until sufficient confidence has been obtained on the safety performance of a geological repository over long time periods. In the previous decade research has mostly focused on the exploration of the suitability of existing salt formations in the northern part of the country as host rock for a radioactive waste repository. Although so far no in situ research was carried out, it could be demonstrated by utilising values of the relevant parameters from other rock salt formations that, in principle, deep underground disposal of radioactive waste is safe. This assessment was made by comparing both with common radiation protection criteria and with risk criteria over long periods of time. However, a decision to proceed with in situ research was postponed in view of the strong opposition from the local population against underground disposal. Instead, the scope of the research was extended to other host rock materials (clay). Additionally, from a sustainability point of view it was demanded that disposal should be conceived as an irreversible process. This means that the waste should be disposed of in such a way that it is retrievable in case better processing methods for the waste would become available. This demand of retrievability derives from the general waste policy to close the life-cycles of raw materials in order not to deprive future generations from their benefits. Consequently, much of the sequential research is now focused on the safety and financial impact of

Disposal of Radioactive Wastes in Natural Salt; Elimination des Dechets Radioactifs dans le Sel Naturel; 0423 0414 ; Evacuacion de Desechos Radiactivos en Formaciones Salinas Naturales

Energy Technology Data Exchange (ETDEWEB)

Parker, F. L.; Boegly, W. J.; Bradshaw, R. L.; Empson, F. M.; Hemphill, L.; Struxness, E. G.; Tamura, T. [Health Physics Division, Oak Ridge National Laboratory, Oak Ridge, TN (United States)

1960-07-01

The proposed use of cavities in salt formations as a disposal site for radioactive wastes is based upon : 1. Existence of salt for geologic time periods, 2. The impermeability of salt to the passage of water; 3. The widespread geographical distribution of salt; 4. The extremely large quantities of salt available; 5. The structural strength of salt; 6. The relatively high thermal conductivity of salt in comparison with other general geologic formations; 7. The possible recovery of valuable fission products in the wastes injected into the salt; 8. The relative ease of forming cavities in salt by mining, and the even greater ease and low cost of developing solution cavities in salt; and 9. The low seismicity in the areas of major salt deposits. Radioactive liquid wastes can be stored in cavities in natural salt formations if the structural properties of the salt are not adversely affected by chemical interaction, pressure, temperature, and radiation. Analytical studies show that it is possible to-store 2-year-old 10,000 MWD/T, 800 gal/ton waste in a sphere of 10 ft diameter without exceeding a temperature of 200 Degree-Sign F. Laboratory tests show that the structural properties and thermal conductivity of rock salt are not greatly altered by high radiation doses, although high temperatures increase the creep rate for both irradiated and unirradiated samples. Chemical interaction of liquid wastes with salt produces chlorine and other chlorine compound gases, but the volumes are not excessive. The migration of nuclides through the salt and deformation of the cavity and chamber can only be studied in undisturbed salt in situ. One-fifth-scale models have been run in a bedded salt deposit in Hutchinson, Kansas, and full-scale field tests are in progress. (author) [French] L'emploi envisage des cavites des gisements de sel comme lieu d'evacuation des dechets radioactifs se-fonde sur les considerations suivantes: 1. L'existence du sel dans des formations correspondant a

Radioresistance of aboriginal earthworms of Absheron peninsula in radioactive model systems

International Nuclear Information System (INIS)

Suleymanova, A.S.; Garibov, A.A.; Abdullayev, A.A.; Naghiyev, J.A.; Farajov, M.F.; Samedov, P.A.

2011-01-01

Full tex:Soil animals are the most suitable biological indicators of radioactive pollution because they are parts of nutritional chains and webs, occur in relatively high numbers and can be collected during most parts of the year.The use of earthworms in the soils from Ramana iodine plant area of Absheron peninsula, which is rich with radionuclide and in the soil mixed with RaCl2 and UO2SO4 salt solutions in different concentrations, for resistance to ionizing radiation in soil is reviewed.Effect of radionuclides on vital functions of earthworms and determination of radionuclides (before and after experiments) in contaminated soils by ?-spectrometer were carried out in laboratory condition during a month. Regarding to ?-spectrometric results there were determined that earthworms had absorbed most of radioactive elements and allocated them as coprogenous substances on the upper layer of soil. In Ramana soils mostly the 238U radionuclides were highly accumulated in gut cells of the earthworms. By the influence of radioactive elements it was shown that the earthworms from Ramana iodine plant territory variants had proved particularly sensitive to an increased Ra-radiation background and to iodine factor.It was interestingly established the proportional dependence between rising level of accumulation in earthworms' body and in their coprolites and increasing of radioactive salts containing in the soils treated by UO2SO4 and RaCl2 solutions.Thus there is different level of radioresistance for earthworms and they are among the best bioindicators of polluted soils. There was an obvious perspective of using of earthworms as bioremediators in polluted soil with radionuclides in future as well.

