WorldWideScience

Sample records for hlw solubility experiments

  1. The solubilities of significant organic compounds in HLW tank supernate solutions -- FY 1995 progress report

    International Nuclear Information System (INIS)

    Barney, G.S.

    1996-01-01

    At the Hanford Site organic compounds were measured in tank supernate simulant solutions during FY 1995. This solubility information will be used to determine if these organic salts could exist in solid phases (saltcake or sludges) in the waste where they might react violently with the nitrate or nitrite salts present in the tanks. Solubilities of sodium glycolate, succinate, and caproate salts; iron and aluminum and butylphosphate salts; and aluminum oxalate were measured in simulated waste supernate solutions at 25 degree C, 30 degree C, 40 degree C, and 50 degree C. The organic compounds were selected because they are expected to exist in relatively high concentrations in the tanks. The solubilities of sodium glycolate, succinate, caproate, and butylphosphate in HLW tank supernate solutions were high over the temperature and sodium hydroxide concentration ranges expected in the tanks. High solubilities will prevent solid sodium salts of these organic acids from precipitating from tank supernate solutions. The total organic carbon concentrations (YOC) of actual tank supernates are generally much lower than the TOC ranges for simulated supernate solutions saturated (at the solubility limit) with the organic salts. This is so even if all the dissolved carbon in a given tank and supernate is due to only one of these eight soluble compounds (an unlikely situation). Metal ion complexes of and butylphosphate and oxalate in supernate solutions were not stable in the presence of the hydroxide concentrations expected in most tanks. Iron and aluminum dibutylphosphate compounds reacted with hydroxide to form soluble sodium dibutylphosphate and precipitated iron and aluminum hydroxides. Aluminum oxalate complexes were also not stable in the basic simulated supernate solutions. Solubilities of all the organic salts decrease with increasing sodium hydroxide concentration because of the common ion effect of Na+. Increasing temperatures raised the solubilities of the organic

  2. Spent fuel and HLW transportation the French experience

    International Nuclear Information System (INIS)

    Giraud, J.P.; Charles, J.L.

    1995-01-01

    With 53 nuclear power plants in operation at EDF and a fuel cycle with recycling policy of the valuable materials, COGEMA is faced with the transport of a wide range of radioactive materials. In this framework, the transport activity is a key link in closing the fuel cycle. COGEMA has developed a comprehensive Transport Organization System dealing with all the sectors of the fuel cycle. The paper will describe the status of transportation of spent fuel and HLW in France and the experience gathered. The Transport Organization System clearly defines the role of all actors where COGEMA, acting as the general coordinator, specifies the tasks to be performed and brings technical and commercial support to its various subcontractors: TRANSNUCLEAIRE, specialized in casks engineering and transport operations, supplies packaging and performs transport operations, LEMARECHAL and CELESTIN operate transport by truck in the Vicinity of the nuclear sites while French Railways are in charge of spent fuel transport by train. HLW issued from the French nuclear program is stored for 30 years in an intermediate storage installation located at the La Hague reprocessing plant. Ultimately, these canisters will be transported to the disposal site. COGEMA has set up a comprehensive transport organization covering all operational aspects including adapted procedures, maintenance programs and personnel qualification

  3. The experiment of affective web risk communication on HLW geological disposal

    International Nuclear Information System (INIS)

    Kugo, Akihide; Yoshikawa, Eiwa; Wakabayashi, Yasunaga; Shimoda, Hiroshi; Uda, Akinobu; Ito, Kyoko

    2006-01-01

    Dialog mode web contents regarding the HLW risk is effective to altruism. To make it more effectively, we introduced affective elements such as facial expression of character agents and sympathetic response on the BBS by experts, which brought us smooth risk communication. This paper describes the result of preliminary experiments surrounding the affective ways to communicate on the risk of HLW geological disposal, leading to enhance the social cooperation, and the public open experiment for one month on the Web. (author)

  4. The solubilities of significant organic compounds in HLW tank supernate solutions

    International Nuclear Information System (INIS)

    Barney, G.S.

    1994-08-01

    Large quantities of organic chemicals used in reprocessing spent nuclear-fuels at the Hanford Site have accumulated in underground high-level radioactive waste tanks. The organic content of these tanks must he known so that the potential for hazardous reactions between organic components and sodium nitrate/nitrite salts in the waste can he evaluated. The solubilities of organic compounds described in this report will help determine if they are present in the solid phases (salt cake and sludges) as well as the liquid phase (interstitial liquor/supernate) in the tanks. The solubilities of five significant sodium salts of carboxylic acids and aminocarboxylic acids [sodium oxalate, formate, citrate, nitrilotriacetate (NTA) and ethylendiaminetetraacetate (EDTA)] were measured in a simulated supernate solution at 25 degrees C, 30 degrees C, 40 degrees C, and 50 degrees C

  5. HLW Tank Space Management, Final Report

    International Nuclear Information System (INIS)

    Sessions, J.

    1999-01-01

    The HLW Tank Space Management Team (SM Team) was chartered to select and recommend an HLW Tank Space Management Strategy (Strategy) for the HLW Management Division of Westinghouse Savannah River Co. (WSRC) until an alternative salt disposition process is operational. Because the alternative salt disposition process will not be available to remove soluble radionuclides in HLW until 2009, the selected Strategy must assure that it safely receives and stores HLW at least until 2009 while continuing to supply sludge slurry to the DWPF vitrification process

  6. Crystallization In High Level Waste (HLW) Glass Melters: Operational Experience From The Savannah River Site

    Energy Technology Data Exchange (ETDEWEB)

    Fox, K. M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2014-02-27

    processing strategy for the Hanford Tank Waste Treatment and Immobilization Plant (WTP). The basis of this alternative approach is an empirical model predicting the crystal accumulation in the WTP glass discharge riser and melter bottom as a function of glass composition, time, and temperature. When coupled with an associated operating limit (e.g., the maximum tolerable thickness of an accumulated layer of crystals), this model could then be integrated into the process control algorithms to formulate crystal tolerant high level waste (HLW) glasses targeting higher waste loadings while still meeting process related limits and melter lifetime expectancies. This report provides a review of the scaled melter testing that was completed in support of the Defense Waste Processing Facility (DWPF) melter. Testing with scaled melters provided the data to define the DWPF operating limits to avoid bulk (volume) crystallization in the un-agitated DWPF melter and provided the data to distinguish between spinels generated by K-3 refractory corrosion versus spinels that precipitated from the HLW glass melt pool. This report includes a review of the crystallization observed with the scaled melters and the full scale DWPF melters (DWPF Melter 1 and DWPF Melter 2). Examples of actual DWPF melter attainment with Melter 2 are given. The intent is to provide an overview of lessons learned, including some example data, that can be used to advance the development and implementation of an empirical model and operating limit for crystal accumulation for WTP. Operation of the first and second (current) DWPF melters has demonstrated that the strategy of using a liquidus temperature predictive model combined with a 100 °C offset from the normal melter operating temperature of 1150 °C (i.e., the predicted liquidus temperature (TL) of the glass must be 1050 °C or less) has been successful in preventing any detrimental accumulation of spinel in the DWPF melt pool, and spinel has not been

  7. Crystallization in high level waste (HLW) glass melters: Savannah River Site operational experience

    Energy Technology Data Exchange (ETDEWEB)

    Fox, Kevin M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Peeler, David K. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Kruger, Albert A. [USDOE Office of River Protection, Richland, WA (United States)

    2015-06-12

    This paper provides a review of the scaled melter testing that was completed for design input to the Defense Waste Processing Facility (DWPF) melter. Testing with prototype melters provided the data to define the DWPF operating limits to avoid bulk (volume) crystallization in the un-agitated DWPF melter and provided the data to distinguish between spinels generated by refractory corrosion versus spinels that precipitated from the HLW glass melt pool. A review of the crystallization observed with the prototype melters and the full-scale DWPF melters (DWPF Melter 1 and DWPF Melter 2) is included. Examples of actual DWPF melter attainment with Melter 2 are given. The intent is to provide an overview of lessons learned, including some example data, that can be used to advance the development and implementation of an empirical model and operating limit for crystal accumulation for a waste treatment and immobilization plant.

  8. Counter current decantation washing of HLW sludge

    International Nuclear Information System (INIS)

    Brooke, J.N.; Peterson, R.A.

    1997-01-01

    The Savannah River Site (SRS) has 51 High Level Waste (HLW) tanks with typical dimensions 25.9 meters (85 feet) diameter and 10 meters (33 feet) high. Nearly 114 million liters (30 M gallons) of HLW waste is stored in these tanks in the form of insoluble solids called sludge, crystallized salt called salt cake, and salt solutions. This waste is being converted to waste forms stable for long term storage. In one of the processes, soluble salts are washed from HLW sludge in preparation for vitrification. At present, sludge is batch washed in a waste tank with one or no reuse of the wash water. Sodium hydroxide and sodium nitrite are added to the wash water for tank corrosion protection; the large volumes of spent wash water are recycled to the evaporator system; additional salt cake is produced; and sodium carbonate is formed in the washed sludge during storage by reaction with CO 2 from the air. High costs and operational concerns with the current washing process prompts DOE and WSRC to seek an improved washing method. A new method should take full advantage of the physical/chemical properties of sludge, experience from other technical disciplines, processing rate requirements, inherent process safety, and use of proven processes and equipment. Counter current solids washing is a common process in the minerals processing and chemical industries. Washing circuits can be designed using thickeners, filters or centrifuges. Realizing the special needs of nuclear work and the low processing rates required, a Counter Current Decantation (CCD) circuit is proposed using small thickeners and fluidic pumps

  9. Preliminary on-surface experiments for backfilling a HLW repository: the ESDRED project

    International Nuclear Information System (INIS)

    Bastiaens, W.

    2007-01-01

    ESDRED is a technological integrated project within the context of the Sixth Framework Program of EURATOM. The project aims to demonstrate the technical feasibility at an industrial scale of specific technologies related to the construction, operation and closure of a deep geological repository for spent fuel and long-lived radioactive waste. The Belgian design for high level waste disposal is based on the so-called Supercontainer concept. Within this concept, the waste is encased in a carbon steel overpack, which is consequently fitted into a 70 cm thick concrete shell, in its turn enveloped by a stainless steel liner. A Supercontainer measures about 2 m in diameter. In the design of the repository, the Supercontainers will be emplaced, one after the other, in disposal galleries. The space between the Supercontainers and the gallery lining needs to be filled up with a solid material. The most essential function of this component, referred to as backfill, is to prevent a collapse of the gallery. A secondary function is to limit the presence of free oxygen, to limit corrosion. In the ESDRED project EIG EURIDICE, together with SCK-CEN and ONDRAF/NIRAS, investigates technologies to apply the backfill. Two options to apply the backfill were investigated within the ESDRED project: fill the gap with a granular material and backfill the gap with a grout. The prime operational target will be to achieve a 100 percent filling of the gap. A wide variety of materials was tested. A number of considerations regarding long-term safety and operational feasibility impose constraints on the backfill component:it should preserve the corrosion-protective environment established by the Supercontainer; it should not act as a thermal isolator; it should not introduce organic materials that can give rise to the formation of migration-enhancing complexes between radionuclides and soluble organic compounds; it should be feasible to construct at a sufficiently high rate; the strength of the

  10. Technetium Chemistry in HLW

    International Nuclear Information System (INIS)

    Hess, Nancy J.; Felmy, Andrew R.; Rosso, Kevin M.; Xia Yuanxian

    2005-01-01

    Tc contamination is found within the DOE complex at those sites whose mission involved extraction of plutonium from irradiated uranium fuel or isotopic enrichment of uranium. At the Hanford Site, chemical separations and extraction processes generated large amounts of high level and transuranic wastes that are currently stored in underground tanks. The waste from these extraction processes is currently stored in underground High Level Waste (HLW) tanks. However, the chemistry of the HLW in any given tank is greatly complicated by repeated efforts to reduce volume and recover isotopes. These processes ultimately resulted in mixing of waste streams from different processes. As a result, the chemistry and the fate of Tc in HLW tanks are not well understood. This lack of understanding has been made evident in the failed efforts to leach Tc from sludge and to remove Tc from supernatants prior to immobilization. Although recent interest in Tc chemistry has shifted from pretreatment chemistry to waste residuals, both needs are served by a fundamental understanding of Tc chemistry

  11. Viability for controlling long-term leaching of radionuclides from HLW glass by amorphous silica additives

    International Nuclear Information System (INIS)

    Inagaki, Y.; Uehara, S.

    2004-01-01

    Dissolution and deterioration experiments in coexistence system of amorphous silica and vitrified wastes have been executed in order to evaluating the effects of amorphous silica addition to high level radioactive vitrified waste (HLW glass) on suppression of nuclide leaching. Geo-chemical reaction mechanism among the vitrified waste, the amorphous silica and water was also evaluated. Dissolution of the silica network was suppressed by addition of the amorphous silica. However, the leaching of soluble nuclides like B proceeded depending on the hydration deterioration reaction. (A. Hishinuma)

  12. HLW Disposal System Development

    Energy Technology Data Exchange (ETDEWEB)

    Choi, J. W.; Choi, H. J.; Lee, J. Y. (and others)

    2007-06-15

    A KRS is suggested through design requirement analysis of the buffer and the canister which are the constituent of disposal system engineered barrier and HLW management plans are proposed. In the aspect of radionuclide retention capacity, the thickness of the buffer is determined 0.5m, the shape to be disc and ring and the dry density to be 1.6 g/cm{sup 3}. The maximum temperature of the buffer is below 100 .deg. which meets the design requirement. And bentonite blocks with 5 wt% of graphite showed more than 1.0 W/mK of thermal conductivity without the addition of sand. The result of the thermal analysis for proposed double-layered buffer shows that decrease of 7 .deg. C in maximum temperature of the buffer. For the disposal canister, the copper for the outer shell material and cast iron for the inner structure material is recommended considering the results analyzed in terms of performance of the canisters and manufacturability and the geochemical properties of deep groundwater sampled from the research area with granite, salt water intrusion, and the heavy weight of the canister. The results of safety analysis for the canister shows that the criticality for the normal case including uncertainty is the value of 0.816 which meets subcritical condition. Considering nation's 'Basic Plan for Electric Power Demand and Supply' and based on the scenario of disposing CANDU spent fuels in the first phase, the disposal system that the repository will be excavated in eight phases with the construction of the Underground Research Laboratory (URL) beginning in 2020 and commissioning in 2040 until the closure of the repository is proposed. Since there is close correlation between domestic HLW management plans and front-end/back-end fuel cycle plans causing such a great sensitivity of international environment factor, items related to assuring the non-proliferation and observing the international standard are showed to be the influential factor and acceptability

  13. HLW immobilization in glass

    International Nuclear Information System (INIS)

    Leroy, P.; Jacquet-Francillon, N.; Runge, S.

    1992-01-01

    The immobilization of High Level Waste in glass in France is a long history which started as early as in the 1950's. More than 30 years of Research and Development have been invested in that field. Two industrial facilities are operating (AVM and R7) and a third one (T7), under cold testing, is planned to start active operation in the mid-92. While vitrification has been demonstrated to be an industrially mastered process, the question of the quality of the final waste product, i.e. the HLW glass, must be addressed. The scope of the present paper is to focus on the latter point from both standpoints of the R and D and of the industrial reality

  14. Safety of HLW shipments

    International Nuclear Information System (INIS)

    1998-01-01

    The third shipment back to Japan of vitrified high-level radioactive waste (HLW) produced through reprocessing in France is scheduled to take place in early 1998. A consignment last March drew protest from interest groups and countries along the shipping route. Requirements governing the shipment of cargoes of this type and concerns raised by Greenpeace that were assessed by an international expert group, were examined in a previous article. A further report prepared on behalf of Greenpeace Pacific has been released. The paper: Transportation accident of a ship carrying vitrified high-level radioactive waste, Part 1 Impact on the Federated States of Micronesia by Resnikoff and Champion, is dated 31 July 1997. A considerable section of the report is given over to discussion of the economic situation of the Federated Statess of Micronesia, and lifestyle and dietary factors which would influence radiation doses arising from a release. It postulates a worst case accident scenario of a collision between the HLW transport ship and an oil tanker 1 km off Pohnpei with the wind in precisely the direction to result in maximum population exposure, and attempts to assess the consequences. In summary, the report postulates accident and exposure scenarios which are conceivable but not credible. It combines a series of worst case scenarios and attempts to evaluate the consequences. Both the combined scenario and consequences have probabilities of occurrence which are negligible. The shipment carried by the 'Pacific Swan' left Cherbourgon 21 January 1998 and comprised 30 tonnes of reprocessed vitrified waste in 60 stainless steel canisters loaded into three shipping casks. (author)

  15. Focusing on clay formation as host media of HLW geological disposal in China

    International Nuclear Information System (INIS)

    Zheng Hualing; Chen Shi; Sun Donghui

    2007-01-01

    Host medium is vitally important for safety for HLW geological disposal. Chinese HLW disposal effort in the past decades were mainly focused on granite formation. However, the granite formation has fatal disadvantage for HLW geological disposal. This paper reviews experiences gained and lessons learned in the international community and analyzes key factors affecting the site selection. It is recommended that clay formation should be taken into consideration and additional effort should be made before decision making of host media of HLW disposal in China. (authors)

  16. Long-term product consistency test of simulated 90-19/Nd HLW glass

    International Nuclear Information System (INIS)

    Gan, X.Y.; Zhang, Z.T.; Yuan, W.Y.; Wang, L.; Bai, Y.; Ma, H.

    2011-01-01

    Chemical durability of 90-19/Nd glass, a simulated high-level waste (HLW) glass in contact with the groundwater was investigated with a long-term product consistency test (PCT). Generally, it is difficult to observe the long term property of HLW glass due to the slow corrosion rate in a mild condition. In order to overcome this problem, increased contacting surface (S/V = 6000 m -1 ) and elevated temperature (150 o C) were employed to accelerate the glass corrosion evolution. The micro-morphological characteristics of the glass surface and the secondary minerals formed after the glass alteration were analyzed by SEM-EDS and XRD, and concentrations of elements in the leaching solution were determined by ICP-AES. In our experiments, two types of minerals, which have great impact on glass dissolution, were found to form on 90-19/Nd HLW glass surface when it was subjected to a long-term leaching in the groundwater. One is Mg-Fe-rich phyllosilicates with honeycomb structure; the other is aluminosilicates (zeolites). Mg and Fe in the leaching solution participated in the formation of phyllosilicates. The main components of phyllosilicates in alteration products of 90-19/Nd HLW glass are nontronite (Na 0.3 Fe 2 Si 4 O 10 (OH) 2 .4H 2 O) and montmorillonite (Ca 0.2 (Al,Mg) 2 Si 4 O 10 (OH) 2 .4H 2 O), and those of aluminosilicates are mordenite ((Na 2 ,K 2 ,Ca)Al 2 Si 10 O 24 .7H 2 O)) and clinoptilolite ((Na,K,Ca) 5 Al 6 Si 30 O 72 .18H 2 O). Minerals like Ca(Mg)SO 4 and CaCO 3 with low solubility limits are prone to form precipitant on the glass surface. Appearance of the phyllosilicates and aluminosilicates result in the dissolution rate of 90-19/Nd HLW glass resumed, which is increased by several times over the stable rate. As further dissolution of the glass, both B and Na in the glass were found to leach out in borax form.

  17. Development Of Glass Matrices For HLW Radioactive Wastes

    International Nuclear Information System (INIS)

    Jantzen, C.

    2010-01-01

    Vitrification is currently the most widely used technology for the treatment of high level radioactive wastes (HLW) throughout the world. Most of the nations that have generated HLW are immobilizing in either borosilicate glass or phosphate glass. One of the primary reasons that glass has become the most widely used immobilization media is the relative simplicity of the vitrification process, e.g. melt waste plus glass forming frit additives and cast. A second reason that glass has become widely used for HLW is that the short range order (SRO) and medium range order (MRO) found in glass atomistically bonds the radionuclides and governs the melt properties such as viscosity, resistivity, sulphate solubility. The molecular structure of glass controls contaminant/radionuclide release by establishing the distribution of ion exchange sites, hydrolysis sites, and the access of water to those sites. The molecular structure is flexible and hence accounts for the flexibility of glass formulations to waste variability. Nuclear waste glasses melt between 1050-1150 C which minimizes the volatility of radioactive components such as Tc 99 , Cs 137 , and I 129 . Nuclear waste glasses have good long term stability including irradiation resistance. Process control models based on the molecular structure of glass have been mechanistically derived and have been demonstrated to be accurate enough to control the world's largest HLW Joule heated ceramic melter in the US since 1996 at 95% confidence.

  18. DEVELOPMENT OF GLASS MATRICES FOR HLW RADIOACTIVE WASTES

    Energy Technology Data Exchange (ETDEWEB)

    Jantzen, C.

    2010-03-18

    Vitrification is currently the most widely used technology for the treatment of high level radioactive wastes (HLW) throughout the world. Most of the nations that have generated HLW are immobilizing in either borosilicate glass or phosphate glass. One of the primary reasons that glass has become the most widely used immobilization media is the relative simplicity of the vitrification process, e.g. melt waste plus glass forming frit additives and cast. A second reason that glass has become widely used for HLW is that the short range order (SRO) and medium range order (MRO) found in glass atomistically bonds the radionuclides and governs the melt properties such as viscosity, resistivity, sulphate solubility. The molecular structure of glass controls contaminant/radionuclide release by establishing the distribution of ion exchange sites, hydrolysis sites, and the access of water to those sites. The molecular structure is flexible and hence accounts for the flexibility of glass formulations to waste variability. Nuclear waste glasses melt between 1050-1150 C which minimizes the volatility of radioactive components such as Tc{sup 99}, Cs{sup 137}, and I{sup 129}. Nuclear waste glasses have good long term stability including irradiation resistance. Process control models based on the molecular structure of glass have been mechanistically derived and have been demonstrated to be accurate enough to control the world's largest HLW Joule heated ceramic melter in the US since 1996 at 95% confidence.

  19. Applicability of thermodynamic database of radioactive elements developed for the Japanese performance assessment of HLW repository

    International Nuclear Information System (INIS)

    Yui, Mikazu; Shibata, Masahiro; Rai, Dhanpat; Ochs, Michael

    2003-01-01

    In 1999 Japan Nuclear Cycle Development Institute (JNC) published a second progress report (also known as H12 report) on high-level radioactive waste (HLW) disposal in Japan (JNC 1999). This report helped to develop confidence in the selected HLW disposal system and to establish the implementation body in 2000 for the disposal of HLW. JNC developed an in-house thermodynamic database for radioactive elements for performance analysis of the engineered barrier system (EBS) and the geosphere for H12 report. This paper briefly presents the status of the JNC's thermodynamic database and its applicability to perform realistic analyses of the solubilities of radioactive elements, evolution of solubility-limiting solid phases, predictions of the redox state of Pu in the neutral pH range under reducing conditions, and to estimate solubilities of radioactive elements in cementitious conditions. (author)

  20. HLW Glass Studies: Development of Crystal-Tolerant HLW Glasses

    Energy Technology Data Exchange (ETDEWEB)

    Matyas, Josef; Huckleberry, Adam R.; Rodriguez, Carmen P.; Lang, Jesse B.; Owen, Antionette T.; Kruger, Albert A.

    2012-04-02

    In our study, a series of lab-scale crucible tests were performed on designed glasses of different compositions to further investigate and simulate the effect of Cr, Ni, Fe, Al, Li, and RuO2 on the accumulation rate of spinel crystals in the glass discharge riser of the HLW melter. The experimental data were used to expand the compositional region covered by an empirical model developed previously (Matyáš et al. 2010b), improving its predictive performance. We also investigated the mechanism for agglomeration of particles and impact of agglomerates on accumulation rate. In addition, the TL was measured as a function of temperature and composition.

  1. COMSOL Multiphysics Model for HLW Canister Filling

    Energy Technology Data Exchange (ETDEWEB)

    Kesterson, M. R. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-04-11

    The U.S. Department of Energy (DOE) is building a Tank Waste Treatment and Immobilization Plant (WTP) at the Hanford Site in Washington to remediate 55 million gallons of radioactive waste that is being temporarily stored in 177 underground tanks. Efforts are being made to increase the loading of Hanford tank wastes in glass while meeting melter lifetime expectancies and process, regulatory, and product quality requirements. Wastes containing high concentrations of Al2O3 and Na2O can contribute to nepheline (generally NaAlSiO4) crystallization, which can sharply reduce the chemical durability of high level waste (HLW) glass. Nepheline crystallization can occur during slow cooling of the glass within the stainless steel canister. The purpose of this work was to develop a model that can be used to predict temperatures of the glass in a WTP HLW canister during filling and cooling. The intent of the model is to support scoping work in the laboratory. It is not intended to provide precise predictions of temperature profiles, but rather to provide a simplified representation of glass cooling profiles within a full scale, WTP HLW canister under various glass pouring rates. These data will be used to support laboratory studies for an improved understanding of the mechanisms of nepheline crystallization. The model was created using COMSOL Multiphysics, a commercially available software. The model results were compared to available experimental data, TRR-PLT-080, and were found to yield sufficient results for the scoping nature of the study. The simulated temperatures were within 60 ºC for the centerline, 0.0762m (3 inch) from centerline, and 0.2286m (9 inch) from centerline thermocouples once the thermocouples were covered with glass. The temperature difference between the experimental and simulated values reduced to 40 ºC, 4 hours after the thermocouple was covered, and down to 20 ºC, 6 hours after the thermocouple was covered

  2. Waste Isolation Pilot Plant in situ experimental program for HLW

    International Nuclear Information System (INIS)

    Molecke, M.A.

    1977-01-01

    The Waste Isolation Pilot Plant (WIPP) will be a facility to demonstrate the environmental and operational safety of storing radioactive wastes in a deep geologic bedded salt facility. The WIPP will be located in southeastern New Mexico, approximately 30 miles east of the city of Carlsbad. The major focus of the pilot plant operation involves ERDA defense related low and intermediate-level transuranic wastes. The scope of the project also specifically includes experimentation utilizing commercially generated high-level wastes, or alternatively, spent unreprocessed fuel elements. WIPP HLW experiments are being conducted in an inter-related laboratory, bench-scale, and in situ mode. This presentation focuses on the planned in situ experiments which, depending on the availability of commercially reprocessed waste plus delays in the construction schedule of the WIPP, will begin in approximately 1985. Such experiments are necessary to validate preceding laboratory results and to provide actual, total conditions of geologic storage which cannot be adequately simulated. One set of planned experiments involves emplacing bare HLW fragments into direct contact with the bedded salt environment. A second set utilizes full-size canisters of waste emplaced in the salt in the same manner as planned for a future HLW repository. The bare waste experiments will study in an accelerated manner waste-salt bed-brine interactions including matrix integrity/degradation, brine leaching, system chemistry, and potential radionuclide migration through the salt bed. Utilization of full-size canisters of HLW in situ permits us to demonstrate operational effectiveness and safety. Experiments will evaluate corrosion and compatibility interactions between the waste matrix, canister and overpack materials, getter materials, stored energy, waste buoyancy, etc. Using full size canisters also allows us to demonstrate engineered retrievability of wastes, if necessary, at the end of experimentation

  3. Effect of composition of simulated intestinal media on the solubility of poorly soluble compounds investigated by design of experiments

    DEFF Research Database (Denmark)

    Madsen, Cecilie Maria; Feng, Kung-I; Leithead, Andrew

    2018-01-01

    The composition of the human intestinal fluids varies both intra- and inter-individually. This will influence the solubility of orally administered drug compounds, and hence, the absorption and efficacy of compounds displaying solubility limited absorption. The purpose of this study was to assess...... studies feasible compared to single SIF solubility studies. Applying this DoE approach will lead to a better understanding of the impact of intestinal fluid composition on the solubility of a given drug compound....

  4. Connecting Solubility, Equilibrium, and Periodicity in a Green, Inquiry Experiment for the General Chemistry Laboratory

    Science.gov (United States)

    Cacciatore, Kristen L.; Amado, Jose; Evans, Jason J.; Sevian, Hannah

    2008-01-01

    We present a novel first-year chemistry laboratory experiment that connects solubility, equilibrium, and chemical periodicity concepts. It employs a unique format that asks students to replicate experiments described in different sample lab reports, each lacking some essential information, rather than follow a scripted procedure. This structure is…

  5. Korean Reference HLW Disposal System

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Heui Joo; Lee, J. Y.; Kim, S. S. (and others)

    2008-03-15

    This report outlines the results related to the development of Korean Reference Disposal System for High-level radioactive wastes. The research has been supported around for 10 years through a long-term research plan by MOST. The reference disposal method was selected via the first stage of the research during which the technical guidelines for the geological disposal of HLW were determined too. At the second stage of the research, the conceptual design of the reference disposal system was made. For this purpose the characteristics of the reference spent fuels from PWR and CANDU reactors were specified, and the material and specifications of the canisters were determined in term of structural analysis and manufacturing capability in Korea. Also, the mechanical and chemical characteristics of the domestic Ca-bentonite were analyzed in order to supply the basic design parameters of the buffer. Based on these parameters the thermal and mechanical analysis of the near-field was carried out. Thermal-Hydraulic-Mechanical behavior of the disposal system was analyzed. The reference disposal system was proposed through the second year research. At the final third stage of the research, the Korean Reference disposal System including the engineered barrier, surface facilities, and underground facilities was proposed through the performance analysis of the disposal system.

  6. A dose of HLW reality

    International Nuclear Information System (INIS)

    Payne, J.

    1993-01-01

    What many people were sure they knew, and some others were fairly confident they knew, was acknowledged by the US Department of Energy in December: A monitored retrievable storage (MRS) facility will not be ready to accept spent fuel by January 31, 1998. A dose of reality has thus been added to the US high-level radioactive waste scene. Perhaps as important as the new reality is the practical, businesslike nature of the DOE's plan. The Department's proposal has the quality of a plan aimed at genuinely solving a problem rather just going through the motions. (In contrast, some readers are familiar with New York State's procedures for siting and licensing a low-level waste facility - procedures so labyrinthine that they are much more likely to protect political careers in that state than they are to achieve an LLW site). The DOE has received a lot of criticism - some justified, some not - about its handling of the HLW program. In this instance, it is proposing what many in the industry might have recommended: Make available storage capacity for spent nuclear fuel at existing federal government sites

  7. SOURCE TERMS FOR HLW GLASS CANISTERS

    International Nuclear Information System (INIS)

    J.S. Tang

    2000-01-01

    This calculation is prepared by the Monitored Geologic Repository (MGR) Waste Package Design Section. The objective of this calculation is to determine the source terms that include radionuclide inventory, decay heat, and radiation sources due to gamma rays and neutrons for the high-level radioactive waste (HLW) from the, West Valley Demonstration Project (WVDP), Savannah River Site (SRS), Hanford Site (HS), and Idaho National Engineering and Environmental Laboratory (INEEL). This calculation also determines the source terms of the canister containing the SRS HLW glass and immobilized plutonium. The scope of this calculation is limited to source terms for a time period out to one million years. The results of this calculation may be used to carry out performance assessment of the potential repository and to evaluate radiation environments surrounding the waste packages (WPs). This calculation was performed in accordance with the Development Plan ''Source Terms for HLW Glass Canisters'' (Ref. 7.24)

  8. Historical fuel reprocessing and HLW management in Idaho

    International Nuclear Information System (INIS)

    Knecht, D.A.; Staiger, M.D.; Christian, J.D.

    1997-01-01

    This article review some of the key decision points in the historical development of spent fuel reprocessing and waste management practices at the Idaho Chemical Processing Plant that have helped ICPP to successfully accomplish its mission safely and with minimal impact on the environment. Topics include ICPP reprocessing development; batch aluminum-uranium dissolution; continuous aluminum uranium dissolution; batch zirconium dissolution; batch stainless steel dissolution; semicontinuous zirconium dissolution with soluble poison; electrolytic dissolution of stainless steel-clad fuel; graphite-based rover fuel processing; fluorinel fuel processing; ICPP waste management consideration and design decisions; calcination technology development; ICPP calcination demonstration and hot operations; NWCF design, construction, and operation; HLW immobilization technology development. 80 refs., 4 figs

  9. Long term corrosion behavior of the WAK-HLW glass in salt solutions

    International Nuclear Information System (INIS)

    Luckscheiter, B.; Nesovic, M.

    1998-01-01

    The corrosion behavior of the HLW glass GP WAK1 containing simulated HLW oxides from the WAK reprocessing plant in Karlsruhe is investigated in long-term corrosion experiments at high S/V ratios in two reference brines at 110 and 190 C. In case of the MgCl 2 -rich solution the leachate becomes increasingly acid with reaction time up to a final pH of about 3.5 at 190 C. In the NaCl-rich solution the pH rises to about 8.5 after one year of reaction. The release of soluble elements in MgCl 2 solution, under Si-saturated conditions, is proportional to the surface area of the sample and the release increases at 190 C according to a t 1/2 rate law. This time dependence may be an indication of diffusion controlled matrix dissolution. However, at 110 C the release of the mobile elements cannot be described by a t 1/2 rate law as the time exponents are much lower than 0.5. This difference in corrosion behavior may be explained by the higher pH of about 5 at 110 C. In case of NaCl solution under alkaline conditions, the release of soluble elements is not proportional to the surface area of the sample and it increases with time exponents much lower than 0.5. After one year of reaction at 190 C a sharp increase of the release values of some elements was observed. This increase might be explained by the high pH of the solution attained after one year. The corrosion mechanism in NaCl solution, as well as in MgCl 2 solution at 110 C, has not yet been explained. By corrosion experiments in water at constant pH values between 2 and 10, it could be shown that the time exponents of the release of Li and B decrease with increasing pH of the solution. This result can explain qualitatively the differences found in the corrosion behavior of the glass under the various conditions

  10. Key Factors to Determine the Borehole Spacing in a Deep Borehole Disposal for HLW

    International Nuclear Information System (INIS)

    Lee, Jongyoul; Choi, Heuijoo; Lee, Minsoo; Kim, Geonyoung; Kim, Kyeongsoo

    2015-01-01

    Deep fluids also resist vertical movement because they are density stratified and reducing conditions will sharply limit solubility of most dose critical radionuclides at the depth. Finally, high ionic strengths of deep fluids will prevent colloidal transport. Therefore, as an alternative disposal concept, i.e., deep borehole disposal technology is under consideration in number of countries in terms of its outstanding safety and cost effectiveness. In this paper, the general concept for deep borehole disposal of spent fuels or high level radioactive wastes which has been developed by some countries according to the rapid advance in the development of drilling technology, as an alternative method to the deep geological disposal method, was reviewed. After then an analysis on key factors for the distance between boreholes for the disposal of HLW was carried out. In this paper, the general concept for deep borehole disposal of spent fuels or HLW wastes, as an alternative method to the deep geological disposal method, were reviewed. After then an analysis on key factors for the determining the distance between boreholes for the disposal of HLW was carried out. These results can be used for the development of the HLW deep borehole disposal system

  11. Key Factors to Determine the Borehole Spacing in a Deep Borehole Disposal for HLW

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jongyoul; Choi, Heuijoo; Lee, Minsoo; Kim, Geonyoung; Kim, Kyeongsoo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    Deep fluids also resist vertical movement because they are density stratified and reducing conditions will sharply limit solubility of most dose critical radionuclides at the depth. Finally, high ionic strengths of deep fluids will prevent colloidal transport. Therefore, as an alternative disposal concept, i.e., deep borehole disposal technology is under consideration in number of countries in terms of its outstanding safety and cost effectiveness. In this paper, the general concept for deep borehole disposal of spent fuels or high level radioactive wastes which has been developed by some countries according to the rapid advance in the development of drilling technology, as an alternative method to the deep geological disposal method, was reviewed. After then an analysis on key factors for the distance between boreholes for the disposal of HLW was carried out. In this paper, the general concept for deep borehole disposal of spent fuels or HLW wastes, as an alternative method to the deep geological disposal method, were reviewed. After then an analysis on key factors for the determining the distance between boreholes for the disposal of HLW was carried out. These results can be used for the development of the HLW deep borehole disposal system.

  12. Processes for consensus building and role sharing. Lessons learned from HLW policies in European countries

    International Nuclear Information System (INIS)

    Nagano, Koji

    2003-01-01

    This report attempts to obtain lessons in implementation of HLW management policies for Japan by reviewing past experiences and present status of policy formulation and implementation as well as reflection of public opinions and consensus building of selected European countries, such as Finland, Sweden and others. After examining the situations of those countries, the author derives four key aspects that need to be addressed; separation of nuclear energy policies and HLW policies, fundamental support shared among national public, sense of controllability, and proper scheme of responsibility sharing. (author)

  13. HIGH ALUMINUM HLW GLASSES FOR HANFORD'S WTP

    International Nuclear Information System (INIS)

    Kruger, A.A.; Joseph, I.; Bowman, B.W.; Gan, H.; Kot, W.; Matlack, K.S.; Pegg, I.L

    2009-01-01

    The world's largest radioactive waste vitrification facility is now under construction at the United State Department of Energy's (DOE's) Hanford site. The Hanford Tank Waste Treatment and Immobilization Plant (WTP) is designed to treat nearly 53 million gallons of mixed hazardous and radioactive waste now residing in 177 underground storage tanks. This multi-decade processing campaign will be one of the most complex ever undertaken because of the wide chemical and physical variability of the waste compositions generated during the cold war era that are stored at Hanford. The DOE Office of River Protection (ORP) has initiated a program to improve the long-term operating efficiency of the WTP vitrification plants with the objective of reducing the overall cost of tank waste treatment and disposal and shortening the duration of plant operations. Due to the size, complexity and duration of the WTP mission, the lifecycle operating and waste disposal costs are substantial. As a result, gains in High Level Waste (HLW) and Low Activity Waste (LAW) waste loadings, as well as increases in glass production rate, which can reduce mission duration and glass volumes for disposal, can yield substantial overall cost savings. EnergySolutions and its long-term research partner, the Vitreous State Laboratory (VSL) of the Catholic University of America, have been involved in a multi-year ORP program directed at optimizing various aspects of the HLW and LAW vitrification flow sheets. A number of Hanford HLW streams contain high concentrations of aluminum, which is challenging with respect to both waste loading and processing rate. Therefore, a key focus area of the ORP vitrification process optimization program at EnergySolutions and VSL has been development of HLW glass compositions that can accommodate high Al 2 O 3 concentrations while maintaining high processing rates in the Joule Heated Ceramic Melters (JHCMs) used for waste vitrification at the WTP. This paper, reviews the

  14. Thermodynamic properties of a smectite and an illite: comparison between solubility experiments and calorimetric results

    International Nuclear Information System (INIS)

    Gailhanou, H.; Gaboreau, S.; Gaucher, E.C.; Blanc, P.; Rogez, J.; Olives, J.; Amouric, M.; Van Miltenburg, J.C.; Michau, N.; Giffaut, E.

    2010-01-01

    Document available in extended abstract form only. In the context of nuclear waste repositories in argillaceous formations, it is necessary to assess the geochemical behaviour of natural and engineered clay barriers, namely by the stability of clay minerals over long periods of time. However, thermodynamic data of clay minerals, which are required for geochemical modelling, are still poorly known. The present study aims to improve our comprehension of clay mineral stability. The thermodynamic properties of smectite MX-80 and illite IMt-2 (Silver Hill, Montana) have already been determined using calorimetric methods between 0 K and 520 K, under both dried and hydrated states in the case of smectite. In parallel, solubility experiments were carried out in order to determine the solubility products of the clay minerals. Such experiments require some particular precautions as published by May et al. (1986) and Aja and Rosenberg (1992). According to these authors, it is namely important to reach the equilibrium from both under and over saturation and to characterize from a mineralogical point of view the end products. Taking advantage of previous calorimetric measurements, we propose to compare the solubility products obtained here from dissolution experiments with respect to calorimetry results, in order to assess the equilibrium achievement for the solubility experiments. The study is integrated in the Thermochimie project, which aims to provide a consistent thermodynamic database (Thermochimie) for modelling purposes. This work dealing with a smectite and an illite is being completed by the study of a set of typical clays, selected to be well-representative of the clay group. Enthalpies of formation of the minerals were determined by isothermal dissolution calorimetry at 25 deg. C, using a HF-HNO 3 solution. They were obtained by measuring the enthalpies of dissolution of (i) the sample (clay mineral + impurities) and (ii) the oxide or hydroxide constituent mixture

  15. Demonstration of pyrometallurgical processing for metal fuel and HLW

    International Nuclear Information System (INIS)

    Tadafumi, Koyama; Kensuke, Kinoshita; Takatoshi, Hizikata; Tadashi, Inoue; Ougier, M.; Rikard, Malmbeck; Glatz, J.P.; Lothar, Koch

    2001-01-01

    CRIEPI and JRC-ITU have started a joint study on pyrometallurgical processing to demonstrate the capability of this type of process for separating actinide elements from spent fuel and HLW. The equipment dedicated for this experiments has been developed and installed in JRC-ITU. The stainless steel box equipped with tele-manipulators is operated under pure Ar atmosphere, and prepared for later installation in a hot cell. Experiments on pyro-processing of un-irradiated U-Pu-Zr metal alloy fuel by molten salt electrorefining has been carried out. Recovery of U and Pu from this type alloy fuel was first demonstrated with using solid iron cathode and liquid Cd cathode, respectively. (author)

  16. Safety assessment of HLW geological disposal system

    International Nuclear Information System (INIS)

    Naito, Morimasa

    2006-01-01

    In accordance with the Japanese nuclear program, the liquid waste with a high level of radioactivity arising from reprocessing is solidified in a stable glass matrix (vitrification) in stainless steel fabrication containers. The vitrified waste is referred to as high-level radioactive waste (HLW), and is characterized by very high initial radioactivity which, even though it decreases with time, presents a potential long-term risk. It is therefore necessary to thoroughly manage HLW from human and his environment. After vitrification, HLW is stored for a period of 30 to 50 years to allow cooling, and finally disposed of in a stable geological environment at depths greater than 300 m below surface. The deep underground environment, in general, is considered to be stable over geological timescales compared with surface environment. By selecting an appropriate disposal site, therefore, it is considered to be feasible to isolate the waste in the repository from man and his environment until such time as radioactivity levels have decayed to insignificance. The concept of geological disposal in Japan is similar to that in other countries, being based on a multibarrier system which combines the natural geological environment with engineered barriers. It should be noted that geological disposal concept is based on a passive safety system that does not require any institutional control for assuring long term environmental safety. To demonstrate feasibility of safe HLW repository concept in Japan, following technical steps are essential. Selection of a geological environment which is sufficiently stable for disposal (site selection). Design and installation of the engineered barrier system in a stable geological environment (engineering measures). Confirmation of the safety of the constructed geological disposal system (safety assessment). For site selection, particular consideration is given to the long-term stability of the geological environment taking into account the fact

  17. Source term measurements on vitrified HLW

    International Nuclear Information System (INIS)

    Hough, A.; Marples, J.A.C.