Journal bearing

Science.gov (United States)

Menke, John R.; Boeker, Gilbert F.

1976-05-11

1. An improved journal bearing comprising in combination a non-rotatable cylindrical bearing member having a first bearing surface, a rotatable cylindrical bearing member having a confronting second bearing surface having a plurality of bearing elements, a source of lubricant adjacent said bearing elements for supplying lubricant thereto, each bearing element consisting of a pair of elongated relatively shallowly depressed surfaces lying in a cylindrical surface co-axial with the non-depressed surface and diverging from one another in the direction of rotation and obliquely arranged with respect to the axis of rotation of said rotatable member to cause a flow of lubricant longitudinally along said depressed surfaces from their distal ends toward their proximal ends as said bearing members are rotated relative to one another, each depressed surface subtending a radial angle of less than 360.degree., and means for rotating said rotatable bearing member to cause the lubricant to flow across and along said depressed surfaces, the flow of lubricant being impeded by the non-depressed portions of said second bearing surface to cause an increase in the lubricant pressure.

Policy and practice of radioactive waste management in India

International Nuclear Information System (INIS)

Sunder Radzhan, N.S.

1986-01-01

The Indian program on radioactive waste management comprising two main variants: engineering subsurface repositories for low- and intermediate-level wastes and deep geological formations for alpha-bearing and high-level wastes (HLW) is presented. One of the problems deals with the matrices with improved properties for HLW inclusion. The other aspect concerns development of management with alpha-emitting radionuclides in HLW. Special attention is paid to the problems of safety

Camshaft bearing arrangement

Energy Technology Data Exchange (ETDEWEB)

Aoi, K.; Ozawa, T.

1986-06-10

A bearing arrangement is described for the camshaft of an internal combustion engine or the like which camshaft is formed along its length in axial order with a first bearing surface, a first cam lobe, a second bearing surface, a second cam lobe, a third bearing surface, a third cam lobe and a fourth bearing surface, the improvement comprising first bearing means extending around substantially the full circumference of the first bearing surface and journaling the first bearing surface, second bearing means extending around substantially less than the circumference of the second bearing surface and journaling the second bearing surface, third bearing means extending around substantially less than the circumference of the third bearing surface and journaling the third bearing surface, and fourth bearing means extending around substantially the full circumference of the fourth bearing surface and journaling the first bearing surface.

Study of the thermal and mechanical sensitivity of bitumen/oxygen salt mixtures

International Nuclear Information System (INIS)

Backof, E.; Diepold, W.

1975-07-01

The safe handling characteristics of radioactive wastes containing nitrate salts to be fixed in bitumen for ultimate storage in salt mines according to a process developed at the Karlsruhe Nuclear Research Center have been examined with respect to their combustibility and shock sensitivity in tests of inactive bitumen/salt mixtures. Samples containing 40% bitumen and 60% nitrates of alkali, alkaline earth, and heavy metals, organic acids and rare earths were used to determine the thermal sensitivity (ignition temperature, duration of burning, heating under contained conditions), the mechanical sensitivity (shock sensitivity) and, in order to simulate major shock stresses, the sensitivity against detonation stresses. A few basic experiments were also performed on some beta-irradiated inactive samples. It appeared that although the addition of nitrates increased the combustibility of bitumen, neither the high thermal nor the detonation stresses resulted in any explosion-type reaction. (orig.) [de

Extension of the M-D model for treating stress drops in salt

International Nuclear Information System (INIS)

Munson, D.E.; DeVries, K.L.; Fossum, A.F.; Callahan, G.D.