    1988-01-01

    The equilibrium concentrations of Tc-99, Np-237, Pu-239/240 and Am-241 have been measured in the presence of materials likely to be present in a vitrified HLW repository: glass, iron, backfill and rock. Results were measured under both oxidising and reducing conditions and at pH values set by the backfill bentonite and cement. Under reducing conditions and with cementitious backfills, the equilibrium concentrations ranged from three to 30 times allowed drinking water levels for the four isotopes. (author)

  18. Strategic management of HLW repository projects

    International Nuclear Information System (INIS)

    Bartlett, J.W.

    1984-01-01

    This paper suggests an approach to strategic management of HLW repository projects based on the premise that a primary objective of project activities is resolution of issues. The approach would be implemented by establishing an issues management function with responsibility to define the issues agenda, develop and apply the tools for assessing progress toward issue resolution, and develop the issue resolution criteria. A principal merit of the approach is that it provides a defensible rationale for project plans and activities. It also helps avoid unnecessary costs and schedule delays, and it helps assure coordination between project functions that share responsibilities for issue resolution

  19. Plutonium solubilities

    International Nuclear Information System (INIS)

    Puigdomnech, I.; Bruno, J.

    1991-02-01

    Thermochemical data has been selected for plutonium oxide, hydroxide, carbonate and phosphate equilibria. Equilibrium constants have been evaluated in the temperature range 0 to 300 degrees C at a pressure of 1 bar to T≤100 degrees C and at the steam saturated pressure at higher temperatures. Measured solubilities of plutonium that are reported in the literature for laboratory experiments have been collected. Solubility data on oxides, hydroxides, carbonates and phosphates have been selected. No solubility data were found at temperatures higher than 60 degrees C. The literature solubility data have been compared with plutonium solubilities calculated with the EQ3/6 geochemical modelling programs, using the selected thermodynamic data for plutonium. (authors)

  20. HLW Long-term Management Technology Development

    International Nuclear Information System (INIS)

    Choi, Jong Won; Kang, C. H.; Ko, Y. K.

    2010-02-01

    Permanent disposal of spent nuclear fuels from the power generation is considered to be the unique method for the conservation of human being and nature in the present and future. In spite of spent nuclear fuels produced from power generation, based on the recent trends on the gap between supply and demand of energy, the advance on energy price and reduction of carbon dioxide, nuclear energy is expected to play a role continuously in Korea. It means that a new concept of nuclear fuel cycle is needed to solve problems on spent nuclear fuels. The concept of the advanced nuclear fuel cycle including PYRO processing and SFR was presented at the 255th meeting of the Atomic Energy Commission. According to the concept of the advanced nuclear fuel cycle, actinides and long-term fissile nuclides may go out of existence in SFR. And then it is possible to dispose of short term decay wastes without a great risk bearing. Many efforts had been made to develop the KRS for the direct disposal of spent nuclear fuels in the representative geology of Korea. But in the case of the adoption of Advanced nuclear fuel cycle, the disposal of PYRO wastes should be considered. For this, we carried out the Safety Analysis on HLW Disposal Project with 5 sub-projects such as Development of HLW Disposal System, Radwaste Disposal Safety Analysis, Feasibility study on the deep repository condition, A study on the Nuclide Migration and Retardation Using Natural Barrier, and In-situ Study on the Performance of Engineered Barriers

  1. HLW disposal in Germany - R and D achievements and outlook

    International Nuclear Information System (INIS)

    Steininger, W.

    2006-01-01

    The paper gives a brief overview of the status of R and D on HLW disposal. Shortly addressed is the current nuclear policy. After describing the responsibilities regarding R and D for disposing of heat-generating high-level (HLW) waste (vitrified waste and spent fuel), selected projects are mentioned to illustrate the state of knowledge in disposing of waste in rock salt. Participation in international projects and programs is described to illustrate the value for the German concepts and ideas for HLW disposal in different rock types. Finally, a condensed outlook on future activities is given. (author)

  2. Neutron radiography experiments for verification of soluble boron mixing and transport modeling under natural circulation conditions

    International Nuclear Information System (INIS)

    Morlang, M.M.; Feltus, M.A.

    1996-01-01

    The use of neutron radiography for visualization of fluid flow through flow visualization modules has been very successful. Current experiments at the Penn State Breazeale Reactor serve to verify the mixing and transport of soluble boron under natural flow conditions as would be experienced in a pressurized water reactor. Different flow geometries have been modeled including holes, slots, and baffles. Flow modules are constructed of aluminum box material 1 1/2 inches by 4 inches in varying lengths. An experimental flow system was built which pumps fluid to a head tank and natural circulation flow occurs from the head tank through the flow visualization module to be radio-graphed. The entire flow system is mounted on a portable assembly to allow placement of the flow visualization module in front of the neutron beam port. A neutron-transparent fluor-inert fluid is used to simulate water at different densities. Boron is modeled by gadolinium oxide powder as a tracer element, which is placed in a mixing assembly and injected into the system a remotely operated electric valve, once the reactor is at power. The entire sequence is recorded on real-time video. Still photographs are made frame-by-frame from the video tape. Computers are used to digitally enhance the video and still photographs. The data obtained from the enhancement will be used for verification of simple geometry predictions using the TRAC and RELAP thermal-hydraulic codes. A detailed model of a reactor vessel inlet plenum, downcomer region, flow distribution area and core inlet is being constructed to model the APGOO plenum. Successive radiography experiments of each section of the model under identical conditions will provide a complete vessel / core model for comparison with the thermal-hydraulic codes

  3. Neutron radiography experiments for verification of soluble boron mixing and transport modeling under natural circulation conditions

    International Nuclear Information System (INIS)

    Feltus, M.A.; Morlang, G.M.

    1996-01-01

    The use of neutron radiography for visualization of fluid flow through flow visualization modules has been very successful. Current experiments at the Penn State Breazeale Reactor serve to verify the mixing and transport of soluble boron under natural flow conditions as would be experienced in a pressurized water reactor. Different flow geometries have been modeled including holes, slots, and baffles. Flow modules are constructed of aluminum box material 1 1/2 inches by 4 inches in varying lengths. An experimental flow system was built which pumps fluid to a head tank and natural circulation flow occurs from the head tank through the flow visualization module to be radiographed. The entire flow system is mounted on a portable assembly to allow placement of the flow visualization module in front of the neutron beam port. A neutron-transparent fluorinert fluid is used to simulate water at different densities. Boron is modeled by gadolinium oxide powder as a tracer element, which is placed in a mixing assembly and injected into the system by remote operated electric valve, once the reactor is at power. The entire sequence is recorded on real-time video. Still photographs are made frame-by-frame from the video tape. Computers are used to digitally enhance the video and still photographs. The data obtained from the enhancement will be used for verification of simple geometry predictions using the TRAC and RELAP thermal-hydraulic codes. A detailed model of a reactor vessel inlet plenum, downcomer region, flow distribution area and core inlet is being constructed to model the AP600 plenum. Successive radiography experiments of each section of the model under identical conditions will provide a complete vessel/core model for comparison with the thermal-hydraulic codes

  4. Diffusion and solubility coefficients determined by permeation and immersion experiments for organic solvents in HDPE geomembrane.

    Science.gov (United States)

    Chao, Keh-Ping; Wang, Ping; Wang, Ya-Ting

    2007-04-02

    The chemical resistance of eight organic solvents in high density polyethylene (HDPE) geomembrane has been investigated using the ASTM F739 permeation method and the immersion test at different temperatures. The diffusion of the experimental organic solvents in HDPE geomembrane was non-Fickian kinetic, and the solubility coefficients can be consistent with the solubility parameter theory. The diffusion coefficients and solubility coefficients determined by the ASTM F739 method were significantly correlated to the immersion tests (pHDPE as barriers in the field.

  5. Compas project stress analysis of HLW containers: behaviour under realistic disposal conditions

    International Nuclear Information System (INIS)

    Ove Arup and Partners, London

    1990-01-01

    The Compas project is concerned with the structural performance of metal overpacks which may be used to encapsulate vitrified high-level waste (HLW) forms before disposal in deep geological repositories. In this final stage of the project, analysis of an HLW overpack of realistic design is performed to predict its behaviour when subjected to likely repository loads. This analysis work is undertaken with the benefit of experience gained in previous phases of the project in which the ability to accurately predict overpack behaviour, when subjected to a uniform external pressure, was demonstrated. Burial in clay, granite and salt environments has been considered and two distinct loading arrangements identified, in an attempt to represent the worst conditions that could be imposed by such media. The analysis successfully demonstrates the ability of the containers to withstand extreme, yet credible, repository loads

  6. Industrial scale-plant for HLW partitioning in Russia

    International Nuclear Information System (INIS)

    Dzekun, E.G.; Glagolenko, Y.V.; Drojko, E.G.; Kurochkin, A.I.

    1996-01-01

    Radiochemical plant of PA > at Ozersk, which was come on line in December 1948 originally for weapon plutonium production and reoriented on the reprocessing of spent fuel, till now keeps on storage HLW of the military program. Application of the vitrification method since 1986 has not essentially reduced HLW volumes. So, as of September 1, 1995 vitrification installations had been processed 9590 m 3 HLW and 235 MCi of radionuclides was included in glass. However only 1100 m 3 and 20.5 MCi is part of waste of the military program. The reason is the fact, that the technology and equipment of vitrification were developed for current waste of Purex-process, for which low contents of corrosion-dangerous impurity to materials of vitrification installation is characteristic of. With reference to HLW, which are growing at PA > in the course of weapon plutonium production, the program of Science-Research Works includes the following main directions of work. Development of technology and equipment of installations for immobilising HLW with high contents of impurity into a solid form at induction melter. Application of High-temperature Adsorption Method for sorption of radionuclides from HLW on silica gel. Application of Partitioning Method of radionuclides from HLW, based on extraction cesium and strontium into cobalt dicarbollyde or crown-ethers, but also on recovery of cesium radionuclides by sorption on inorganic sorbents. In this paper the results of work on creation of first industrial scale-plant for partitioning HLW by the extraction and sorption methods are reported

  7. SNF/HLW Transfer System Description Document

    International Nuclear Information System (INIS)

    W. Holt

    2005-01-01

    The purpose of this system description document (SDD) is to establish requirements that drive the design of the spent nuclear fuel (SNF)/high-level radioactive waste (HLW) transfer system and associated bases, which will allow the design effort to proceed to license application. This SDD will be revised at strategic points as the design matures. This SDD identifies the requirements and describes the system design, as it currently exists, with emphasis on attributes of the design provided to meet the requirements. This SDD is an engineering tool for design control. Accordingly, the primary audience and users are design engineers. This SDD is part of an iterative design process. It leads the design process with regard to the flowdown of upper tier requirements onto the system. Knowledge of these requirements is essential in performing the design process. The SDD follows the design with regard to the description of the system. The description provided in this SDD reflects the current results of the design process

  8. Stress analysis of HLW containers. Compas project

    International Nuclear Information System (INIS)

    1989-01-01

    This document reports the work carried out for the Compas project which looked at the performance of various computer codes in a selected benchmark exercise. This exercise consisted of several analyses on simplified models which have features typical of HLW containers. These analyses comprise two groups; one related to thick walled, stressed shell overpacks, the other related to thin walled, supported shell overpacks with a lead filler. The first set of analyses looked at an elastic-plastic behaviour and large deformation of a cylinder representative of the main body of thick walled containers). The second set looked at creep behaviour of the lead filler, and the shape the base of thin walled containers will take up, after hundreds of years in the repository. On the thick walled analyses with the cylinder subject to an external pressure all the codes gave consistent results in the elastic region and there is good agreement in the yield pressures. Once in the plastic region there is more divergence in the results although a consistent trend is predicted. One of the analyses predicted a non-axisymmetric mode of deformation as would be expected in reality. Fewer results were received for the creep analysis, however the transient creep results showed consistency, and were bounded by the final-state results

  9. MELT RATE ENHANCEMENT FOR HIGH ALUMINUM HLW (HIGH LEVEL WASTE) GLASS FORMULATION FINAL REPORT 08R1360-1

    Energy Technology Data Exchange (ETDEWEB)

    KRUGER AA; MATLACK KS; KOT W; PEGG IL; JOSEPH I; BARDAKCI T; GAN H; GONG W; CHAUDHURI M

    2010-01-04

    . The WTP HLW melter design, unlike earlier DOE melter designs, incorporates an active glass bubbler system. The bubblers create active glass pool convection and thereby improve heat transfer and glass melting rate. The WTP HLW melter has a glass surface area of 3.75 m{sup 2} and depth of {approx}1.1 m. The two melters in the HLW facility together are designed to produce up to 7.5 MT of glass per day at 100% availability. Further increases in HLW waste processing rates can potentially be achieved by increasing the melter operating temperature above 1150 C and by increasing the waste loading in the glass product. Increasing the waste loading also has the added benefit of decreasing the number of canisters for storage. The current estimates and glass formulation efforts have been conservative in terms of achievable waste loadings. These formulations have been specified to ensure that the glasses are homogenous, contain essentially no crystalline phases, are processable in joule-heated, ceramic-lined melters and meet WTP Contract terms. The WTP's overall mission will require the immobilization of tank waste compositions that are dominated by mixtures of aluminum (Al), chromium (Cr), bismuth (Bi), iron (Fe), phosphorous (P), zirconium (Zr), and sulfur (S) compounds as waste-limiting components. Glass compositions for these waste mixtures have been developed based upon previous experience and current glass property models. Recently, DOE has initiated a testing program to develop and characterize HLW glasses with higher waste loadings. Results of this work have demonstrated the feasibility of increases in wasteloading from about 25 wt% to 33-50 wt% (based on oxide loading) in the glass depending on the waste stream. It is expected that these higher waste loading glasses will reduce the HLW canister production requirement by about 25% or more.

  10. MELT RATE ENHANCEMENT FOR HIGH ALUMINUM HLW (HIGH LEVEL WASTE) GLASS FORMULATION. FINAL REPORT 08R1360-1

    International Nuclear Information System (INIS)

    Kruger, A.A.; Matlack, K.S.; Kot, W.; Pegg, I.L.; Joseph, I.; Bardakci, T.; Gan, H.; Gong, W.; Chaudhuri, M.

    2010-01-01

    . The WTP HLW melter design, unlike earlier DOE melter designs, incorporates an active glass bubbler system. The bubblers create active glass pool convection and thereby improve heat transfer and glass melting rate. The WTP HLW melter has a glass surface area of 3.75 m 2 and depth of ∼1.1 m. The two melters in the HLW facility together are designed to produce up to 7.5 MT of glass per day at 100% availability. Further increases in HLW waste processing rates can potentially be achieved by increasing the melter operating temperature above 1150 C and by increasing the waste loading in the glass product. Increasing the waste loading also has the added benefit of decreasing the number of canisters for storage. The current estimates and glass formulation efforts have been conservative in terms of achievable waste loadings. These formulations have been specified to ensure that the glasses are homogenous, contain essentially no crystalline phases, are processable in joule-heated, ceramic-lined melters and meet WTP Contract terms. The WTP's overall mission will require the immobilization of tank waste compositions that are dominated by mixtures of aluminum (Al), chromium (Cr), bismuth (Bi), iron (Fe), phosphorous (P), zirconium (Zr), and sulfur (S) compounds as waste-limiting components. Glass compositions for these waste mixtures have been developed based upon previous experience and current glass property models. Recently, DOE has initiated a testing program to develop and characterize HLW glasses with higher waste loadings. Results of this work have demonstrated the feasibility of increases in wasteloading from about 25 wt% to 33-50 wt% (based on oxide loading) in the glass depending on the waste stream. It is expected that these higher waste loading glasses will reduce the HLW canister production requirement by about 25% or more.

  11. HLW Flexible jumper materials compatibility evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Skidmore, T. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-11-13

    H-Tank Farm Engineering tasked SRNL/Materials Science & Technology (MS&T) to evaluate the compatibility of Goodyear Viper® chemical transfer hose with HLW solutions. The hose is proposed as a flexible Safety Class jumper for up to six months service. SRNL/MS&T performed various tests to evaluate the effects of radiation, high pH chemistry and elevated temperature on the hose, particularly the inner liner. Test results suggest an upper dose limit of 50 Mrad for the hose. Room temperature burst pressure values at 50 Mrad are estimated at 600- 800 psi, providing a safety factor of 4.0-5.3X over the anticipated operating pressure of 150 psi and a safety factor of 3.0-4.0X over the working pressure of the hose (200 psi), independent of temperature effects. Radiation effects are minimal at doses less than 10 Mrad. Doses greater than 50 Mrad may be allowed, depending on operating conditions and required safety factors, but cannot be recommended at this time. At 250 Mrad, burst pressure values are reduced to the hose working pressure. At 300 Mrad, burst pressures are below 150 psi. At a bounding continuous dose rate of 57,870 rad/hr, the 50 Mrad dose limit is reached within 1.2 months. Actual dose rates may be lower, particularly during non-transfer periods. Refined dose calculations are therefore recommended to justify longer service. This report details the tests performed and interpretation of the results. Recommendations for shelf-life/storage, component quality verification, and post-service examination are provided.

  12. Compas project stress analysis of HLW containers intermediate testwork

    International Nuclear Information System (INIS)

    Ove Arup and Partners London

    1990-01-01

    The Compas project is concerned with the structural performance of metal overpacks which may be used to encapsulate vitrified high-level waste forms before disposal in deep geological repositories. This document describes the series of experiments and associated calculations performed in the Intermediate testwork phase of this project. Seven mild steel, one-third scale simplified models of HLW containers were manufactured in a variety of configurations of geometry and weld type. The effects of reducing the wall thickness, corroding the external surface of the container, and using different welding methods were all investigated. The containers were tested under the action of a uniform external pressure up to their respective failure points. All containers failed by buckling at pressures of between 42 and 87 MPa dependent upon the particular geometric and weld configuration. The outer surface of each container was comprehensively strain-gauged in order to provide strain histories at positions of interest. The Compas project partners, from five different European countries, independently modelled the behaviour of three of the five different containers. Test results and computer predictions were compared and an assessment of the overall performance of the codes demonstrated good agreement in the initial loading of each container. However once stresses exceeded the material yield point there was a considerable spread in the predicted container behaviour

  13. Database and Interim Glass Property Models for Hanford HLW Glasses

    International Nuclear Information System (INIS)

    Hrma, Pavel R; Piepel, Gregory F; Vienna, John D; Cooley, Scott K; Kim, Dong-Sang; Russell, Renee L

    2001-01-01

    The purpose of this report is to provide a methodology for an increase in the efficiency and a decrease in the cost of vitrifying high-level waste (HLW) by optimizing HLW glass formulation. This methodology consists in collecting and generating a database of glass properties that determine HLW glass processability and acceptability and relating these properties to glass composition. The report explains how the property-composition models are developed, fitted to data, used for glass formulation optimization, and continuously updated in response to changes in HLW composition estimates and changes in glass processing technology. Further, the report reviews the glass property-composition literature data and presents their preliminary critical evaluation and screening. Finally the report provides interim property-composition models for melt viscosity, for liquidus temperature (with spinel and zircon primary crystalline phases), and for the product consistency test normalized releases of B, Na, and Li. Models were fitted to a subset of the screened database deemed most relevant for the current HLW composition region

  14. Solubility of grape seed oil in supercritical CO2: Experiments and modeling

    International Nuclear Information System (INIS)

    Duba, Kurabachew Simon; Fiori, Luca

    2016-01-01

    Highlights: • Solubility of grape seed oil in SC-CO 2 for P: 20–50 MPa and T: 313–343 K. • Experimental procedure: dynamic method and oil dispersed on the surface of glass beads. • Eight density-based models and a thermodynamic model to fit the experimental data. • All the models predict the solubility of grape seed oil in SC-CO 2 to a reasonable degree. • Models by Chrastil, del Valle and Aguilera, Kumar and Johnston, and the thermodynamic model are preferable. - Abstract: The solubility of grape (Vitis vinifera L.) seed oil in supercritical CO 2 was measured in the temperature range 313–343 K and pressure range 20–50 MPa using the dynamic technique. Several data and global trends were reported. The results show that, at constant temperature, the solubility increases with the increase in pressure, while the effect of the temperature is different for low and high pressure. The experimental data were modeled by eight density-based models and a thermodynamic model based on the Peng-Robinson equation of state. By best fitting procedures, the “free parameters” of the various models were calculated: in general, all the tested models have proved to be able to predict the solubility of grape seed oil in supercritical CO 2 . Differences in model capabilities have been discussed based on the main characteristics of the various models, evidencing their distinct and common features. The predictive capability of the thermodynamic model was comparable to that of the density-based models.