1993-01-01

Development of the multimechanism deformation (M-D) constitutive model for steady state creep, which incorporates irreversible workhardening and recovery transient strains, was motivated by the need to predict very long term closures in underground rooms for radioactive waste repositories in salt. The multimechanism deformation model for the creep deformation of salt is extended to treat the response of salt to imposed stress drops. Stress drop tests produce a very distinctive behavior where both reversible elastic strain and reversible time dependent strain occur. These transient strains are negative compared to the positive transient strains produced by the normal creep workhardening and recovery processes. A simple micromechanical evolutionary process is defined to account for the accumulation of these reversible strains, and their subsequent release with decreases in stress. A number of experimental stress drop tests for various stress drop magnitudes and temperatures are adequately simulated with the model

Disposal of high-level waste from nuclear power plants in Denmark. Salt dome investigations. v.5

International Nuclear Information System (INIS)

1981-01-01

The present report deals with safety evaluation as part of the investigations regarding a repository for high-level waste in a salt dome. It is volume 5 of five volumes that together constitute the final report on the Danish utilities' salt dome investigations. Two characteristics of the waste are of special importance for the safety evaluation: the encasing of the waste in steel casks with 15 cm thick walls affording protection against corrosion, protecting the surroundings against radiation, and protecting the glass cylinders from mechanical damage resulting from the pressure at the bottom of the disposal hole, and the modest generation of heat in the waste at the time of disposal resulting in a maximum temperature increase in the salt close to the waste of approx. 40 deg. C. These characteristics proved to considerably improve the safety margin with respect to unforeseen circumstances. The character of the salt dome and of the salt in the proposed disposal area offers in itself good protection against contact with the ground water outside the dome. The relatively large depth of 1200 and 2500 m of the salt surface also means that neither dome nor disposal facility will be appreciably influenced by glaciations or earthquakes. The chalk above the proposed disposal area is very tight and to retain radioactive matter effectively even in the precence of high concentrations of NaCL. The safety investigations included a number of natural processes and probable events such as the segregation of crystal water from overlooked salt minerals, faulty sealings of disposal holes, permeable fault zones in the chalk overlying the dome, the risk in connection with human penetration into the dome. These conditions will neither lead to the destruction of the waste casks or to the release of waste from the dome. Leaching of a cavern is the only situation which proved to result in a release of radioactive material to the biosphere, but the resulting doses was found to be small

Computer-based supervisory control and data acquisition system for the radioactive waste evaporator

International Nuclear Information System (INIS)

Pope, N.G.; Schreiber, S.B.; Yarbro, S.L.; Gomez, B.G.; Nekimken, H.L.; Sanchez, D.E.; Bibeau, R.A.; Macdonald, J.M.

1994-12-01

The evaporator process at TA-55 reduces the amount of transuranic liquid radioactive waste by separating radioactive salts from relatively low-level radioactive nitric acid solution. A computer-based supervisory control and data acquisition (SCADA) system has been installed on the process that allows the operators to easily interface with process equipment. Individual single-loop controllers in the SCADA system allow more precise process operation with less human intervention. With this system, process data can be archieved in computer files for later analysis. Data are distributed throughout the TA-55 site through a local area network so that real-time process conditions can be monitored at multiple locations. The entire system has been built using commercially available hardware and software components

Deep geologic disposal of mixed waste in bedded salt: The Waste Isolation Pilot Plant

International Nuclear Information System (INIS)

Rempe, N.T.

1993-01-01

Mixed waste (i.e., waste that contains both chemically hazardous and radioactive components) poses a moral, political, and technical challenge to present and future generations. But an international consensus is emerging that harmful byproducts and residues can be permanently isolated from the biosphere in a safe and environmentally responsible manner by deep geologic disposal. To investigate and demonstrate such disposal for transuranic mixed waste, derived from defense-related activities, the US Department of Energy has prepared the Waste Isolation Pilot Plant (WIPP) near Carlsbad, New Mexico. This research and development facility was excavated approximately at the center of a 600 m thick sequence of salt (halite) beds, 655 m below the surface. Proof of the long-term tectonic and hydrological stability of the region is supplied by the fact that these salt beds have remained essentially undisturbed since they were deposited during the Late Permian age, approximately 225 million years ago. Plutonium-239, the main radioactive component of transuranic mixed waste, has a half-life of 24,500 years. Even ten half-lives of this isotope - amounting to about a quarter million years, the time during which its activity will decline to background level represent only 0.11 percent of the history of the repository medium. Therefore, deep geologic disposal of transuranic mixed waste in Permian bedded salt appears eminently feasible

The Design of High Reliability Magnetic Bearing Systems for Helium Cooled Reactor Machinery

International Nuclear Information System (INIS)

Swann, M.; Davies, N.; Jayawant, R.; Leung, R.; Shultz, R.; Gao, R.; Guo, Z.

2014-01-01

The requirements for magnetic bearing equipped machinery used in high temperature, helium cooled, graphite moderated reactor applications present a set of design considerations that are unlike most other applications of magnetic bearing technology in large industrial rotating equipment, for example as used in the oil and gas or other power generation applications. In particular, the bearings are typically immersed directly in the process gas in order to take advantage of the design simplicity that comes about from the elimination of ancillary lubrication and cooling systems for bearings and seals. Such duty means that the bearings will usually see high temperatures and pressures in service and will also typically be subject to graphite particulate and attendant radioactive contamination over time. In addition, unlike most industrial applications, seismic loading events become of paramount importance for the magnetic bearings system, both for actuators and controls. The auxiliary bearing design requirements, in particular, become especially demanding when one considers that the whole mechanical structure of the magnetic bearing system is located inside an inaccessible pressure vessel that should be rarely, if ever, disassembled over the service life of the power plant. Lastly, many machinery designs for gas cooled nuclear power plants utilize vertical orientation. This circumstance presents its own unique requirements for the machinery dynamics and bearing loads. Based on the authors’ experience with machine design and supply on several helium cooled reactor projects including Ft. St. Vrain (US), GT-MHR (Russia), PBMR (South Africa), GTHTR (Japan), and most recently HTR-PM (China), this paper addresses many of the design considerations for such machinery and how the application of magnetic bearings directly affects machinery reliability and availability, operability, and maintainability. Remote inspection and diagnostics are a key focus of this paper. (author)

REAKTOR INNOVATIVE MOLTEN SALT (IMSR DENGAN SISTEM KESELAMATAN PASIF MENYELURUH

Directory of Open Access Journals (Sweden)

Andang Widiharto

2015-04-01

Full Text Available Pengembangan Teknologi Reaktor Nuklir pada masa mendatang mengarah pada peningkatan aspek keselamatan, peningkatan pendayagunaan bahan bakar, reduksi limbah radioaktif, ketahanan terhadap proliferasi bahan-bakar nuklir dan peningkatan aspek ekonomi. reaktor Innovative Molten Salt (IMSR adalah reaktor nuklir yang menggunakan bahan bakar cair berupa garam lebur fluoride (7LiF-ThF4-UF4-MaFx. Reaktor IMSR didesain sebagai reaktor pembiak termal, yaitu membiakkan U-233 dari Th-232. Hal ini untuk menjawab permasalahan sustainabilitas ketersedian sumber daya bahan bakar nuklir dan reduksi limbah radioaktif. Dalam aspek keselamatan, desain reaktor IMSR memiliki sifat inherent safe, yaitu koefisien umpan balik daya yang negatif serta memiliki fitur-fitur keselamatan pasif. Fitur-fitur keselamatan pasif terdiri dari sistem shutdown pasif, sistem pendinginan pasif pasca shutdown serta sistem pendinginan pasif untuk produk fisi. Kecelakaan yang berpotensi terjadi pada IMSR, yaitu kecelakaan kehilangan aliran bahan bakar, kecelakaan kehilangan aliran pendingin, kecelakaan kehilangan kemampuan pengambilan kalor serta kecelakaan kerusakan integritas sistem reaktor, dapat ditangani sepenuhnya secara pasif hingga mencapai kondisi shutdown selamat. Kata kunci: keselamatan pasif, inherent safe, IMSR   The next Nuclear Reactor Technology developments are directed to the increasing of the aspects of safety, fuel utility, radioactive waste reduction, proliferation retention and economy. Innovative Molten Salt Reactor (IMSR is a nuclear reactor design that uses fluoride molten salt (7LiF-ThF4-UF4-MaFx. IMSR is designed as a thermal breeder reactor, i.e. to produce U-233 from Th-232. This is the answer of natural nuclear fuel sustainability and radioactive waste problems. In term of safety aspect, IMSR design has inherent safe characteristics, i.e. negative power feedback coefficient, and passive safety features. The passive safety features are passive shutdown

Transport of radioactive materials of the C. A. E. [CEA (France)]. Le transport des matieres radioactives au C.E.A.

Energy Technology Data Exchange (ETDEWEB)

Labrousse, M.