  15. R and D on HLW Partitioning in Russia

    International Nuclear Information System (INIS)

    Khaperskaya, A.; Babain, V.; Alyapyshev, M.

    2015-01-01

    Results of more than thirty years investigations on high level radioactive waste (HLW) partitioning in Russia are described. The objectives of research and development is to assess HLW partitioning technical feasibility and its advantages compared to direct vitrification of long-lived radionuclides. Many technological flowsheets for long-lived nuclides (cesium, strontium and minor actinides) separation were developed and tested with simulated and actual HLW. Different classes of extractants, including carbamoyl-phosphine oxides, dialkyl-phosphoric acids, crown ethers and diamides of heterocyclic acids were studied. Some of these processes were tested at PA 'Mayak' and MCC. Many extraction systems based on chlorinated cobalt dicarbollide (CCD), including UNEX-extractant and its modifications, were also observed. Diamides of diglycolic acid and diamides of heterocyclic acids in polar diluents have shown promising properties for minor actinide-lanthanide extraction and separation. Comparison of different solvents and possible ways of implementing new flowsheets in radiochemical technology are also discussed. (authors)

  16. HLW Canister and Can-In-Canister Drop Calculation

    International Nuclear Information System (INIS)

    H. Marr

    1999-01-01

    The purpose of this calculation is to evaluate the structural response of the standard high-level waste (HLW) canister and the HLW canister containing the cans of immobilized plutonium (''can-in-canister'' throughout this document) to the drop event during the handling operation. The objective of the calculation is to provide the structure parameter information to support the canister design and the waste handling facility design. Finite element solution is performed using the commercially available ANSYS Version (V) 5.4 finite element code. Two-dimensional (2-D) axisymmetric and three-dimensional (3-D) finite element representations for the standard HLW canister and the can-in-canister are developed and analyzed using the dynamic solver

  17. Long-term storage or disposal of HLW-dilemma

    International Nuclear Information System (INIS)

    Ninkovic, M. M.; Raicevic, J.

    1995-01-01

    In this paper, a new concept approach to HLW management founded on deterministic safety philosophy - i.e. long-term storage with final objective of destroying was justified and proposed instead of multi barrier concept with final disposal in extra stable environmental conditions, which are founded on probabilistic safety approach model. As a support to this new concept some methods for destruction of waste which are now accessible, on scientific stage only, as transmutation in fast reactors and accelerators of heavy ions were briefly discussed . It is justified to believe that industrial technology for destruction of HLW would be developed in not so far future. (author).

  18. Cooling and cracking of technical HLW glass products

    International Nuclear Information System (INIS)

    Kienzler, B.

    1989-01-01

    The author discusses various cooling procedures applied to canisters filled with inactive simulated HLW glass and the measured temperature distributions compared with numerically computed data. Stress computations of the cooling process were carried out with a finite element method. Only those volume elements having temperatures below the transformation temperature Tg were assumed to contribute thermoelastically to the developing stresses. Model calculations were extended to include real HLW glass canisters with inherent thermal power. The development of stress as a function of variations of heat flow conditions and of the radioactive decay was studied

  19. Active geothermal systems as natural analogs of HLW repositories

    International Nuclear Information System (INIS)

    Elders, W.A.; Williams, A.E.; Cohen, L.H.

    1988-01-01

    Geologic analogs of long-lived processes in high-level waste (HLW) repositories have been much studied in recent years. However, most of these occurrences either involve natural processes going on today at 25 degree C, or, if they are concerned with behavior at temperatures similar to the peak temperatures anticipated near HLW canisters, have long since ended. This paper points out the usefulness of studying modern geothermal systems as natural analogs, and to illustrate the concept with a dramatic example, the Salton Sea geothermal system (SSGS)

  20. Influence of the reprocessing flow sheet on the HLW solidification technology

    International Nuclear Information System (INIS)

    Baetsle, L.H.

    1981-01-01

    The introduction of Pu recycled LWR and CMFBR fuel will require the addition of a second dissolution step to quantitation recover Pu. If process modifications can be brought to the head-end procedures it is advisable to remove Ru, Te, Mo, Pd by high performance centrifugation and to volatilize soluble RuNO(NO 3 ) 2 by sparing with ozone. This changes improve the liquid extraction efficiency and simplify the off gas treatment during calcination and vitrification of HAWC. The conversion of HAW to HAWC by evaporation is accompagnied by some-volatilization of Ru and Cs. Organic reductants reduce the Ru volatilization. The introduction of salt free reagents during feed adjustment steps will decrease Na content in the HLW. The main impact of the use of salt free reagent will have its bearing on the LAW and ILIW treatment and conditioning. (DG)

  1. Enhanced sludge processing of HLW: Hydrothermal oxidation of chromium, technetium, and complexants by nitrate. 1997 mid-year progress report

    International Nuclear Information System (INIS)

    Buelow, S.

    1997-01-01

    'Treatment of High Level Waste (HLW) is the second most costly problem identified by OEM. In order to minimize costs of disposal, the volume of HLW requiring vitrification and long term storage must be reduced. Methods for efficient separation of chromium from waste sludges, such as the Hanford Tank Wastes (HTW), are key to achieving this goal since the allowed level of chromium in high level glass controls waste loading. At concentrations above 0.5 to 1.0 wt.% chromium prevents proper vitrification of the waste. Chromium in sludges most likely exists as extremely insoluble oxides and minerals, with chromium in the plus III oxidation state [1]. In order to solubilize and separate it from other sludge components, Cr(III) must be oxidized to the more soluble Cr(VI) state. Efficient separation of chromium from HLW could produce an estimated savings of $3.4B[2]. Additionally, the efficient separation of technetium [3], TRU, and other metals may require the reformulation of solids to free trapped species as well as the destruction of organic complexants. New chemical processes are needed to separate chromium and other metals from tank wastes. Ideally they should not utilize additional reagents which would increase waste volume or require subsequent removal. The goal of this project is to apply hydrothermal processing for enhanced chromium separation from HLW sludges. Initially, the authors seek to develop a fundamental understanding of chromium speciation, oxidation/reduction and dissolution kinetics, reaction mechanisms, and transport properties under hydrothermal conditions in both simple and complex salt solutions. The authors also wish to evaluate the potential of hydrothermal processing for enhanced separations of technetium and TRU by examining technetium and TRU speciation at hydrothermal conditions optimal for chromium dissolution.'

  2. CCN activation experiments with adipic acid: effect of particle phase and adipic acid coatings on soluble and insoluble particles

    Directory of Open Access Journals (Sweden)

    S. S. Hings

    2008-07-01

    Full Text Available Slightly soluble atmospherically relevant organic compounds may influence particle CCN activity and therefore cloud formation. Adipic acid is a frequently employed surrogate for such slightly soluble organic materials. The 11 published experimental studies on the CCN activity of adipic acid particles are not consistent with each other nor do they, in most cases, agree with the Köhler theory. The CCN activity of adipic acid aerosol particles was studied over a significantly wider range of conditions than in any previous single study. The work spans the conditions of the previous studies and also provides alternate methods for producing "wet" (deliquesced solution droplets and dry adipic acid particles without the need to produce them by atomization of aqueous solutions. The experiments suggest that the scatter in the previously published CCN measurements is most likely due to the difficulty of producing uncontaminated adipic acid particles by atomization of solutions and possibly also due to uncertainties in the calibration of the instruments. The CCN activation of the small (dm<150 nm initially dry particles is subject to a deliquescence barrier, while for the larger particles the activation follows the Köhler curve. Wet adipic acid particles follow the Köhler curve over the full range of particle diameters studied. In addition, the effect of adipic acid coatings on the CCN activity of both soluble and insoluble particles has also been studied. When a water-soluble core is coated by adipic acid, the CCN-hindering effect of particle phase is eliminated. An adipic acid coating on hydrophobic soot yields a CCN active particle. If the soot particle is relatively small (dcore≤102 nm, the CCN activity of the coated particles approaches the deliquescence line of adipic acid, suggesting that the total size of the particle determines CCN activation and the soot core acts as a scaffold.

  3. DM100 AND DM1200 MELTER TESTING WITH HIGH WASTE LOADING GLASS FORMULATIONS FOR HANFORD HIGH-ALUMINUM HLW STREAMS

    Energy Technology Data Exchange (ETDEWEB)

    KRUGER AA; MATLACK KS; KOT WK; PEGG IL; JOSEPH I

    2009-12-30

    transfer and glass melting rate. The WTP HLW melter has a glass surface area of 3.75 m{sup 2} and depth of {approx}1.1 m. The two melters in the HLW facility together are designed to produce up to 7.5 MT of glass per day at 100% availability. Further increases in HLW waste processing rates can potentially be achieved by increasing the melter operating temperature above 1150 C and by increasing the waste loading in the glass product Increasing the waste loading also has the added benefit of decreasing the number of canisters for storage. The current estimates and glass formulation efforts have been conservative in terms of achievable waste loadings. These formulations have been specified to ensure that the glasses are homogenous, contain essentially no crystalline phases, are processable in joule-heated, ceramic-lined melters and meet WTP contract requirements. The WTP's overall mission will require the immobilization oftank waste compositions that are dominated by mixtures of aluminum (Al), chromium (Cr), bismuth (Bi), iron (Fe), phosphorous (P), zirconium (Zr), and sulfur (S) compounds as waste-limiting components. Glass compositions for these waste mixtures have been developed based upon previous experience and current glass property models. Recently, DOE has initiated a testing program to develop and characterize HLW glasses with higher waste loadings. Results of this work have demonstrated the feasibility of increases in waste-loading from about 25 wt% to 33-50 wt% (based on oxide loading) in the glass depending on the waste stream. It is expected that these higher waste loading glasses will reduce the HLW canister production requirement by about 25% or more.

  4. Development of database and QA systems for post closure performance assessment on a potential HLW repository

    International Nuclear Information System (INIS)

    Hwang, Y. S.; Kim, S. G.; Kang, C. H.

    2002-01-01

    In TSPA of long-term post closure radiological safety on permanent disposal of HLW in Korea, appropriate management of input and output data through QA is necessary. The robust QA system is developed using the T2R3 principles applicable for five major steps in R and D's. The proposed system is implemented in the web-based system so that all participants in TSRA are able to access the system. In addition, the internet based input database for TSPA is developed. Currently data from literature surveys, domestic laboratory and field experiments as well as expert elicitation are applied for TSPA

  5. HLW immobilization in glass: industrial operation and product quality

    International Nuclear Information System (INIS)

    Jacquet-Francillon, N.; Leroy, P.; Runge, S.

    1992-01-01

    This extended summary discusses the immobilization of high level wastes from the viewpoint of the quality of the final product, i.e. the HLW glass. The R and D studies comprise 3 steps: glass formulation, glass characterization and long term behaviour studies

  6. Influence of Glass Property Restrictions on Hanford HLW Glass Volume

    International Nuclear Information System (INIS)

    Kim, Dong-Sang; Vienna, John D.

    2001-01-01

    A systematic evaluation of Hanford High-Level Waste (HLW) loading in alkali-alumino-borosilicate glasses was performed. The waste feed compositions used were obtained from current tank waste composition estimates, Hanford's baseline retrieval sequence, and pretreatment processes. The waste feeds were sorted into groups of like composition by cluster analysis. Glass composition optimization was performed on each cluster to meet property and composition constraints while maximizing waste loading. Glass properties were estimated using property models developed for Hanford HLW glasses. The impacts of many constraints on the volume of HLW glass to be produced at Hanford were evaluated. The liquidus temperature, melting temperature, chromium concentration, formation of multiple phases on cooling, and product consistency test response requirements for the glass were varied one- or many-at-a-time and the resultant glass volume was calculated. This study shows clearly that the allowance of crystalline phases in the glass melter can significantly decrease the volume of HLW glass to be produced at Hanford.

  7. Actinide partitioning from HLW in a continuous DIDPA extraction process by means of centrifugal extractors

    International Nuclear Information System (INIS)

    Morita, Y.; Kubota, M.; Glatz, J.P.; Koch, L.; Pagliosa, G.; Roemer, K.; Nicholl, A.

    1996-01-01

    An experiment on actinide partitioning from real high level waste (HLW) was performed in a continuous process by extraction with diisodecylphosphoric acid (DIDPA) using a battery of 12 centrifugal extractors installed in a hot cell. The HNO 3 concentration of the HLW was adjusted to 0.5 M by dilution. The extraction section had 8 stages, and H 2 O 2 was added to extract Np effectively. After extraction, Am and Cm were back-extracted with 4 M HNO 3 in 4 stages and Np and Pu were stripped with 0.8 M H 2 C 2 O 4 in 8 stages. The actinides, expect Np, were extracted from HLW with a very high yield. Although only 84% of the Np were recovered in the present experiment, the recovery would be improved to 99.7 % by increasing the temperature to 45 degree C, the number of stages from 8 to 16 and the H 2 O 2 concentration from 1 M to 2 M. Long-lived Tc and the main heat and radiation emitters Cs and Sr were not extracted and were thus separated from the actinides with high decontamination factors. About 98% of Am and Cm were recovered from the loaded solvent in the first stripping step with 4M HNO 3 . About 86% of Np and about 92% of Pu were back-extracted with 0.8 M H 2 C 2 O 4 . These incomplete recoveries would be improved by increasing the number of stages and by optimizing the other process parameters. 18 refs., 5 figs., 3 tabs

  8. Putting HLW performance assessment results in perspective

    International Nuclear Information System (INIS)

    Neall, F.; Smith, P.; Sumerling, T.; Umeki, H.

    1995-01-01

    According to performance assessment results for the different disposal concepts investigated, the maximum radiation doses to the population lie well below the limit set in the official Swiss Protection Objective and below the level of present-day natural background radiation. A comparison of different performance assessments has shown that the following key factors determine radionuclide release from a repository: radionuclide inventory, canister material and failure mode, nuclide solubility limits, the permeability of the buffer material, retardation during transport through the near-field, the presence of an excavation disturbed zone in the rock, the distance to the nearest major water-bearing fracture zone, the conceptual model for transport in fractured rock and near-surface dilution and dose factors. (author) 2 figs., 2 tabs

  9. Solubility and speciation of actinides in salt solutions and migration experiments of intermediate level waste in salt formations

    International Nuclear Information System (INIS)

    1986-01-01

    A comprehensive study into the solubility of the actinides americium and plutonium in concentrated salt solutions, the release of radionuclides from various forms of conditioned ILW and the migration behaviour of these nuclides through geological material specific to the Gorleben site in Lower Saxony is described. A detailed investigation into the characterization of four highly concentrated salt solutions in terms of their pH, Eh, inorganic carbon contents and their densities is given and a series of experiments investigating the solubility of standard americium(III) and plutonium(IV) hydroxides in these solutions is described. Transuranic mobility studies for solutions derived from the standard hydroxides through salt and sand have shown the presence of at least two types of species present of widely differing mobility; one migrating with approximately the same velocity as the solvent front and the other strongly retarded. Actinide mobility data are presented and discussed for leachates derived from the simulated ILW in cement and data are also presented for the migration of the fission products in leachates derived from real waste solidified in cement and bitumen. Relatively high plutonium mobilities were observed in the case of the former and in the case of the real waste leachates, cesium was found to be the least retarded. The sorption of ruthenium was found to be largely associated with the insoluble residues of the natural rock salt rather than the halite itself. (orig./RB)

  10. TWRS HLW interim storage facility search and evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Calmus, R.B., Westinghouse Hanford

    1996-05-16

    The purpose of this study was to identify and provide an evaluation of interim storage facilities and potential facility locations for the vitrified high-level waste (HLW) from the Phase I demonstration plant and Phase II production plant. In addition, interim storage facilities for solidified separated radionuclides (Cesium and Technetium) generated during pretreatment of Phase I Low-Level Waste Vitrification Plant feed was evaluated.

  11. Management strategy for site characterization at candidate HLW repository sites

    International Nuclear Information System (INIS)

    Bartlett, J.W.

    1988-01-01

    This paper describes a management strategy for HLW repository site characterization which is aimed at producing an optimal characterization trajectory for site suitability and licensing evaluations. The core feature of the strategy is a matrix of alternative performance targets and alternative information-level targets which can be used to allocate and justify program effort. Strategies for work concerning evaluation of expected and disrupted repository performance are distinguished, and the need for issue closure criteria is discussed

  12. R and D programme for HLW disposal in Japan

    International Nuclear Information System (INIS)

    Tsuboya, Takao

    1997-01-01

    The Power Reactor and Nuclear Fuel Development Corporation (PNC) has been active in developing an R and D programme for high-level radioactive waste (HLW) disposal in accordance with the overall HLW management programme defined by the Atomic Energy Commission (AEC) of Japan. The aim of the R and D activities at the current stage is to provide a scientific and technical basis for the geological disposal of HLW in Japan, which is turn promotes understanding of the safety concept not only in the scientific and technical community but also by the general public. As a major milestone in the R and D programme, PNC submitted a first progress report, referred to as H3, in September 1992. H3 summarised the results of R and D activities up to March 1992 and identified priority issues for further study. The second progress report, scheduled to be submitted around 2000, and should demonstrated more rigorously and transparently the feasibility of the specified disposal concept. It should also provide input for the siting and regulatory processes, which will be set in motion after the year 2000. (author). 10 refs., 4 figs

  13. Modelling spent fuel and HLW behaviour in repository conditions

    Energy Technology Data Exchange (ETDEWEB)

    Esparza, A M; Esteban, J A

    2003-07-01

    The aim of this report is to give the reader an overall insight of the different models, which are used to predict the long-term behaviour of the spent fuels and HLW disposed in a repository. The models must be established on basic data and robust kinetics describing the mechanisms controlling spent fuel alteration/dissolution in a repository. The UO2 matrix, or source term, contains embedded in it the , majority of radionuclides of the spent fuel (some are in the gap cladding). For this reason the SF radionuclides release models play a significant role in the performance assessment of radioactive waste disposal. The differences existing between models published in the literature are due to the conceptual understanding of the processes and the degree of the conservatism used with the parameter values, and the boundary conditions. They mainly differ in their level of simplification and their final objective. Sometimes are focused the show compliance with regulatory requirements, other to support decision making, to increase the level of confidence of public and scientific community, could be empirical, semi-empirical or analytical. The models take into account the experimental results from radionuclides releases and their extrapolation to the very long term. Its necessary a great statistics for have a representative dissolution rate, due at the number of experimental results is not very high and many of them show a great scatter, independently of theirs different compositions by axial and radial variations, due to linear power or local burnup. On the other hand, it is difficult to predict the spent fuel behaviour over the long term, based in short term experiments. In this report is given a little description of the radionuclides distribution in the spent fuel and also in the cladding/pellet gap, grain boundary, cracks and rim zones (the matrix rim zone can be considered with an especial characteristics very different to the rest of the spent fuel), and structural

  14. Modelling spent fuel and HLW behaviour in repository conditions

    International Nuclear Information System (INIS)

    Esparza, A. M.; Esteban, J. A.