1974-03-15

Since the publication, in 1967, of the two issues of the Bull. Inform. Sci. Tech. devoted to the transport of radioactive materials, an important evolution has taken place, bearing both on the nature of the transports--where natural uranium hexafluoride, irradiated fuel, and wastes are becoming comparatively more important than miscellaneous small packages--and the construction of packagings, which are becoming more and more elaborate. This evolution appears in the reports selected for the BIST that are briefly introduced. (8 fig.)

Amount and nature of occluded water in bedded salt, Palo Duro Basin, Texas

International Nuclear Information System (INIS)

Fisher, R.S.

1987-01-01

The quantity and types of fluids within bedded salt cores from the Permian San Andres Formation, Palo Duro, Texas, were evaluated at the Texas Bureau of Economic Geology. Bedded halite from the San Andres Formation and other salt-bearing units were selected to represent the variety of salt types present, and were then analyzed. The mean water content of ''pure'' samples (more than 90% halite) is 0.4 weight percent, with none observed greater than 1.0 weight percent. Samples that contain more than 10 weight percent clay or mudstone display a trend of increasing water content with increasing clastic material. Chaotic mudstone-halite samples have as much as 5 weight percent water; halite-cemented mudstone interlayers, common throughout the bedded salts, may have water content values as high as 10 to 15 weight percent based on extrapolation of existing data that range from 0 to about 6%. No significant difference exists between the mean water content values of ''pure salt'' from the upper San Andres, lower San Andres Cycle 5, and lower San Andres Cycle 4 salt units. The fraction of total water present as mobile intergranular water is highly variable and not readily predicted from observed properties of the salt sample. The amount of water that would be affected by a high-level nuclear waste repository can be estimated if the volume of halite, the volume of clastic interlayers, and the amount and type of impurity in halite are known. Appendix contains seven vugraphs

Effects of gaseous radioactive nuclides on the design and operation of repositories for spent LWR fuel in rock salt

International Nuclear Information System (INIS)

Jenks, G.H.

1979-12-01

Information relating to the identities and amounts of gaseous radionuclides present in spent LWR fuel and to their release from canistered spent fuel under plausible storage and disposal conditions was assembled, reviewed, and analyzed. Information was also reviewed and analyzed on several other subjects that relate to the integrity of the carbon steel canister in which the spent fuel is to be encapsulated and to the expected rates of transfer of gaseous radionuclides through crushed salt backfill within a disposal room in a reference repository in rock salt. The advantages and disadvantages were considered for several different canister-backfill materials, and recommendations were made regarding preferred materials. Other recommendations relate to encapsulation procedures and specifications and to needs for additional experimental studies. The objective of this work was to provide reference information, conclusions, and recommendations that could be used to establish design and operating conditions and procedures for a bedded salt repository for spent LWR fuel and that could also be used to help evaluate the safety of the repository. The results of this work will also generally apply to spent fuel repositories in domal salt. However, because the domal salt may have little or no brine inclusions within it, there may be little or no possibility that brine will migrate into open spaces around an emplaced canister. Addordingly, some of the concerns that result from the possible occurrence of brine migration in bedded salt may be of no importance in domal salt

Computer simulation of an internally pressurized radioactive waste disposal room in a bedded salt formation

International Nuclear Information System (INIS)

Brown, W.T.; Weatherby, J.R.

1991-01-01

The Waste Isolation Pilot Plant (WIPP) near Carlsbad, New Mexico was created by the U.S. Department of Energy as an underground research and development facility to demonstrate the safe storage of transuranic waste generated from defense activities. This facility consists of storage rooms mined from a bedded salt formation at a depth of about 650 meters. Each room will accommodate about 6800 55-gallon drums filled with waste. After waste containers are emplaced, the storage rooms are to be backfilled with mined salt or other backfill materials. As time passes, reconsolidation of this backfill will reduce the hydraulic conductivity of the room. However, gases produced by decomposition and corrosion of waste and waste containers may cause a slow build-up of pressure which can retard consolidation of the waste and backfilled salt. The authors have developed a finite-element model of an idealized disposal room which is assumed to be perfectly sealed. The assumption that no gas escapes from the disposal room is a highly idealized and extreme condition which does not account for leakage paths, such as interbeds, that exist in the surrounding salt formation. This model has been used in a parametric study to determine how reconsolidation is influenced by various assumed gas generation rates and total amounts of gas generated. Results show that reductions in the gas generation, relative to the baseline case, can increase the degree of consolidation and reduce the peak gas pressure in disposal rooms. Even higher degrees of reconsolidation can be achieved by reducing both amounts and rates of gas generation. 8 refs., 4 figs., 1 tab

Low- and intermediate level radioactive waste from Risoe, Denmark. Location studies for potential disposal areas. Report no. 4. Characterization and description of areas. Bornholm

Energy Technology Data Exchange (ETDEWEB)

Gravesen, P.; Nilsson, B.; Schack Pedersen, S.A.; Binderup, M.