    2003-01-01

    The aim of this report is to give the reader an overall insight of the different models, which are used to predict the long-term behaviour of the spent fuels and HLW disposed in a repository. The models must be established on basic data and robust kinetics describing the mechanisms controlling spent fuel alteration/dissolution in a repository. The UO2 matrix, or source term, contains embedded in it the , majority of radionuclides of the spent fuel (some are in the gap cladding). For this reason the SF radionuclides release models play a significant role in the performance assessment of radioactive waste disposal. The differences existing between models published in the literature are due to the conceptual understanding of the processes and the degree of the conservatism used with the parameter values, and the boundary conditions. They mainly differ in their level of simplification and their final objective. Sometimes are focused the show compliance with regulatory requirements, other to support decision making, to increase the level of confidence of public and scientific community, could be empirical, semi-empirical or analytical. The models take into account the experimental results from radionuclides releases and their extrapolation to the very long term. Its necessary a great statistics for have a representative dissolution rate, due at the number of experimental results is not very high and many of them show a great scatter, independently of theirs different compositions by axial and radial variations, due to linear power or local burnup. On the other hand, it is difficult to predict the spent fuel behaviour over the long term, based in short term experiments. In this report is given a little description of the radionuclides distribution in the spent fuel and also in the cladding/pellet gap, grain boundary, cracks and rim zones (the matrix rim zone can be considered with an especial characteristics very different to the rest of the spent fuel), and structural

  15. Enhanced HLW glass formulations for the waste treatment and immobilization plant

    Energy Technology Data Exchange (ETDEWEB)

    Kruger, Albert A. [DOE-WTP Project Office, US Department of Energy, Richland, Washington (United States)

    2013-07-01

    Current estimates and glass formulation efforts are conservative vis-a-vis achievable waste loadings. These formulations have been specified to ensure that glasses are homogenous, contain essentially no crystalline phases, are processable in joule-heated, ceramic-lined melters and meet WTP Contract terms. The WTP's overall mission will require the immobilization of tank waste compositions that are dominated by mixtures of aluminum, chromium, bismuth, iron, phosphorous, zirconium, and sulfur compounds as waste-limiting components. Glass compositions for these waste mixtures have been developed based upon previous experience and current glass property models. DOE has a testing program to develop and characterize HLW glasses with higher waste loadings. This work has demonstrated the feasibility of increases in waste loading from 25 wt% to 33-50 wt% (based on oxide loading) in the glass depending on the waste stream. It is expected these higher waste loading glasses will reduce the HLW canister production requirement by 25% or more. (authors)

  16. High Level Waste (HLW) Processing Experience with Increased Waste Loading

    International Nuclear Information System (INIS)

    JANTZEN, CAROL

    2004-01-01

    The Defense Waste Processing Facility (DWPF) Engineering requested characterization of glass samples that were taken after the second melter had been operational for about 5 months. After the new melter had been installed, the waste loading had been increased to about 38 weight percentage after a new quasicrystalline liquidus model had been implemented. The DWPF had also switched from processing with refractory Frit 200 to a more fluid Frit 320. The samples were taken after DWPF observed very rapid buildup of deposits in the upper pour spout bore and on the pour spout insert while processing the high waste loading feedstock. These samples were evaluated using various analytical techniques to determine the cause of the crystallization. The pour stream sample was homogeneous, amorphous, and representative of the feed batch from which it was derived. Chemical analysis of the pour stream sample indicated that a waste loading of 38.5 weight per cent had been achieved. The data analysis indicated that surface crystallization, induced by temperature and oxygen fugacity gradients in the pour spout, caused surface crystallization to occur in the spout and on the insert at the higher waste loadings even though there was no crystallization in the pour stream

  17. HLW vitrification in France industrial experience and glass quality

    International Nuclear Information System (INIS)

    Desvaux, J.L.; Delahaye, P.

    1994-01-01

    This paper describes the vitrification process, the technology and process improvements at the La Hague plant in R 7 and T 7 facilities. The main achievements relate to the process flexibility, the reliability of the equipment and solid waste management. The quality of the vitrified glass produced and canisters compliance with agreed specifications are demonstrated through characterization studies. Since the active start-up of R 7/T 7 facilities, canisters compliance with specifications relies upon a complete quality assurance/quality control program including process control. 1 tab., 1 fig

  18. Spent fuel and HLW transportation: The French experience

    International Nuclear Information System (INIS)

    Giraud, J.P.; Charles, J.L.

    1996-01-01

    Transportation of nuclear materials is a key component of the nuclear industry. Transportation takes place in the public domain. This relation with both population and environment has put transportation in the middle of a hard-toned debate. The nuclear transportation system has demonstrated its maturity in terms of safety, reliability, cost-efficiency and environmental respect, through a remarkable track record of successful accomplishments. For 30 years, large quantities of nuclear materials have been shipped smoothly and safely in France, Europe and overseas. General principles and safety rules have been carefully established by international experts, published as recommendations by the IAEA and enforced worldwide through national legislations. The international nuclear industry closely follows this framework, operating with comprehensive quality assurance programs

  19. Design and validation of the THMC China-Mock-Up test on buffer material for HLW disposal

    Directory of Open Access Journals (Sweden)

    Yuemiao Liu

    2014-04-01

    Full Text Available According to the preliminary concept of the high-level radioactive waste (HLW repository in China, a large-scale mock-up facility, named China-Mock-Up was constructed in the laboratory of Beijing Research Institute of Uranium Geology (BRIUG. A heater, which simulates a container of radioactive waste, is placed inside the compacted Gaomiaozi (GMZ-Na-bentonite blocks and pellets. Water inflow through the barrier from its outer surface is used to simulate the intake of groundwater. The numbers of water injection pipes, injection pressure and the insulation layer were determined based on the numerical modeling simulations. The current experimental data of the facility are herein analyzed. The experiment is intended to evaluate the thermo-hydro-mechano-chemical (THMC processes occurring in the compacted bentonite-buffer during the early stage of HLW disposal and to provide a reliable database for numerical modeling and further investigation of engineered barrier system (EBS, and the design of HLW repository.

  20. Development of geological disposal system; localization of element cost data and cost evaluation on the HLW repository

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Byung Sik; Kim, Kil Jung; Yang, Young Jin; Kim, Sung Chun [KOPEC, Taejeon (Korea)

    2002-03-01

    To estimate Total Life Cycle Cost (TSLCC) for Korea HLW Repository through localization of element cost data, we review and re-organize each basic element cost data for reference repository system, localize various element cost and finally estimate TSLCC considering economic parameters. As results of the study, TSLCC is estimated as 17,167,689 million won, which includes costs for site preparation, surface facilities, underground facilities and management/integration. Since HLW repository Project is an early stage of pre-conceptual design at present, the information of design and project information are not enough to perform cost estimate and cost localization for the Project. However, project cost structure is re-organized based on the local condition and Total System Life Cycle Cost is estimated using the previous cost data gathered from construction experience of the local nuclear power plant. Project results can be used as basic reference data to assume total construction cost for the local HLW repository and should be revised to more reliable cost data with incorporating detail project design information into the cost estimate in a future. 20 refs. (Author)

  1. Situation concerning the HLW repository in Germany

    International Nuclear Information System (INIS)

    Lempert, J.P.

    1992-01-01

    Final disposal of radioactive waste has been defined in Germany as: maintenance-free, safe emplacement of radioactive waste, time unlimited and no intention of retrievability. The responsibility for final disposal lies in the hands of the German Federal Government, which has assigned a federal authority to plan, erect and operate the federal facilities for long-term storage of nuclear waste. The federal authority has in lack of industrial experience contracted my company DBE which is responsible for the engineering, erection and operation of all German nuclear waste repositories. (author)

  2. Investigation of water-soluble organic matter extracted from shales during leaching experiments

    Science.gov (United States)

    Zhu, Yaling; Vieth-Hillebrand, Andrea; Wilke, Franziska D. H.; Horsfield, Brian

    2017-04-01

    The huge volumes and unknown composition of flowback and produced waters cause major public concerns about the environmental and social compatibility of hydraulic fracturing and the exploitation of gas from unconventional reservoirs. Flowback and produced waters contain not only residues of fracking additives but also chemical species that are dissolved from the shales themselves during fluid-rock interaction. Knowledge of the composition, size and structure of dissolved organic carbon (DOC) as well as the main controls on the release of DOC are a prerequisite for a better understanding of these interactions and its effects on composition of flowback and produced water. Black shales from four different geological settings and covering a maturity range Ro = 0.3-2.6% were extracted with deionized water. The DOC yields were found to decrease rapidly with increasing diagenesis and remain low throughout catagenesis. Four DOC fractions have been qualitatively and quantitatively characterized using size-exclusion chromatography. The concentrations of individual low molecular weight organic acids (LMWOA) decrease with increasing maturity of the samples except for acetate extracted from the overmature Posidonia shale, which was influenced by hydrothermal brines. The oxygen content of the shale organic matter also shows a significant influence on the release of organic acids, which is indicated by the positive trend between oxygen index (OI) and the concentrations of formate and acetate. Based on our experiments, both the properties of the organic matter source and the thermal maturation progress of the shale organic matter significantly influence the amount and quality of extracted organic compounds during the leaching experiments.

  3. Concept development for HLW disposal research tunnel

    International Nuclear Information System (INIS)

    Queon, S. K.; Kim, K. S.; Park, J. H.; Jeo, W. J.; Han, P. S.

    2003-01-01

    In order to dispose high-level radioactive waste in a geological formation, it is necessary to assess the safety of a disposal concept by excavating a research tunnel in the same geological formation as the host rock mass. The design concept of a research tunnel depends on the actual disposal concept, repository geometry, experiments to be carried at the tunnel, and geological conditions. In this study, analysis of the characteristics of the disposal research tunnel, which is planned to be constructed at KAERI site, calculation of the influence of basting impact on neighbor facilities, and computer simuation for mechanical stability analysis using a three-dimensional code, FLAC3D, had been carried out to develop the design concept of the research tunnel

  4. 12 Flasktransport of vitrified High Level Waste (HLW)

    Energy Technology Data Exchange (ETDEWEB)

    Verdier, A.; Lancelot, J. [COGEMA Logistics (AREVA Group) (France); Gisbertz, A.; Graf, W. [GNS (Germany); Bartagnon, O. [COGEMA (AREVA Group) (France)

    2004-07-01

    The return of HLW to Germany has started in 1996 with the first attribution of 28 glass canisters to German utilities by COGEMA. After several transports comprising 1, 2 and 6 flasks per shipment German and French Authorities requested to transport 12 flasks in a single shipment. The first of these 12-flask-transports was performed with the type CASTOR {sup registered} HAW 20/28 CG flask in 2002 and the second followed in 2003. COGEMA LOGISTICS is responsible for the overall transport assigned by GNS (Gesellschaft fuer Nuklear-Service mbH) being itself entrusted by the German utilities with the return of reprocessing residues.

  5. Chemical compatibility of HLW borosilicate glasses with actinides

    International Nuclear Information System (INIS)

    Walker, C.T.; Scheffler, K.; Riege, U.

    1978-11-01

    During liquid storage of HLLW the formation of actinide enriched sludges is being expected. Also during melting of HLW glasses an increase of top-to-bottom actinide concentrations can take place. Both effects have been studied. Besides, the vitrification of plutonium enriched wastes from Pu fuel element fabrication plants has been investigated with respect to an isolated vitrification process or a combined one with the HLLW. It is shown that the solidification of actinides from HLLW and actinide waste concentrates will set no principal problems. The leaching of actinides has been measured in salt brine at 23 0 C and 115 0 C. (orig.) [de

  6. Rheology of Savannah River site tank 42 HLW radioactive sludge

    International Nuclear Information System (INIS)

    Ha, B.C.

    1997-01-01

    Knowledge of the rheology of the radioactive sludge slurries at the Savannah River Site is necessary in order to ensure that they can be retrieved from waste tanks and processed for final disposal. At Savannah River Site, Tank 42 sludge represents on of the first HLW radioactive sludges to be vitrified in the Defense Waste Processing Facility. The rheological properties of unwashed Tank 42 sludge slurries at various solids concentrations were measured remotely in the Shielded Cells at the Savannah River Technology Center using a modified Haake Rotovisco viscometer

  7. 12 Flasktransport of vitrified High Level Waste (HLW)

    International Nuclear Information System (INIS)

    Verdier, A.; Lancelot, J.; Gisbertz, A.; Graf, W.; Bartagnon, O.

    2004-01-01

    The return of HLW to Germany has started in 1996 with the first attribution of 28 glass canisters to German utilities by COGEMA. After several transports comprising 1, 2 and 6 flasks per shipment German and French Authorities requested to transport 12 flasks in a single shipment. The first of these 12-flask-transports was performed with the type CASTOR registered HAW 20/28 CG flask in 2002 and the second followed in 2003. COGEMA LOGISTICS is responsible for the overall transport assigned by GNS (Gesellschaft fuer Nuklear-Service mbH) being itself entrusted by the German utilities with the return of reprocessing residues

  8. Role of international collaboration in PNC's R ampersand D programme for HLW disposal

    International Nuclear Information System (INIS)

    Masuda, Sumio; Umeki, Hiroyuki; Yamakawa, Minoru

    1996-01-01

    PNC has been active in promoting international cooperation in connection with the Japanese HLW disposal programme, based on both a bilateral and multilateral approach. Both types of cooperation are extremely useful; in particular, bilateral cooperation has the advantage of providing opportunities for in-depth discussions in mutual areas of interest. By way of contrast, multilateral cooperation also provides an international arena for broader discussion and corroboration of output from individual R ampersand D programmes. International collaboration also provides young researchers with an opportunity to learn from experience. Depending on the issues to be tackled, appropriate forms of collaboration have been integrated into PNC's strategy for maximizing output. The lessons learned from collaboration are very valuable and can be used directly in their programme to enhance its credibility. The format of collaboration has also been extensively developed: it has been found that resources can be utilized more effectively by sharing them appropriately

  9. Effect of composition on peraluminous glass properties: An application to HLW containment

    Science.gov (United States)

    Piovesan, V.; Bardez-Giboire, I.; Perret, D.; Montouillout, V.; Pellerin, N.

    2017-01-01

    Part of the Research and Development program concerning high level nuclear waste (HLW) glasses aims to assess new glass formulations able to incorporate a high waste content with enhanced properties in terms of thermal stability, chemical durability, and process ability. This study focuses on peraluminous glasses of the SiO2 - Al2O3 - B2O3 - Na2O - Li2O - CaO - La2O3 system, defined by an excess of aluminum ions Al3+ in comparison with modifier elements such as Na+, Li+ or Ca2+. To understand the effect of composition on physical properties of glasses (viscosity, density, Tg), a Design Of Experiments (DOE) approach was applied to investigate the peraluminous glass domain. The influence of each oxide was quantified to build predictive models for each property. Lanthanum and lithium oxides appear to be the most influential factors on peraluminous glass properties.

  10. The chemical stockpile intergovernmental consultation program: Lessons for HLW public involvement

    International Nuclear Information System (INIS)

    Feldman, D.L.

    1991-01-01

    This paper assesses the appropriateness of the US Army's Chemical Stockpile Disposal Program's (CSDP) Intergovernmental Consultation and Coordination Boards (ICCBs) as models for incorporating public concerns in the future siting of HLW repositories by DOE. ICCB structure, function, and implementation are examined, along with other issues relevant to the HLW context. 27 refs

  11. Comparison of risks due to HLW and SURF repositories in bedded salt

    International Nuclear Information System (INIS)

    Chu, M.S.Y.; Ortiz, N.R.; Wahi, K.K.

    1983-01-01

    A methodology was developed for use in the analysis of risks from geologic disposal of nuclear wastes. This methodology is applied to two conceptual nuclear waste repositories in bedded salt containing High-Level Waste (HLW) and Spent Un-Reprocessed Fuel (SURF), respectively. A comparison of the risk estimated from the HLW and SURF repositories is presented

  12. Melter Throughput Enhancements for High-Iron HLW

    Energy Technology Data Exchange (ETDEWEB)

    Kruger, A. A. [Department of Energy, Office of River Protection, Richland, Washington (United States); Gan, Hoa [The Catholic University of America, Washington, DC (United States); Joseph, Innocent [The Catholic University of America, Washington, DC (United States); Pegg, Ian L. [The Catholic University of America, Washington, DC (United States); Matlack, Keith S. [The Catholic University of America, Washington, DC (United States); Chaudhuri, Malabika [The Catholic University of America, Washington, DC (United States); Kot, Wing [The Catholic University of America, Washington, DC (United States)

    2012-12-26

    This report describes work performed to develop and test new glass and feed formulations in order to increase glass melting rates in high waste loading glass formulations for HLW with high concentrations of iron. Testing was designed to identify glass and melter feed formulations that optimize waste loading and waste processing rate while meeting all processing and product quality requirements. The work included preparation and characterization of crucible melts to assess melt rate using a vertical gradient furnace system and to develop new formulations with enhanced melt rate. Testing evaluated the effects of waste loading on glass properties and the maximum waste loading that can be achieved. The results from crucible-scale testing supported subsequent DuraMelter 100 (DM100) tests designed to examine the effects of enhanced glass and feed formulations on waste processing rate and product quality. The DM100 was selected as the platform for these tests due to its extensive previous use in processing rate determination for various HLW streams and glass compositions.

  13. NOx AND HETEROGENEITY EFFECTS IN HIGH LEVEL WASTE (HLW)

    International Nuclear Information System (INIS)

    Meisel, Dan; Camaioni, Donald M.; Orlando, Thom

    2000-01-01

    We summarize contributions from our EMSP supported research to several field operations of the Office of Environmental Management (EM). In particular we emphasize its impact on safety programs at the Hanford and other EM sites where storage, maintenance and handling of HLW is a major mission. In recent years we were engaged in coordinated efforts to understand the chemistry initiated by radiation in HLW. Three projects of the EMSP (''The NOx System in Nuclear Waste,'' ''Mechanisms and Kinetics of Organic Aging in High Level Nuclear Wastes, D. Camaioni--PI'' and ''Interfacial Radiolysis Effects in Tanks Waste, T. Orlando--PI'') were involved in that effort, which included a team at Argonne, later moved to the University of Notre Dame, and two teams at the Pacific Northwest National Laboratory. Much effort was invested in integrating the results of the scientific studies into the engineering operations via coordination meetings and participation in various stages of the resolution of some of the outstanding safety issues at the sites. However, in this Abstract we summarize the effort at Notre Dame

  14. Argon solubility in liquid steel

    NARCIS (Netherlands)

    Boom, R; Dankert, O; Van Veen, A; Kamperman, AA

    2000-01-01

    Experiments have been performed to establish the solubility of argon in liquid interstitial-free steel. The solubility appears to be lower than 0.1 at ppb, The results are in line with argon solubilities reported in the literature on liquid iron. Semiempirical theories and calculations based on the

  15. Mineral surface processes responsible for the decreased retardation (or enhanced mobilization) of 137Cs from HLW tank discharges. 1998 annual progress report

    International Nuclear Information System (INIS)

    Bertsch, P.M.; Zachara, J.M.

    1998-01-01

    'Cesium (137) is a major component of high level weapons waste. At Hanford, single shell tanks (SST''s) with high level wastes (HLW) have leaked supernate containing over 10 6 Ci of 137 Cs and other co-contaminants into the vadose zone. In select locations, 137 Cs has migrated further than expected from retardation experiments and performance assessment calculations. Deep 137 Cs migration has been observed beneath the SX tank farm at Hanford with REDOX wastes as the carrier causing regulatory and stakeholder concern. The causes for expedited migration are unclear. This research is investigating how the sorption chemistry of Cs on Hanford vadose zone sediments changes after contact with solutions characteristic of HLW. The central scientific hypothesis is that the high Na concentration of HLW will suppress surface-exchange reactions of Cs, except those to highly-selective frayed edge sites (FES) of the micaceous fraction. The authors further speculate that the concentrations, ion selectivity, and structural aspects of the FES will change after contact with HLW and that these changes will be manifest in the macroscopic sorption behavior of Cs. The authors believe that migration predictions of Cs can be improved substantially if such changes are understood and quantified. The research has three objectives: (1.) identify how the multi-component surface exchange behavior of Cs on Hanford sediments changes after contact with HLW simulants that span a range of relevant chemical (Na, OH, Al, K) and temperature conditions (23-80 C); (2) reconcile changes in sorption chemistry with microscopic and molecular changes in site distribution, chemistry, mineralogy, and surface structure of the micaceous fraction; (3) integrate mass-action-solution exchange measurements with changes in the structure/site distribution of the micaceous fraction to yield a multicomponent exchange model relevant to high ionic strength and hydroxide for prediction of environmental Cs sorption.'

  16. DM100 AND DM1200 MELTER TESTING WITH HIGH WASTE LOADING FORMULATIONS FOR HANFORD HIGH-ALUMINUM HLW STREAMS, TEST PLAN 09T1690-1

    International Nuclear Information System (INIS)

    Kruger, A.A.; Matlack, K.S.; Kot, W.K.; Pegg, I.L.; Joseph, I.