2011-07-01

The low - and intermediate level radioactive waste from Risoe: the nuclear reactor buildings, different types of material from the research periods and waste from hospitals and research institutes have to be stored in a final disposal in Denmark for at least 300 years. The task is to locate and recognize sediments or rocks with low permeability which can isolate the radioactive waste from the surrounding deposits, the groundwater resources, the recipients and from human activities. The sediments or rocks shall also act as a protection if the waste disposal leaks radioactive material to the surroundings. This goal can be reached by low water flow possibilities and high sorption potentials of the sediments or rocks. The investigation of geological deposits as potential waste disposals for high radioactive waste from nuclear power plants has earlier been focused on deep seated salt deposits and basement rocks, but the Tertiary clays were also mapped. The salt diapirs and the salt deposits are not included in the present study. The task is to find approximately 20 areas where a waste disposal potentially can be located. The 20 areas have to be reduced to 2-3 more precise locations, where detailed field investigations of the geological, hydrogeological-hydrochemical and technical conditions will be performed. The present report describes areas 1 and 2 on Bornholm, East Denmark. (LN)

Low- and intermediate level radioactive waste from Risoe, Denmark. Location studies for potential disposal areas. Report no. 4. Characterization and description of areas. Bornholm

International Nuclear Information System (INIS)

Gravesen, P.; Nilsson, B.; Schack Pedersen, S.A.; Binderup, M.

2011-01-01

The low - and intermediate level radioactive waste from Risoe: the nuclear reactor buildings, different types of material from the research periods and waste from hospitals and research institutes have to be stored in a final disposal in Denmark for at least 300 years. The task is to locate and recognize sediments or rocks with low permeability which can isolate the radioactive waste from the surrounding deposits, the groundwater resources, the recipients and from human activities. The sediments or rocks shall also act as a protection if the waste disposal leaks radioactive material to the surroundings. This goal can be reached by low water flow possibilities and high sorption potentials of the sediments or rocks. The investigation of geological deposits as potential waste disposals for high radioactive waste from nuclear power plants has earlier been focused on deep seated salt deposits and basement rocks, but the Tertiary clays were also mapped. The salt diapirs and the salt deposits are not included in the present study. The task is to find approximately 20 areas where a waste disposal potentially can be located. The 20 areas have to be reduced to 2-3 more precise locations, where detailed field investigations of the geological, hydrogeological-hydrochemical and technical conditions will be performed. The present report describes areas 1 and 2 on Bornholm, East Denmark. (LN)

Low- and intermediate level radioactive waste from Risoe, Denmark. Location studies for potential disposal areas. Report no. 6. Characterization and description of areas. Sjaelland

International Nuclear Information System (INIS)

Gravesen, P.; Nilsson, B.; Schack Pedersen, S.A.; Binderup, M.

2011-01-01

The low and intermediate level radioactive waste from Risoe: the nuclear reactor buildings, different types of material from the research periods and waste from hospitals and research institutes have to be stored in a final disposal in Denmark for at least 300 years. The task is to locate and recognize sediments or rocks with low permeability which can isolate the radioactive waste from the surrounding deposits, the groundwater resources, the recipients and from human activities. The sediments or rocks shall also act as a protection if the waste disposal leaks radioactive material to the surroundings. This goal can be reached by low water flow possibilities, strong sorption capacity for many radionuclides and self-sealing properties. The investigation of geological deposits as potential waste disposals for high radioactive waste from nuclear power plants has earlier focused on deep seated salt deposits and basement rocks. Nevertheless, the Tertiary clays were mapped as well. The salt diapirs and the salt deposits are not included in the present study. The task is to find approximately 20 areas potentially useful for a waste disposal. The 20 areas have to be reduced to 1-3 most potential locations where detailed field investigations of the geological, hydrogeological - hydrochemical and geotechnical conditions will be performed. The present report describes the areas 5 and 6 on Zealand. (LN)

Low- and intermediate level radioactive waste from Risoe, Denmark. Location studies for potential disposal areas. Report no. 8. Characterization and description of areas. OEstjylland

Energy Technology Data Exchange (ETDEWEB)

Gravesen, P.; Nilsson, B.; Schack Pedersen, S.A.; Binderup, M.

2011-07-01

The low and intermediate level radioactive waste from Risoe: the nuclear reactor buildings, different types of material from the research periods and waste from hospitals and research institutes have to be stored in a final disposal in Denmark for at least 300 years. The task is to locate and recognize sediments or rocks with low permeability which can isolate the radioactive waste from the surrounding deposits, the groundwater resources, the recipients and from human activities. The sediments or rocks shall also act as a protection if the waste disposal leaks radioactive material to the surroundings. This goal can be reached by low water flow possibilities, high sorption capacity for many radionuclides and self-sealing properties. The investigation of geological deposits as potential waste disposals for high radioactive waste from nuclear power plants has earlier focused on deep seated salt deposits and basement rocks. Nevertheless, the Tertiary clays were mapped as well. The salt diapirs and the salt deposits are not included in the present study. The task is to find approximately 20 areas where a waste disposal potentially can be located. The 20 areas have to be reduced to 1-3 most potential locations where detailed field investigations of the geological, hydrogeological - hydrochemical and geotechnical conditions will be performed. The present report describes the areas 12,13,14 and 15 in Eastern Jutland. (LN)

Low- and intermediate level radioactive waste from Risoe, Denmark. Location studies for potential disposal areas. Report no. 10. Characterization and description of areas. Nordjylland

Energy Technology Data Exchange (ETDEWEB)

Gravesen, P.; Nilsson, B.; Schack Pedersen, S.A.; Binderup, M.

2011-07-01

The low and intermediate level radioactive waste from Risoe: the nuclear reactor buildings, different types of material from the research periods and waste from hospitals and research institutes have to be stored in a final disposal in Denmark for at least 300 years. The task is to locate and recognize sediments or rocks with low permeability which can isolate the radioactive waste from the surrounding deposits, the groundwater resources, the recipients and from human activities. The sediments or rocks shall also act as a protection if the waste disposal leaks radioactive material to the surroundings. This goal can be reached by low water flow possibilities, strong sorption capacity for many radionuclides and self-sealing properties. The investigation of geological deposits as potential waste disposals for high radioactive waste from nuclear power plants has earlier focused on deep seated salt deposits and basement rocks. Nevertheless, the Tertiary clays were mapped as well. The salt diapirs and the salt deposits are not included in the present study. The task is to find approximately 20 areas potentially useful for a waste disposal. The 20 areas have to be reduced to 1-3 most potential locations where detailed field investigations of the geological, hydrogeological, hydrochemical and geotechnical conditions will be performed. The present report describes the area 22 in Northern Jutland. (LN)

Low- and intermediate level radioactive waste from Risoe, Denmark. Location studies for potential disposal areas. Report no. 9. Characterization and description of areas. Limfjorden

International Nuclear Information System (INIS)

Gravesen, P.; Nilsson, B.; Schack Pedersen, S.A.; Binderup, M.

2011-01-01

The low and intermediate level radioactive waste from Risoe: the nuclear reactor buildings, different types of material from the research periods and waste from hospitals and research institutes have to be stored in a final disposal in Denmark for at least 300 years. The task is to locate and recognize sediments or rocks with low permeability which can isolate the radioactive waste from the surrounding deposits, the groundwater resources, the recipients and from human activities. The sediments or rocks shall also act as a protection if the waste disposal leaks radioactive material to the surroundings. This goal can be reached by low water flow possibilities, strong sorption capacity for many radionuclides and self-sealing properties. The investigation of geological deposits as potential waste disposals for high radioactive waste from nuclear power plants has earlier focused on deep seated salt deposits and basement rocks. Nevertheless, the Tertiary clays were mapped as well. The salt diapirs and the salt deposits are not included in the present study. The task is to find approximately 20 areas potentially useful for a waste disposal. The 20 areas have to be reduced to 1-3 most potential locations where detailed field investigations of the geological, hydrogeological, hydrochemical and geotechnical conditions will be performed. The present report describes the areas 16,17,18,19,20 and 21 around Limfjorden. (LN)

Low- and intermediate level radioactive waste from Risoe, Denmark. Location studies for potential disposal areas. Report no. 8. Characterization and description of areas. Oestjylland

International Nuclear Information System (INIS)

Gravesen, P.; Nilsson, B.; Schack Pedersen, S.A.; Binderup, M.