    2009-01-01

    glass melting rate. The WTP HLW melter has a glass surface area of 3.75 m 2 and depth of ∼1.1 m. The two melters in the HLW facility together are designed to produce up to 7.5 MT of glass per day at 100% availability. Further increases in HLW waste processing rates can potentially be achieved by increasing the melter operating temperature above 1150 C and by increasing the waste loading in the glass product Increasing the waste loading also has the added benefit of decreasing the number of canisters for storage. The current estimates and glass formulation efforts have been conservative in terms of achievable waste loadings. These formulations have been specified to ensure that the glasses are homogenous, contain essentially no crystalline phases, are processable in joule-heated, ceramic-lined melters and meet WTP contract requirements. The WTP's overall mission will require the immobilization oftank waste compositions that are dominated by mixtures of aluminum (Al), chromium (Cr), bismuth (Bi), iron (Fe), phosphorous (P), zirconium (Zr), and sulfur (S) compounds as waste-limiting components. Glass compositions for these waste mixtures have been developed based upon previous experience and current glass property models. Recently, DOE has initiated a testing program to develop and characterize HLW glasses with higher waste loadings. Results of this work have demonstrated the feasibility of increases in waste-loading from about 25 wt% to 33-50 wt% (based on oxide loading) in the glass depending on the waste stream. It is expected that these higher waste loading glasses will reduce the HLW canister production requirement by about 25% or more.

  17. Suggestions on selection of clay site as a key alternative of underground repository for HLW geological disposal in China

    International Nuclear Information System (INIS)

    Zheng Hualing; Fu Bingjun; Fan Xianhua; Chen Shi; Sun Donghui

    2006-01-01

    Site selection for the underground repository is a vital problem with respect to the HLW geological disposal. Over the past decades, we have been focusing our attention on granite as a priority in China. However, there are some problems have to be discussed on this matter. In this paper, both experiences gained and lessons learned in the international community regarding the site selection are described. And then, after analyzing a lot of some key factors affecting the site selection, some comments and suggestions on selection of clay site as a key alternative before final decision making in China are presented. (authors)

  18. RECENT PROCESS IMPROVEMENTS TO INCREASE HLW THROUGHPUT AT THE DWPF

    International Nuclear Information System (INIS)

    Herman, C

    2007-01-01

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

  19. Development of gap filling technique in HLW repository

    International Nuclear Information System (INIS)

    Nakashima, Hitoshi; Saito, Akira; Ishii, Takashi; Toguri, Satohito; Okihara, Mitsunobu; Iwasa, Kengo

    2016-01-01

    HLW is supposed to be disposed underground at depths more than 300 m in Japan. Buffer is an artificial barrier that controls radionuclides migrating into the groundwater. The buffer would be made of a natural swelling clay, bentonite. Construction technology for the buffer has been studied for many years, but studies for the gaps surrounding the buffer are little. The proper handling of the gaps is important for guaranteeing the functions of the buffer. In this paper, gap filling techniques using bentonite pellets have been developed in order to the gap having the same performance as the buffer. A new method for manufacturing high-density spherical pellets has been developed to fill the gap higher density ever reported. For the bentonite pellets, the filling performance and how to use were determined. And full-scale filling tests provided availability of the bentonite pellets and filling techniques. (author)

  20. Comparing technical concepts for disposal of Belgian vitrified HLW

    International Nuclear Information System (INIS)

    Bel, J.; Bock, C. de; Boyazis, J.P.

    2004-01-01

    The choice of a suitable repository design for different categories of radioactive waste is an important element in the decisional process that will eventually lead to the waste disposal in geological ground layers during the next decades. Most countries are in the process of elaborating different technical solutions for their EBS '. Considering possible design alternatives offers more flexibility to cope with remaining uncertainties and allows optimizing some elements of the EBS in the future. However, it is not feasible to continue carrying out detailed studies for a large number of alternative design options. At different stages in the decisional process, choices, even preliminary ones, have to be made. Although the impact of different stakeholders (regulator, waste agencies, waste producers, research centers,...) in making these design choices can differ from one country to another, the choices should be based on sound, objective, clear and unambiguous justification grounds. Moreover, the arguments should be carefully reported and easy to understand by the decision makers. ONDRAF/NIRAS recently elaborated three alternative designs for the disposal of vitrified HLW. These three designs are briefly described in the next section. A first series of technological studies pointed out that the three options are feasible. It would however be unreasonable to continue R and D work on all three alternatives in parallel. It is therefore planned to make a preliminary choice of a reference design for the vitrified HLW in 2003. This selection will depend on the way the alternative design options can be evaluated against a number of criteria, mainly derived from general repository design requirements. The technique of multi-criteria analysis (MCA) will be applied as a tool for making the optimum selection, considering all selection criteria and considering different strategic approaches. This paper describes the used methodology. The decision on the actual selection will be

  1. Setting up a safe deep repository for long-lived HLW and ILW in Russia: Current state of the works

    International Nuclear Information System (INIS)

    Polyakov, Yu.D.; Porsov, A.Yu.; Beigul, V.P.; Palenov, M.V.

    2014-01-01

    The concept of RW disposal in Russia in accordance with the Federal Law 'On Radioactive Waste Management and Amendments to Specific Legal Acts of the Russian Federation' No. 190-FL dated 11 July 2011, is oriented at the ultimate disposal of waste, without an intent for their subsequent retrieval. The law 190-FL has it as follows: - A radioactive waste repository is a radioactive waste storage facility intended for disposal of the radioactive wastes without an intent for their subsequent retrieval. - Disposal of solid long-lived high-level waste and solid long-lived intermediate-level waste is carried out in deep repositories for radioactive waste. - Import into the Russian Federation of radioactive waste for the purpose of its storage, processing and disposal, except for spent sealed sources of ionising radiation originating from the Russian Federation, is prohibited. For safe final disposal of long-lived HLW and ILW, it is planned to construct a deep repository for radioactive waste (DRRW) in a low-pervious monolith rock massif in the Krasnoyarsk region in the production territory of the Mining and Chemical Combine (FSUE 'Gorno-khimicheskiy kombinat'). According to the IAEA recommendations and in line with the international experience in feasibility studies for setting up of HLW and SNF underground disposal facilities, the first mandatory step is the construction of an underground research laboratory. An underground laboratory serves the following purposes: - itemised research into the characteristics of enclosing rock mass, with verification of massive material suitability for safe disposal of long-lived HLW and ILW; - research into and verification of the isolating properties of an engineering barrier system; - development of engineering solutions and transportation and process flow schemes for construction and running of a future RW ultimate isolation facility. (authors)

  2. Cocrystal solubility-pH and drug solubilization capacity of sodium dodecyl sulfate – mass action model for data analysis and simulation to improve design of experiments

    Directory of Open Access Journals (Sweden)

    Alex Avdeef

    2018-06-01

    solubility product of cocrystals (coupled with predicted k values described here allowed for simulations of solubility-pH speciation profiles of cocrystal systems in the presence of SDS. Well in advance of any actual measurements, these simulations can be used to probe conditions favorable to the design of cocrystal experiments where SDS stabilizes cocrystal suspensions against drug precipitation over a predicted range of pH values.

  3. Redox Control For Hanford HLW Feeds VSL-12R2530-1, REV 0

    International Nuclear Information System (INIS)

    Kruger, A. A.; Matlack, Keith S.; Pegg, Ian L.; Kot, Wing K.; Joseph, Innocent

    2012-01-01

    The principal objectives of this work were to investigate the effects of processing simulated Hanford HLW at the estimated maximum concentrations of nitrates and oxalates and to identify strategies to mitigate any processing issues resulting from high concentrations of nitrates and oxalates. This report provides results for a series of tests that were performed on the DM10 melter system with simulated C-106/AY-102 HLW. The tests employed simulated HLW feeds containing variable amounts of nitrates and waste organic compounds corresponding to maximum concentrations proj ected for Hanford HLW streams in order to determine their effects on glass production rate, processing characteristics, glass redox conditions, melt pool foaming, and the tendency to form secondary phases. Such melter tests provide information on key process factors such as feed processing behavior, dynamic effects during processing, processing rates, off-gas amounts and compositions, foaming control, etc., that cannot be reliably obtained from crucible melts

  4. Redox Control For Hanford HLW Feeds VSL-12R2530-1, REV 0

    Energy Technology Data Exchange (ETDEWEB)

    Kruger, A. A. [Department of Energy, Office of River Protection, Richland, Washington (United States); Matlack, Keith S. [The Catholic University of America, Washington, DC (United States); Pegg, Ian L. [The Catholic University of America, Washington, DC (United States); Kot, Wing K. [The Catholic University of America, Washington, DC (United States); Joseph, Innocent [The Catholic University of America, Washington, DC (United States)

    2012-12-13

    The principal objectives of this work were to investigate the effects of processing simulated Hanford HLW at the estimated maximum concentrations of nitrates and oxalates and to identify strategies to mitigate any processing issues resulting from high concentrations of nitrates and oxalates. This report provides results for a series of tests that were performed on the DM10 melter system with simulated C-106/AY-102 HLW. The tests employed simulated HLW feeds containing variable amounts of nitrates and waste organic compounds corresponding to maximum concentrations proj ected for Hanford HLW streams in order to determine their effects on glass production rate, processing characteristics, glass redox conditions, melt pool foaming, and the tendency to form secondary phases. Such melter tests provide information on key process factors such as feed processing behavior, dynamic effects during processing, processing rates, off-gas amounts and compositions, foaming control, etc., that cannot be reliably obtained from crucible melts.

  5. Conclusions on the two technical panels on HLW-disposal and waste treatment processes respectively

    International Nuclear Information System (INIS)

    Dinkespiller, J.A.; Dejonghe, P.; Feates, F.

    1986-01-01

    The paper reports the concluding panel session at the European Community Conference on radioactive waste management and disposal, Luxembourg 1985. The panel considered the conclusions of two preceeding technical panels on high level waste (HLW) disposal and waste treatment processes. Geological disposal of HLW, waste management, safety assessment of waste disposal, public opinion, public acceptance of the manageability of radioactive wastes, international cooperation, and waste management in the United States, are all discussed. (U.K.)

  6. Legal precedents regarding use and defensibility of risk assessment in Federal transportation of SNF and HLW

    International Nuclear Information System (INIS)

    Bentz, E.J. Jr.; Bentz, C.B.; O'Hora, T.D.; Chen, S.Y.

    1997-01-01

    Risk assessment has become an increasingly important and essential tool in support of Federal decision-making regarding the handling, storage, disposal, and transportation of spent nuclear fuel (SNF) and high-level radioactive waste (HLW). This paper analyzes the current statutory and regulatory framework and related legal precedents with regard to SNF and HLW transportation. The authors identify key scientific and technical issues regarding the use and defensibility of risk assessment in Federal decision-making regarding anticipated shipments

  7. Nuclide transport models for HLW repository safety assessment in Finland, Japan, Sweden, and Canada

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Young Myoung; Kang, Chul Hyung; Hwang, Yong Soo; Choi, Jong Won; Kim, Sung Gi; Koh, Won Il

    1997-10-01

    Disposal and design concepts in such countries as Sweden, Finland, Canada and Japan which have already published safety assessment reports for the HLW repositories have been reviewed mainly in view of nuclide transport models used in their assessment. This kind of review would be very helpful in doing similar research in Korea where research program regarding HLW has been just started. (author). 44 refs., 2 tabs., 30 figs

  8. Simulation of HTM processes in buffer-rock barriers based on the French HLW disposal concept

    International Nuclear Information System (INIS)

    Li, Xiaoshuo; Roehlig, Klaus-Juergen; Zhang, Chunliang

    2012-01-01

    Document available in extended abstract form only. The main objectives of this paper are to gain experience with modelling and analysis of HTM processes in clay rock and bentonite buffer surrounding heat-generating radioactive waste. The French concept for HLW disposal in drifts with backfilled bentonite buffer considered in numerical calculations which are carried out by using the computer code CODE-BRIGHT developed by the Technical University of Catalonia in Barcelona. The French repository designed by ANDRA is located in the middle of the Callovo-Oxfordian argillaceous formation (COX) of 250 m thickness at a depth of 500 to 630 m below the surface. The French concept has been simplified at this simulation work. A drift is considered to be excavated at a depth of 500 m below the surface. It has a diameter of 2.2 m and a length of 20 m. A large volume of the rock mass around the drift is taken into account by an axisymmetric model of 100 m radius and 100 m length. In fact, this model represents a cylindrical rock-buffer-system with the central axis of the containers, as shown in Figure 1. Some points are selected in the buffer and the rock along the radial line (dash yellow line) in the middle of the drift for recording HTM parameters with time. The display and analysis of the results at this paper are chiefly along this line. The simulation work has been divided to two time steps. At the first step, the drift excavation and ventilation is simulated by reducing the stress normal to the drift wall down to zero and circulating gas along the drift wall with relative humidity of 85 %. Following the drift excavation and ventilation, the HLW containers and the bentonite are emplaced in the drift as the second step of the simulation. This is simulated by simultaneously applying the initial conditions of the buffer and the decayed heat emitting from the waste containers as thermal boundary conditions. Two materials (Clay rock and bentonite buffer) are taken into account

  9. Public Perspectives in the Japanese HLW Disposal Program

    Energy Technology Data Exchange (ETDEWEB)

    Inatsugu, Shigefumi; Takeuchi, Mitsuo; Kato, Toshiaki [Nuclear Waste Management Organization of Japan (NUNIO), Tokyo (Japan)

    2006-09-15

    Following legislation entitled the 'Specified Radioactive Waste Final Disposal Act', the Nuclear Waste Management Organization of Japan (NUMO) was established in October 2000 as the implementing organization for geological disposal of vitrified high-level waste (HLW). Implementation of NUMO's disposal project will be based on three principles: 1) respecting public initiative and opinion, 2) adopting a stepwise approach and 3) ensuring transparency in information disclosure. NUMO has decided to adopt an open solicitation approach to finding volunteer municipalities for Preliminary Investigation Areas (PIAs). The official announcement of the start of the open solicitation program was made in 2002. Although no official applications had been received from volunteer municipalities by the end of 2005, NUMO has been continuing to carry out various activities aimed specifically at public communication and encouraging dialogue about the deep geological disposal project This paper summarizes the results obtained and lessons learned so far and identifies the issues that NUMO must tackle immediately in the areas of communication and dialogue.

  10. Stress analysis of HLW containers advanced test work Compas project

    International Nuclear Information System (INIS)

    Ove Arup and Partners

    1990-01-01

    The Compas project is concerned with the structural performance of metal overpacks which may be used to encapsulate vitrified high-level waste forms before disposal in deep geological repositories. This document describes the activities performed between June and August 1989 forming the advanced test work phase of this project. This is the culmination of two years' analysis and test work to demonstrate whether the analytical ability exists to model containers subjected to realistic loads. Three mild steel containers were designed and manufactured to be one-third scale models of a realistic HLW container, modified to represent the effect of anisotropic loading and to facilitate testing. The containers were tested under a uniform external pressure and all failed by buckling in the mid-body region. The outer surface of each container was comprehensively strain-gauged to provide strain history data at all positions of interest. In parallel with the test work, Compas project partners, from five different European countries, independently modelled the behaviour of each of the containers using their computer codes to predict the failure pressure and produce strain history data at a number of specified locations. The first axisymmetric container was well modelled but predictions for the remaining two non-axisymmetric containers were much more varied, with differences of up to 50% occurring between failure predictions and test data

  11. Technical and economic optimization study for HLW waste management

    International Nuclear Information System (INIS)

    Deffes, A.

    1989-01-01

    This study was conducted to assess the technical and economic aspects of high level waste (HLW) management with the objective of optimizing the interim storage duration and the dimensions of the underground repository site. The procedure consisted in optimizing the economic criterion under specified constraints. The results are intended to identify trends and guide the choice from among available options; simple and highly flexible models were therefore used in this study, and only nearfield thermal constraints were taken into consideration. Because of the present uncertainty on the physicochemical properties of the repository environment and on the unit cost figures, this study focused on developing a suitable method rather than on obtaining definitive results. With the physical and economic data bases used for the two media investigated (granite and salt) the optimum values found show that it is advisable to minimize the interim storage time, and that the geological repository should feature a high degree of spatial dilution. These results depend to a considerable extent on the assumption of high interim storage costs

  12. Public Perspectives in the Japanese HLW Disposal Program

    International Nuclear Information System (INIS)

    Inatsugu, Shigefumi; Takeuchi, Mitsuo; Kato, Toshiaki

    2006-01-01

    Following legislation entitled the 'Specified Radioactive Waste Final Disposal Act', the Nuclear Waste Management Organization of Japan (NUMO) was established in October 2000 as the implementing organization for geological disposal of vitrified high-level waste (HLW). Implementation of NUMO's disposal project will be based on three principles: 1) respecting public initiative and opinion, 2) adopting a stepwise approach and 3) ensuring transparency in information disclosure. NUMO has decided to adopt an open solicitation approach to finding volunteer municipalities for Preliminary Investigation Areas (PIAs). The official announcement of the start of the open solicitation program was made in 2002. Although no official applications had been received from volunteer municipalities by the end of 2005, NUMO has been continuing to carry out various activities aimed specifically at public communication and encouraging dialogue about the deep geological disposal project This paper summarizes the results obtained and lessons learned so far and identifies the issues that NUMO must tackle immediately in the areas of communication and dialogue

  13. Thermal analysis of the vertical disposal for HLW

    International Nuclear Information System (INIS)

    Zhao Honggang; Wang Ju; Liu Yuemiao; Su Rui

    2013-01-01

    The temperature on the canister surface is set to be no more than 100℃ in the high level radioactive waste (HLW) repository, it is a criterion to dictate the thermal dimension of the repository. The factors that affect the temperature on the canister surface include the initial power of the canister, the thermal properties of material as the engineered barrier system (EBS), the gaps around the canister in the EBS, the initial ground temperature and thermal properties of the host rock, the repository layout, etc. This article examines the thermal properties of the material in host rock and the EBS, the thermal conductivity properties of the different gaps in the EBS, the temperature evolution around the single canister by using the analysis method and the numerical method. The findings are as follows: 1) The most important and the sensitive parameter is the initial disposal power of the canister; 2) The two key factors that affect the highest temperature on the canister surface are the parameter of uncertainty and nature variability of material as the host rock and the EBS, and the gaps around the canister in the EBS; 3) The temperature difference between the canister and bentonite is no more than 10℃ , and the bigger the inner gaps are, the bigger the temperature difference will be; when the gap between the bentonite and the host rock is filled with water, the temperature difference becomes small, but it will be 1∼3℃ higher than the gaps filled will air. (authors)

  14. Biosphere modelling for a HLW repository - scenario and parameter variations

    International Nuclear Information System (INIS)

    Grogan, H.

    1985-03-01

    In Switzerland high-level radioactive wastes have been considered for disposal in deep-lying crystalline formations. The individual doses to man resulting from radionuclides entering the biosphere via groundwater transport are calculated. The main recipient area modelled, which constitutes the base case, is a broad gravel terrace sited along the south bank of the river Rhine. An alternative recipient region, a small valley with a well, is also modelled. A number of parameter variations are performed in order to ascertain their impact on the doses. Finally two scenario changes are modelled somewhat simplistically, these consider different prevailing climates, namely tundra and a warmer climate than present. In the base case negligibly low doses to man in the long term, resulting from the existence of a HLW repository have been calculated. Cs-135 results in the largest dose (8.4E-7 mrem/y at 6.1E+6 y) while Np-237 gives the largest dose from the actinides (3.6E-8 mrem/y). The response of the model to parameter variations cannot be easily predicted due to non-linear coupling of many of the parameters. However, the calculated doses were negligibly low in all cases as were those resulting from the two scenario variations. (author)

  15. Rheology of Savannah River Site Tank 51 HLW radioactive sludge

    International Nuclear Information System (INIS)

    Ha, B.C.

    1993-01-01

    Savannah River Site (SRS) Tank 51 HLW radioactive sludge represents a major portion of the first batch of sludge to be vitrified in the Defense Waste Processing Facility (DWPF) at SRS. The rheological properties of Tank 51 sludge will determine if the waste sludge can be pumped by the current DWPF process cell pump design and the homogeneity of melter feed slurries. The rheological properties of Tank 51 sludge and sludge/frit slurries at various solids concentrations were measured remotely in the Shielded Cells Operations (SCO) at the Savannah River Technology Center (SRTC) using a modified Haake Rotovisco viscometer system. Rheological properties of Tank 51 radioactive sludge/Frit 202 slurries increased drastically when the solids content was above 41 wt %. The yield stresses of Tank 51 sludge and sludge/frit slurries fall within the limits of the DWPF equipment design basis. The apparent viscosities also fall within the DWPF design basis for sludge consistency. All the results indicate that Tank 51 waste sludge and sludge/frit slurries are pumpable throughout the DWPF processes based on the current process cell pump design, and should produce homogeneous melter feed slurries

  16. Tc Chemistry in HLW: Role of Organic Complexants

    International Nuclear Information System (INIS)

    Hess, Nancy S.; Conradsen, Steven D.