2011-01-01

The low and intermediate level radioactive waste from Risoe: the nuclear reactor buildings, different types of material from the research periods and waste from hospitals and research institutes have to be stored in a final disposal in Denmark for at least 300 years. The task is to locate and recognize sediments or rocks with low permeability which can isolate the radioactive waste from the surrounding deposits, the groundwater resources, the recipients and from human activities. The sediments or rocks shall also act as a protection if the waste disposal leaks radioactive material to the surroundings. This goal can be reached by low water flow possibilities, high sorption capacity for many radionuclides and self-sealing properties. The investigation of geological deposits as potential waste disposals for high radioactive waste from nuclear power plants has earlier focused on deep seated salt deposits and basement rocks. Nevertheless, the Tertiary clays were mapped as well. The salt diapirs and the salt deposits are not included in the present study. The task is to find approximately 20 areas where a waste disposal potentially can be located. The 20 areas have to be reduced to 1-3 most potential locations where detailed field investigations of the geological, hydrogeological - hydrochemical and geotechnical conditions will be performed. The present report describes the areas 12,13,14 and 15 in Eastern Jutland. (LN)

Low- and intermediate level radioactive waste from Risoe, Denmark. Location studies for potential disposal areas. Report no. 9. Characterization and description of areas. Limfjorden

Energy Technology Data Exchange (ETDEWEB)

Gravesen, P.; Nilsson, B.; Schack Pedersen, S.A.; Binderup, M.

2011-07-01

The low and intermediate level radioactive waste from Risoe: the nuclear reactor buildings, different types of material from the research periods and waste from hospitals and research institutes have to be stored in a final disposal in Denmark for at least 300 years. The task is to locate and recognize sediments or rocks with low permeability which can isolate the radioactive waste from the surrounding deposits, the groundwater resources, the recipients and from human activities. The sediments or rocks shall also act as a protection if the waste disposal leaks radioactive material to the surroundings. This goal can be reached by low water flow possibilities, strong sorption capacity for many radionuclides and self-sealing properties. The investigation of geological deposits as potential waste disposals for high radioactive waste from nuclear power plants has earlier focused on deep seated salt deposits and basement rocks. Nevertheless, the Tertiary clays were mapped as well. The salt diapirs and the salt deposits are not included in the present study. The task is to find approximately 20 areas potentially useful for a waste disposal. The 20 areas have to be reduced to 1-3 most potential locations where detailed field investigations of the geological, hydrogeological, hydrochemical and geotechnical conditions will be performed. The present report describes the areas 16,17,18,19,20 and 21 around Limfjorden. (LN)

Low- and intermediate level radioactive waste from Risoe, Denmark. Location studies for potential disposal areas. Report no. 6. Characterization and description of areas. Sjaelland

Energy Technology Data Exchange (ETDEWEB)

Gravesen, P.; Nilsson, B.; Schack Pedersen, S.A.; Binderup, M.

2011-07-01

The low and intermediate level radioactive waste from Risoe: the nuclear reactor buildings, different types of material from the research periods and waste from hospitals and research institutes have to be stored in a final disposal in Denmark for at least 300 years. The task is to locate and recognize sediments or rocks with low permeability which can isolate the radioactive waste from the surrounding deposits, the groundwater resources, the recipients and from human activities. The sediments or rocks shall also act as a protection if the waste disposal leaks radioactive material to the surroundings. This goal can be reached by low water flow possibilities, strong sorption capacity for many radionuclides and self-sealing properties. The investigation of geological deposits as potential waste disposals for high radioactive waste from nuclear power plants has earlier focused on deep seated salt deposits and basement rocks. Nevertheless, the Tertiary clays were mapped as well. The salt diapirs and the salt deposits are not included in the present study. The task is to find approximately 20 areas potentially useful for a waste disposal. The 20 areas have to be reduced to 1-3 most potential locations where detailed field investigations of the geological, hydrogeological - hydrochemical and geotechnical conditions will be performed. The present report describes the areas 5 and 6 on Zealand. (LN)

New scenario for the accumulation and release of radiation damage in rock salt and related materials

NARCIS (Netherlands)