    2003-01-01

    Tc complexation with organic compounds in tank waste plays a significant role in the redox chemistry of Tc and the partitioning of Tc between the supernatant and sludge components in waste tanks. These processes need to be understood so that strategies to effectively remove Tc from high-level nuclear waste prior to waste immobilization can be developed and so that long-term consequences of Tc remaining in residual waste after sludge removal can be evaluated. Only limited data on the stability of Tc-organic complexes exists and even less thermodynamic data on which to develop predictive models of Tc chemical behavior is available. To meet these challenges we are conducting a research program to study to develop thermodynamic data on Tc-organic complexation over a wide range of chemical conditions. We will attempt to characterize Tc-speciation in actual tank waste using state-of-the-art analytical organic chemistry, separations, and speciation techniques to validate our model. On the basis of such studies we will develop credible model of Tc chemistry in HLW that will allow prediction of Tc speciation in tank waste and Tc behavior during waste pretreatment processing and in waste tank residuals

  17. 'Practicality' as a key constraint to HLW repository design

    International Nuclear Information System (INIS)

    Kitayama, Kazumi; Sakabe, Yasushi; Ishiguro, Katsuhiko

    2007-01-01

    Designs of repositories in Japan for HLW have focused very much on demonstration of post-closure safety. Safety can be assured using very simple assessment techniques, which make many conservative simplifications. Such a situation is reasonable for the early stages of generic concept demonstration, but becomes less appropriate as NUMO moves towards siting, where a number of issues involved with construction and operation of a repository - generally grouped together as 'practicality'. The engineering logistics and conventional safety of repository construction and operation have been relatively little studied and present major challenges. Current designs emphasise a minimum of infrastructure in the emplacement tunnels and remote-handled operation. This would be difficult enough, but such operations need to be carried out to strict quality limits and need to be robust in the event of equipment failure or disruptive events. The paper will first examine how designs can be modified from the viewpoint of logistics. The implications of such modifications on operational robustness and associated safety in case of perturbation scenarios are then considered. (author)

  18. Modelling of radionuclide migration and heat transport from an High-Level-Radioactive-Waste-repository (HLW) in Boom clay

    International Nuclear Information System (INIS)

    Put, M.; Henrion, P.

    1992-01-01

    For the modelling of the migration of radionuclides in the Boom clay formation, the analytical code MICOF has been updated with a 3-dimensional analytical solution for discrete sources. the MICOF program is used for the calculation of the release of α and β emitters from the HIGH LEVEL RADIOACTIVE WASTES (HLW). A coherent conceptual model is developed which describes all the major physico-chemical phenomena influencing the migration of radionuclides in the Boom clay. The concept of the diffusion accessible porosity is introduced and included in the MICOF code. Different types of migration experiments are described with their advantages and disadvantages. The thermal impact of the HLW disposal in the stratified Boom clay formation has been evaluated by a finite element simulation of the coupled heat and mass transport equation. The results of the simulations show that under certain conditions thermal convection cells may form, but the convective heat transfer in the clay formation is negligible. 6 refs., 19 figs., 2 tabs., 5 appendices

  19. Effect of composition on peraluminous glass properties: An application to HLW containment

    Energy Technology Data Exchange (ETDEWEB)

    Piovesan, V. [CEA, DEN, DTCD, SECM, LDMC – Marcoule, F-30207 Bagnols sur Cèze (France); CNRS, CEMHTI UPR3079, Univ. Orléans, F-45071 Orléans (France); Bardez-Giboire, I., E-mail: isabelle.giboire@cea.fr [CEA, DEN, DTCD, SECM, LDMC – Marcoule, F-30207 Bagnols sur Cèze (France); Perret, D. [CEA, DEN, DTCD, SECM, LDMC – Marcoule, F-30207 Bagnols sur Cèze (France); Montouillout, V.; Pellerin, N. [CNRS, CEMHTI UPR3079, Univ. Orléans, F-45071 Orléans (France)

    2017-01-15

    Part of the Research and Development program concerning high level nuclear waste (HLW) glasses aims to assess new glass formulations able to incorporate a high waste content with enhanced properties in terms of thermal stability, chemical durability, and process ability. This study focuses on peraluminous glasses of the SiO{sub 2} – Al{sub 2}O{sub 3} – B{sub 2}O{sub 3} – Na{sub 2}O – Li{sub 2}O – CaO – La{sub 2}O{sub 3} system, defined by an excess of aluminum ions Al{sup 3+} in comparison with modifier elements such as Na{sup +}, Li{sup +} or Ca{sup 2+}. To understand the effect of composition on physical properties of glasses (viscosity, density, T{sub g}), a Design Of Experiments (DOE) approach was applied to investigate the peraluminous glass domain. The influence of each oxide was quantified to build predictive models for each property. Lanthanum and lithium oxides appear to be the most influential factors on peraluminous glass properties. - Highlights: • A Design of Experiment approach to link composition and glass properties. • Adding alkali decreases glass transition temperature. • Adding La{sub 2}O{sub 3} strongly decreases glass melt viscosity. • Adding La{sub 2}O{sub 3} increases density.

  20. Stress analysis of HLW containers. Preliminary ring test exercise Compas project

    International Nuclear Information System (INIS)

    1989-01-01

    This document describes the series of experiments and associated calculations performed as the Compas preliminary ring test exercise. A number of mild steel rings, representative of sections through HLW containers, some notched and pre-cracked, were tested in compression right up to and beyond their ultimate load. The Compas project partners independently modelled the behaviour of these rings using their finite element codes. Four different ring types were tested, and each test was repeated three times. For three of the ring types, the three test repetitions gave identical results. The fourth ring, which was not modelled by the partners, had a 4 mm thick layer of weld metal deposited on its surface. The three tests on this ring did not give identical results and suggested that the effect of welding methods should be addressed at a later stage of the project. Fracture was not found to be a significant cause of ring failure. The results of the ring tests were compared with the partners predictions, and additionally some time was spent assessing where the use of the codes could be improved. This exercise showed that the partners codes have the ability to produce results within acceptable limits. Most codes were unable to model stable crack growth. There were indications that some codes would not be able to cope with a significantly more complex three-dimensional analysis

  1. Environmental risk assessment: its contribution to criteria development for HLW disposal

    International Nuclear Information System (INIS)

    Smith, G.M.; Little, R.H.; Watkins, B.M.

    1999-01-01

    Principles for radioactive waste management have been provided by the International Atomic Energy Agency in Safety Series No.111-F, which was published in 1995. This has been a major step forward in the process of achieving acceptance for proposals for disposal of radioactive waste, for example, for High Level Waste disposal in deep repositories. However, these principles have still to be interpreted and developed into practical radiation protection criteria. Without prejudicing final judgements on the acceptability of waste proposals, an important aspect is that practical demonstration of compliance (or the opposite) with these criteria must be possible. One of the IAEA principles requires that radioactive waste shall be managed in such a way as to provide an acceptable level of protection of the environment. There has been and continues to be considerable debate as to how to demonstrate compliance with such a principle. This paper briefly reviews the current status and considers how experience in other areas of environmental protection could contribute to criteria development for HLW disposal

  2. Solubility of Lead Sulfate in Water and in Sodium Sulfate Solutions: An Experiment in Atomic Absorption Spectrophotometry.

    Science.gov (United States)

    Lehman, Thomas A.; Everett, Wayne W.

    1982-01-01

    Describes a set of undergraduate laboratory experiments which provide experience in deuteration and derivatization procedures applied to infrared spectroscopy. Basic skills in vacuum-line technique are also taught while measuring infrared spectra of deuterated solid samples and demonstrating the value of derivatization as an aid to interpreting…

  3. Development of thermal analysis method for the near field of HLW repository using ABAQUS

    Energy Technology Data Exchange (ETDEWEB)

    Kuh, Jung Eui; Kang, Chul Hyung; Park, Jeong Hwa [Korea Atomic Energy Research Institute, Taejon (Korea)

    1998-10-01

    An appropriate tool is needed to evaluate the thermo-mechanical stability of high level radioactive waste (HLW) repository. In this report a thermal analysis methodology for the near field of HLW repository is developed to use ABAQUS which is one of the multi purpose FEM code and has been used for many engineering area. The main contents of this methodology development are the structural and material modelling to simulate a repository, setup of side conditions, e.g., boundary and load conditions, and initial conditions, and the procedure to selection proper material parameters. In addition to these, the interface programs for effective production of input data and effective change of model size for sensitivity analysis for disposal concept development are developed. The results of this work will be apply to evaluate the thermal stability and to use as main input data for mechanical analysis of HLW repository. (author). 20 refs., 15 figs., 5 tabs.

  4. The Results of HLW Processing Using Zirconium Salt of Dibutyl phosphoric Acid in Hot Cell

    Energy Technology Data Exchange (ETDEWEB)

    Fedorov, Yu.S.; Zilberman, B.Ya.; Shmidt, O.V. [Khlopin Radium Institute, 2nd Murinsky Ave., 28, Saint-Petersburg, 194021 (Russian Federation)

    2008-07-01

    Zirconium salt of dibutyl phosphoric acid (ZS HDBP), is an effective solvent for liquid HLW and ILW (high and intermediate level wastes) processing with radionuclide partitioning into different groups for further immobilization according to radiotoxicity. The rig trials in mixer-settles in hot cells were carried out using 30 L of real HLW containing transplutonium (TPE), rare earths (RE), Sr and Cs in 2 mol/L HNO{sub 3}, characterized by total specific activity 520 MBk/L. The recovery factor for TPE and RE was as high as 10{sup 4}, but only 10 for Sr. Purification factor of TPE and RE from Cs and Sr was 10{sup 4}, and that of Sr from TPE and Cs was 10{sup 3}. Almost all Cs was localized in the second cycle raffinate. So Zr salt of HDBP can be used in HLW processing with radionuclide partitioning with respect to the categories of radiotoxicity. (authors)

  5. Regulatory status on the safety assessment of a HLW repository in other countries

    International Nuclear Information System (INIS)

    Lee, Sung Ho; Hwang, Yong Soo

    2008-12-01

    To construct a HLW repository, it is essential to meet the requirements on the regulation for a deep geological disposal. Even if the construction of a HLW repository is determined positively, technical standards which assert the performance of a repository will be needed. Among various technical standards, safety assessment based on the repository evolution in the future will play an important role in the licensing process. The foreign countries' technical standards on the safety assessment of a HLW repository may be an indicator to carry out the R and D activities on geological disposal effectively. In this report, assessment period, limit of radiation dose and uncertainty related to the safety assessment are investigated and analyzed in detail. Especially, the technical reviews of USA regulation bodies seems to be reasonable in the point of the intrinsic attribute of safety assessment

  6. Study on evaluation method for potential effect of natural phenomena on a HLW disposal system

    International Nuclear Information System (INIS)

    Kawamura, Makoto; Makino, Hitoshi; Umeda, Koji; Osawa, Hideaki; Seo, Toshihiro; Ishimaru, Tsuneaki

    2005-01-01

    Evaluation for the potential effect of natural phenomena on a HLW disposal system is an important issue in safety assessment. A scenario construction method for the effects on a HLW disposal system condition and performance has been developed for two purposes: the first being effective elicitation and organization of information from investigators of natural phenomena and performance assessor and the second being, maintenance of traceability of scenario construction processes with suitable records. In this method, a series of works to construct scenarios is divided into pieces to facilitate and to elicit the features of potential effect of natural phenomena on a HLW disposal system and is organized to create reasonable scenarios with consistency, traceability and adequate conservativeness within realistic view. (author)

  7. Support for HLW Direct Feed - Phase 2, VSL-15R3440-1

    Energy Technology Data Exchange (ETDEWEB)

    Matlack, K. S. [The Catholic Univ. of America, Washington, DC (United States); Pegg, I. [The Catholic Univ. of America, Washington, DC (United States); Joseph, I. [EnergySolutions, Columbia, MD (United States); Kot, W. K. [The Catholic Univ. of America, Washington, DC (United States)

    2017-03-20

    This report describes work performed to develop and test new glass and feed formulations originating from a potential flow-sheet for the direct vitrification of High Level Waste (HLW) with minimal or no pretreatment. In the HLW direct feed option that is under consideration for early operations at the Hanford Tank Waste Treatment and Immobilization Plant (WTP), the pretreatment facility would be bypassed in order to support an earlier start-up of the vitrification facility. For HLW, this would mean that the ultrafiltration and caustic leaching operations that would otherwise have been performed in the pretreatment facility would either not be performed or would be replaced by an interim pretreatment function (in-tank leaching and settling, for example). These changes would likely affect glass formulations and waste loadings and have impacts on the downstream vitrification operations. Modification of the pretreatment process may result in: (i) Higher aluminum contents if caustic leaching is not performed; (ii) Higher chromium contents if oxidative leaching is not performed; (iii) A higher fraction of supernate in the HLW feed resulting from the lower efficiency of in-tank washing; and (iv) A higher water content due to the likely lower effectiveness of in-tank settling compared to ultrafiltration. The HLW direct feed option has also been proposed as a potential route for treating HLW streams that contain the highest concentrations of fast-settling plutoniumcontaining particles, thereby avoiding some of the potential issues associated with such particles in the WTP Pretreatment facility [1]. In response, the work presented herein focuses on the impacts of increased supernate and water content on wastes from one of the candidate source tanks for the direct feed option that is high in plutonium.

  8. Cesium and strontium fractionation from HLW for thermal-stress reduction in a geologic repository

    International Nuclear Information System (INIS)

    McKee, R.W.

    1983-02-01

    Results are described for a study to assess the benefits and costs of fractionating the cesium and strontium components in commercial high-level waste (HLW) to a separate waste stream for the purpose of reducing geologic repository thermal stresses. System costs are developed for a broad range of conditions comparing the Cs/Sr fractionation concept with disposal of 10-year old vitrified HLW and vitrified HLW aged to achieve (through decay) the same heat output as the fractionated high-level waste (FHLW). All comparisons are based on a 50,000 metric ton equivalent (MTE) system. The FHLW and the Cs/Sr waste are both disposed of a vitrified waste but emplaced in separate areas of a basalt repository. The FHLW is emplaced in high-integrity packages at relatively high waste loading but low heat loading, while the Cs/Sr waste is emplaced in minimum integrity packages at relatively high heat loading. System cost comparisons are based on minimum cost combinations of canister diameter, waste concentration, and canister spacing in a basalt repository for each waste type. The effects on both long- and near-term safety considerations are also addressed. The major conclusion is that the Cs/Sr fractionation concept offers, potentially, a substantial total system cost advantage for HLW disposal if reduced HLW package temperatures in a basalt repository are desired. However, there is no cost advantage if currently designated maximum design temperatures are acceptable. Aging the HLW for 50 to 100 years can accomplish similar results at equivalent or loser costs

  9. Using process instrumentation to obviate destructive examination of canisters of HLW glass

    International Nuclear Information System (INIS)

    Kuhn, W.L.; Slate, S.C.

    1983-01-01

    An important concern of a manufacturer of packages of solidified high-level waste (HLW) is quality assurance of the waste form. The vitrification of HLW as a borosilicate glass is considered, and, based on a reference vitrification process, it is proposed that information from process instrumentation may be used to assure quality without the need for additional information obtained by destructive examining (core drilling) canisters of glass. This follows mainly because models of product performance and process behavior must be previously established in order to confidently select the desired glass formulation, and to have confidence that the process is well enough developed to be installed and operated in a nuclear facility

  10. LIQUIDUS TEMPERATURE AND ONE PERCENT CRYSTAL CONTENT MODELS FOR INITIAL HANFORD HLW GLASSES

    International Nuclear Information System (INIS)

    Vienna, John D.; Edwards, Tommy B.; Crum, Jarrod V.; Kim, Dong-Sang; Peeler, David K.

    2005-01-01

    Preliminary models for liquidus temperature (TL) and temperature at 1 vol% crystal (T01) applicable to WTP HLW glasses in the spinel primary phase field were developed. A series of literature model forms were evaluated using consistent sets of data form model fitting and validation. For TL, the ion potential and linear mixture models performed best, while for T01 the linear mixture model out performed all other model forms. TL models were able to predict with smaller uncertainty. However, the lower T01 values (even with higher prediction uncertainties) were found to allow for a much broader processing envelope for WTP HLW glasses

  11. Solubility and speciation results from oversaturation experiments on neptunium, plutonium and americium in a neutral electrolyte with a total carbonate similar to water from Yucca Mountain Region Well UE- 25p No. 1

    International Nuclear Information System (INIS)

    Torretto, P.; Becraft, K.; Prussin, T.; Roberts, K.; Carpenter, S.; Hobart, D.; Nitsche, H.

    1995-12-01

    Solubility and speciation are important in understanding aqueous radionuclide transport through the geosphere. They define the source term for transport retardation processes such as sorption and colloid formation. Solubility and speciation data are useful in verifying the validity of geochemical codes that are a part of predictive transport models. Solubility experiments will approach solution equilibrium from both oversaturation and undersaturation. In these experiments, we have approached the solubility equilibrium from oversaturation, Results are given for solubility and speciation experiments from oversaturation of 237 NpO 2 + 239 Pu 4+ , and 241 Am 3+ /Nd 3+ in a neutral electrolyte containing a total carbonate concentration similar to groundwater from the Yucca Mountain region, Nevada, which is being investigated as a potential high-level nuclear waste disposal site, at 25 degrees C and three pH values. In these experiments, the solubilitycontrolling steady-state solids were identified and the speciation and/or oxidation states present in the supernatant solutions were determined

  12. Advances in Glass Formulations for Hanford High-Alumimum, High-Iron and Enhanced Sulphate Management in HLW Streams-13000

    International Nuclear Information System (INIS)

    Kruger, Albert A.

    2013-01-01

    The current estimates and glass formulation efforts have been conservative in terms of achievable waste loadings. These formulations have been specified to ensure that the glasses are homogenous, contain essentially no crystalline phases, are processable in joule-heated, ceramic-lined melters and meet Hanford Tank Waste Treatment and Immobilization Plant (WTP) Contract terms. The WTP's overall mission will require the immobilization of tank waste compositions that are dominated by mixtures of aluminum (Al), chromium (Cr), bismuth (Bi), iron (Fe), phosphorous (P), zirconium (Zr), and sulphur (S) compounds as waste-limiting components. Glass compositions for these waste mixtures have been developed based upon previous experience and current glass property models. Recently, DOE has initiated a testing program to develop and characterize HLW glasses with higher waste loadings and higher throughput efficiencies. Results of this work have demonstrated the feasibility of increases in waste loading from about 25 wt% to 33-50 wt% (based on oxide loading) in the glass depending on the waste stream. In view of the importance of aluminum limited waste streams at Hanford (and also Savannah River), the ability to achieve high waste loadings without adversely impacting melt rates has the potential for enormous cost savings from reductions in canister count and the potential for schedule acceleration. Consequently, the potential return on the investment made in the development of these enhancements is extremely favorable. Glass composition development for one of the latest Hanford HLW projected compositions with sulphate concentrations high enough to limit waste loading have been successfully tested and show tolerance for previously unreported tolerance for sulphate. Though a significant increase in waste loading for high-iron wastes has been achieved, the magnitude of the increase is not as substantial as those achieved for high-aluminum, high-chromium, high-bismuth or sulphur

  13. The production of advanced glass ceramic HLW forms using cold crucible induction melter

    International Nuclear Information System (INIS)

    Rutledge, V.J.; Maio, V.

    2013-01-01

    Cold Crucible Induction Melters (CCIM) will favorably change how High-Level radioactive Waste (from nuclear fuel recovery) is treated in a near future. Unlike the existing Joule-Heated Melters (JHM) currently in operation for the glass-based immobilization of High-Level Waste (HLW), CCIM offers unique material features that will increase melt temperatures, increase throughput, increase mixing, increase loading in the waste form, lower melter foot prints, eliminate melter corrosion and lower costs. These features not only enhance the technology for producing HLW forms, but also provide advantageous attributes to the waste form by allowing more durable alternatives to glass. It is concluded that glass ceramic waste forms that are tailored to immobilize fission products of HLW can be can be made from the HLW processed with the CCIM. The advantageous higher temperatures reached with the CCIM and unachievable with JHM allows the lanthanides, alkali, alkaline earths, and molybdenum to dissolve into a molten glass. Upon controlled cooling they go into targeted crystalline phases to form a glass ceramic waste form with higher waste loadings than achievable with borosilicate glass waste forms. Natural cooling proves to be too fast for the formation of all targeted crystalline phases

  14. A GoldSim Based Biosphere Assessment Model for a HLW Repository

    International Nuclear Information System (INIS)

    Lee, Youn-Myoung; Hwang, Yong-Soo; Kang, Chul-Hyung

    2007-01-01

    To demonstrate the performance of a repository, the dose exposure to a human being due to nuclide releases from a repository should be evaluated and the results compared to the dose limit presented by the regulatory bodies. To evaluate a dose rate to an individual due to a long-term release of nuclides from a HLW repository, biosphere assessment models and their implemented codes such as ACBIO1 and ACBIO2 have been developed with the aid of AMBER during the last few years. BIOMASS methodology has been adopted for a HLW repository currently being considered in Korea, which has a similar concept to the Swedish KBS-3 HLW repository. Recently, not just only for verifying the purpose for biosphere assessment models but also for varying the possible alternatives to assess the consequences in a biosphere due to a HLW repository, another version of the assessment modesl has been newly developed in the frame of development programs for a total system performance assessment modeling tool by utilizing GoldSim. Through a current study, GoldSim approach for a biosphere modeling is introduced. Unlike AMBER by which a compartment scheme can be rather simply constructed with an appropriate transition rate between compartments, GoldSim was designed to facilitate the object-oriented modules by which specific models can be addressed in an additional manner, like solving jig saw puzzles

  15. Design options for HLW repository operation technology. (4) Shotclay technique for seamless construction of EBS

    International Nuclear Information System (INIS)

    Kobayashi, Ichizo; Fujisawa, Soh; Nakajima, Makoto; Toida, Masaru; Nakashima, Hitoshi; Asano, Hidekazu

    2011-01-01

    The shotclay method is construction method of the high density bentonite engineered barrier by spraying method. Using this method, the dry density of 1.6 Mg/m 3 , which was considered impossible with the spray method, is achieved. In this study, the applicability of the shotclay method to HLW bentonite-engineered barriers was confirmed experimentally. In the tests, an actual scale vertical-type HLW bentonite-engineered barrier was constructed. This was a bentonite-engineered barrier with a diameter of 2.22 m and a height of 3.13 m. The material used was bentonite with 30% silica sand, and water content was adjusted by mixing chilled bentonite with powdered ice before thawing. Work progress was 11.2 m 3 and the weight was 21.7 Mg. The dry density of the entire buffer was 1.62 Mg/m 3 , and construction time was approximately 8 hours per unit. After the formworks were removed, the core and block of the actual scale HLW bentonite-engineered barrier were sampled to confirm homogeneity. As a result, homogeneity was confirmed, and no gaps were observed between the formwork and the buffer material and between the simulated waste and the buffer material. The applicability to HLW of the shotclay method has been confirmed through this examination. (author)

  16. Disposal of defense spent fuel and HLW from the Idaho Chemical Processing Plant

    International Nuclear Information System (INIS)

    Ermold, L.F.; Loo, H.H.; Klingler, R.D.; Herzog, J.D.; Knecht, D.A.

    1992-12-01

    Acid high-level radioactive waste (HLW) resulting from fuel reprocessing at the Idaho Chemical Processing Plant (ICPP) for the US Department of Energy (DOE) has been solidified to a calcine since 1963 and stored in stainless steel bins enclosed by concrete vaults. Several different types of unprocessed irradiated DOE-owned fuels are also in storage ate the ICPP. In April, 1992, DOE announced that spent fuel would no longer be reprocessed to recover enriched uranium and called for a shutdown of the reprocessing facilities at the ICPP. A new Spent Fuel and HLW Technology Development program was subsequently initiated to develop technologies for immobilizing ICPP spent fuels and HLW for disposal, in accordance with the Nuclear Waste Policy Act. The Program elements include Systems Analysis, Graphite Fuel Disposal, Other Spent Fuel Disposal, Sodium-Bearing Liquid Waste Processing, Calcine Immobilization, and Metal Recycle/Waste Minimization. This paper presents an overview of the ICPP radioactive wastes and current spent fuels, with an emphasis on the description of HLW and spent fuels requiring repository disposal

  17. HLW Salt Disposition Alternatives Identification Preconceptual Phase I Summary Report (Including Attachments)

    International Nuclear Information System (INIS)

    Piccolo, S.F.

    1999-01-01

    The purpose of this report is to summarize the process used by the Team to systematically develop alternative methods or technologies for final disposition of HLW salt. Additionally, this report summarizes the process utilized to reduce the total list of identified alternatives to an ''initial list'' for further evaluation. This report constitutes completion of the team charter major milestone Phase I Deliverable

  18. Final Report Tests On The Duramelter 1200 HLW Pilot Melter System Using AZ-101 HLW Simulants VSL-02R0100-2, Rev. 1, 2/17/03

    International Nuclear Information System (INIS)

    Kruger, A.A.; Matlack, K.S.; Kot, W.K.; Bardakci, T.; Gong, W.; D'Angelo, N.A.; Schatz, T.R.; Pegg, I.L.

    2011-01-01

    This document provides the final report on data and results obtained from a series of nine tests performed on the one-third scale DuraMelter(trademark) 1200 (DM1200) HLW Pilot Melter system that has been installed at VSL with an integrated prototypical off-gas treatment system. That system has replaced the DM1000 system that was used for HLW throughput testing during Part B1 (1). Both melters have similar melt surface areas (1.2 m 2 ) but the DM1200 is prototypical of the present RPP-WTP HLW melter design whereas the DM1000 was not. These tests were performed under a corresponding RPP-WTP Test Specification and associated Test Plans. The nine tests reported here were preceded by an initial series of short-duration tests conducted to support the start-up and commissioning of this system. This report is a followup to the previously issued Preliminary Data Summary Reports. The DM1200 system was deployed for testing and confirmation of basic design, operability, flow sheet, and process control assumptions as well as for support of waste form qualification and permitting. These tests include data on processing rates, off-gas treatment system performance, recycle stream compositions, as well as process operability and reliability. Consequently, this system is a key component of the overall HLW vitrification development strategy. The primary objective of the present series of tests was to determine the effects of a variety of parameters on the glass production rate in comparison to the RPP-WTP HL W design basis of 400 kg/m 2 /d. Previous testing on the DMIOOO system (1) concluded that achievement of that rate with simulants of projected WTP melter feeds (AZ-101 and C-106/AY-102) was unlikely without the use of bubblers. As part of those tests, the same feed that was used during the cold-commissioning of the West Valley Demonstration Project (WVDP) HLW vitrification system was run on the DM1000 system. The DM1000 tests reproduced the rates that were obtained at the larger

  19. FINAL REPORT TESTS ON THE DURAMELTER 1200 HLW PILOT MELTER SYSTEM USING AZ-101 HLW SIMULANTS VSL-02R0100-2 REV 1 2/17/03

    Energy Technology Data Exchange (ETDEWEB)

    KRUGER AA; MATLACK KS; KOT WK; BARDAKCI T; GONG W; D' ANGELO NA; SCHATZ TR; PEGG IL

    2011-12-29

    This document provides the final report on data and results obtained from a series of nine tests performed on the one-third scale DuraMelter{trademark} 1200 (DM1200) HLW Pilot Melter system that has been installed at VSL with an integrated prototypical off-gas treatment system. That system has replaced the DM1000 system that was used for HLW throughput testing during Part B1 [1]. Both melters have similar melt surface areas (1.2 m{sup 2}) but the DM1200 is prototypical of the present RPP-WTP HLW melter design whereas the DM1000 was not. These tests were performed under a corresponding RPP-WTP Test Specification and associated Test Plans. The nine tests reported here were preceded by an initial series of short-duration tests conducted to support the start-up and commissioning of this system. This report is a followup to the previously issued Preliminary Data Summary Reports. The DM1200 system was deployed for testing and confirmation of basic design, operability, flow sheet, and process control assumptions as well as for support of waste form qualification and permitting. These tests include data on processing rates, off-gas treatment system performance, recycle stream compositions, as well as process operability and reliability. Consequently, this system is a key component of the overall HLW vitrification development strategy. The primary objective of the present series of tests was to determine the effects of a variety of parameters on the glass production rate in comparison to the RPP-WTP HL W design basis of 400 kg/m{sup 2}/d. Previous testing on the DMIOOO system [1] concluded that achievement of that rate with simulants of projected WTP melter feeds (AZ-101 and C-106/AY-102) was unlikely without the use of bubblers. As part of those tests, the same feed that was used during the cold-commissioning of the West Valley Demonstration Project (WVDP) HLW vitrification system was run on the DM1000 system. The DM1000 tests reproduced the rates that were obtained at the

  20. Dissolution of ORNL HLW sludge and partitioning of the actinides using the TRUEX process

    International Nuclear Information System (INIS)

    Spencer, B.B.; Egan, B.Z.; Beahm, E.C.; Chase, C.W.; Dillow, T.A.

    1997-01-01

    Experiments were conducted to evaluate the transuranium extraction (TRUEX) process for partitioning actinides from actual dissolved high-level radioactive waste (HLW) sludge. Samples of sludge from melton Valley Storage Tank W-25 were rinsed with mild caustic (0.2 M NaOH) to reduce the concentrations of nitrates and fission products associated with the interstitial liquid. In one campaign the rinsed sludge was leached in nitric acid, and about 50% of the dry mass of the sludge was dissolved. The resulting solution contained total metal concentrations of ∼ 1.8 M with a nitric acid concentration of 2.9 M. In the other campaign the sludge was neutralized with nitric acid to destroy the carbonates, then leached with 2.6 M NaOH for ∼ 6 h before rinsing with the mild caustic. The sludge was then leached in nitric acid, and about 80% of the sludge dissolved. The resulting solution contained total metal concentrations of ∼ 0.6 M with a nitric acid concentration of 1.7 M. Chemical analyses of both phases were used to evaluate the process. Evaluation was based on two metrics: the fraction of TRU elements removed from the dissolved sludge and comparison of the results with predictions made with the Generic TRUEX Model (GTM). The fractions of Eu, Pu, Cm, Th and U species removed from aqueous solution in only one extraction stage were > 95% and were close to the values predicted by the GTM. Mercury was also found to be strongly extracted, with a one-stage removal of > 92%. In one test, vanadium appeared to be moderately extracted

  1. Optimization of Soluble Expression and Purification of Recombinant Human Rhinovirus Type-14 3C Protease Using Statistically Designed Experiments: Isolation and Characterization of the Enzyme.

    Science.gov (United States)

    Antoniou, Georgia; Papakyriacou, Irineos; Papaneophytou, Christos

    2017-10-01

    Human rhinovirus (HRV) 3C protease is widely used in recombinant protein production for various applications such as biochemical characterization and structural biology projects to separate recombinant fusion proteins from their affinity tags in order to prevent interference between these tags and the target proteins. Herein, we report the optimization of expression and purification conditions of glutathione S-transferase (GST)-tagged HRV 3C protease by statistically designed experiments. Soluble expression of GST-HRV 3C protease was initially optimized by response surface methodology (RSM), and a 5.5-fold increase in enzyme yield was achieved. Subsequently, we developed a new incomplete factorial (IF) design that examines four variables (bacterial strain, expression temperature, induction time, and inducer concentration) in a single experiment. The new design called Incomplete Factorial-Strain/Temperature/Time/Inducer (IF-STTI) was validated using three GST-tagged proteins. In all cases, IF-STTI resulted in only 10% lower expression yields than those obtained by RSM. Purification of GST-HRV 3C was optimized by an IF design that examines simultaneously the effect of the amount of resin, incubation time of cell lysate with resin, and glycerol and DTT concentration in buffers, and a further 15% increase in protease recovery was achieved. Purified GST-HRV 3C protease was active at both 4 and 25 °C in a variety of buffers.

  2. Ibuprofen nanocrystals developed by 22 factorial design experiment: A new approach for poorly water-soluble drugs

    Directory of Open Access Journals (Sweden)

    A.R. Fernandes

    2017-12-01

    Full Text Available The reduction of the particle size of drugs of pharmaceutical interest down to the nano-sized range has dramatically changed their physicochemical properties. The greatest disadvantage of nanocrystals is their inherent instability, due to the risk of crystal growth. Thus, the selection of an appropriate stabilizer is crucial to obtain long-term physicochemically stable nanocrystals. High pressure homogenization has enormous advantages, including the possibility of scaling up, lack of organic solvents and the production of small particles diameter with low polydispersity index. The sequential use of high shear homogenization followed by high pressure homogenization, can modulate nanoparticles’ size for different administration routes. The present study focuses on the optimization of the production process of two formulations composed of different surfactants produced by High Shear Homogenization followed by hot High Pressure Homogenization. To build up the surface response charts, a 22 full factorial design experiment, based on 2 independent variables, was used to develop optimized formulations. The effects of the production process on the mean particle size and polydispersity index were evaluated. The best ibuprofen nanocrystal formulations were obtained using 0.20% Tween 80 and 1.20% PVP K30 (F1 and 0.20% Tween 80 and 1.20% Span 80 (F2. The estimation of the long-term stability of the aqueous suspensions of ibuprofen nanocrystals was studied using the LUMISizer. The calculated instability index suggests that F1 was more stable when stored at 4 °C and 22 °C, whereas F2 was shown to be more stable when freshly prepared.

  3. Chemical analysis of simulated high level waste glasses to support stage III sulfate solubility modeling

    Energy Technology Data Exchange (ETDEWEB)

    Fox, K. M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-03-17

    The U.S. Department of Energy (DOE), Office of Environmental Management (EM) is sponsoring an international, collaborative project to develop a fundamental model for sulfate solubility in nuclear waste glass. The solubility of sulfate has a significant impact on the achievable waste loading for nuclear waste forms within the DOE complex. These wastes can contain relatively high concentrations of sulfate, which has low solubility in borosilicate glass. This is a significant issue for low-activity waste (LAW) glass and is projected to have a major impact on the Hanford Tank Waste Treatment and Immobilization Plant (WTP). Sulfate solubility has also been a limiting factor for recent high level waste (HLW) sludge processed at the Savannah River Site (SRS) Defense Waste Processing Facility (DWPF). The low solubility of sulfate in glass, along with melter and off-gas corrosion constraints, dictate that the waste be blended with lower sulfate concentration waste sources or washed to remove sulfate prior to vitrification. The development of enhanced borosilicate glass compositions with improved sulfate solubility will allow for higher waste loadings and accelerate mission completion.The objective of the current scope being pursued by SHU is to mature the sulfate solubility model to the point where it can be used to guide glass composition development for DWPF and WTP, allowing for enhanced waste loadings and waste throughput at these facilities. A series of targeted glass compositions was selected to resolve data gaps in the model and is identified as Stage III. SHU fabricated these glasses and sent samples to SRNL for chemical composition analysis. SHU will use the resulting data to enhance the sulfate solubility model and resolve any deficiencies. In this report, SRNL provides chemical analyses for the Stage III, simulated HLW glasses fabricated by SHU in support of the sulfate solubility model development.

  4. Concept of grouping in partitioning of HLW for self-consistent fuel cycle

    International Nuclear Information System (INIS)

    Kitamoto, A.; Mulyanto

    1993-01-01

    A concept of grouping for partitioning of HLW has been developed in order to examine the possibility of a self-consistent fuel recycle. The concept of grouping of radionuclides is proposed herein, such as Group MA1 (MA below Cm), Group MA2 (Cm and higher MA), Group A ( 99 Tc and I), Group B (Cs and Sr) and Group R (the partitioned remain of HLW). Group B is difficult to be transmuted by neutron reaction, so a radiation application in an industrial scale should be developed in the future. Group A and Group MA1 can be burned by a thermal reactor, on the other hand Group MA2 should be burned by a fast reactor. P-T treatment can be optimized for the in-core and out-core system, respectively

  5. Study on the properties of Gaomiaozi bentonite as the buffer/backfilling materials for HLW disposal

    International Nuclear Information System (INIS)

    Liu Xiaodong; Luo Taian; Zhu Guoping; Chen Qingchun

    2007-12-01

    Systematic studies including mineral composition and structure, physico- chemical properties and thermal properties have been conducted on Gaomiaozi bentonite, Xinghe County, Inner Mongolia Autonomous Region. The compaction characteristics of bentonite and the influence of additive to bentonite have been discussed. The analysis of mineral composition and structure show that the bentonite ores are dominated by montmorillonite. Preliminary studies of the characteristics of ores indicated that No-type bentonite from the deposit has good absorption, excellent swelling and high cation exchangeability. The compressibility of bentonite will be improved by adding the additives such as quartz sand. The studies indicated that the characteristics of Gaomiaozi bentonite can satisfy the requirement of buffer/backfilling materials for HLW repository and the ores can be selected as the preferential candidate to provide buffer/backfill- ing materials for HLW repository in China. (authors)

  6. The interpretation of remote sensing image on the stability of fault zone at HLW repository site

    International Nuclear Information System (INIS)

    Liu Linqing; Yu Yunxiang

    1994-01-01

    It is attempted to interpret the buried fault at the preselected HLW repository site in western Gansu province with a remote sensing image. The authors discuss the features of neotectonism of Shule River buried fault zone and its two sides in light of the remote sensing image, geomorphology, stream pattern, type and thickness difference of Quaternary sediments, and structural basin, etc.. The stability of Shule River fault zone is mainly dominated by the neotectonic movement pattern and strength of its two sides. Although there exist normal and differential vertical movements along it, their strengths are small. Therefore, this is a weakly-active passive fault zone. The east Beishan area north to Shule River fault zone is weakliest active and is considered as the target for further pre-selection for HLW repository site

  7. Study on the properties of Gaomiaozi bentonite as the buffer/backfilling materials for HLW disposal

    Energy Technology Data Exchange (ETDEWEB)

    Xiaodong, Liu [East China Inst. of Technology, Fuzhou (China); [Key Laboratory of Nuclear Resources and Environment of Ministry of Education, Fuzhou (China); Taian, Luo; Guoping, Zhu; Qingchun, Chen [East China Inst. of Technology, Fuzhou (China)

    2007-12-15

    Systematic studies including mineral composition and structure, physico- chemical properties and thermal properties have been conducted on Gaomiaozi bentonite, Xinghe County, Inner Mongolia Autonomous Region. The compaction characteristics of bentonite and the influence of additive to bentonite have been discussed. The analysis of mineral composition and structure show that the bentonite ores are dominated by montmorillonite. Preliminary studies of the characteristics of ores indicated that No-type bentonite from the deposit has good absorption, excellent swelling and high cation exchangeability. The compressibility of bentonite will be improved by adding the additives such as quartz sand. The studies indicated that the characteristics of Gaomiaozi bentonite can satisfy the requirement of buffer/backfilling materials for HLW repository and the ores can be selected as the preferential candidate to provide buffer/backfill- ing materials for HLW repository in China. (authors)

  8. Evaluation Of The Integrated Solubility Model, A Graded Approach For Predicting Phase Distribution In Hanford Tank Waste

    International Nuclear Information System (INIS)

    Pierson, Kayla L.; Belsher, Jeremy D.; Seniow, Kendra R.

    2012-01-01

    The mission of the DOE River Protection Project (RPP) is to store, retrieve, treat and dispose of Hanford's tank waste. Waste is retrieved from the underground tanks and delivered to the Waste Treatment and Immobilization Plant (WTP). Waste is processed through a pretreatment facility where it is separated into low activity waste (LAW), which is primarily liquid, and high level waste (HLW), which is primarily solid. The LAW and HLW are sent to two different vitrification facilities and glass canisters are then disposed of onsite (for LAW) or shipped off-site (for HLW). The RPP mission is modeled by the Hanford Tank Waste Operations Simulator (HTWOS), a dynamic flowsheet simulator and mass balance model that is used for mission analysis and strategic planning. The integrated solubility model (ISM) was developed to improve the chemistry basis in HTWOS and better predict the outcome of the RPP mission. The ISM uses a graded approach to focus on the components that have the greatest impact to the mission while building the infrastructure for continued future improvement and expansion. Components in the ISM are grouped depending upon their relative solubility and impact to the RPP mission. The solubility of each group of components is characterized by sub-models of varying levels of complexity, ranging from simplified correlations to a set of Pitzer equations used for the minimization of Gibbs Energy

  9. Evaluation Of The Integrated Solubility Model, A Graded Approach For Predicting Phase Distribution In Hanford Tank Waste

    Energy Technology Data Exchange (ETDEWEB)

    Pierson, Kayla L.; Belsher, Jeremy D.; Seniow, Kendra R.

    2012-10-19

    The mission of the DOE River Protection Project (RPP) is to store, retrieve, treat and dispose of Hanford's tank waste. Waste is retrieved from the underground tanks and delivered to the Waste Treatment and Immobilization Plant (WTP). Waste is processed through a pretreatment facility where it is separated into low activity waste (LAW), which is primarily liquid, and high level waste (HLW), which is primarily solid. The LAW and HLW are sent to two different vitrification facilities and glass canisters are then disposed of onsite (for LAW) or shipped off-site (for HLW). The RPP mission is modeled by the Hanford Tank Waste Operations Simulator (HTWOS), a dynamic flowsheet simulator and mass balance model that is used for mission analysis and strategic planning. The integrated solubility model (ISM) was developed to improve the chemistry basis in HTWOS and better predict the outcome of the RPP mission. The ISM uses a graded approach to focus on the components that have the greatest impact to the mission while building the infrastructure for continued future improvement and expansion. Components in the ISM are grouped depending upon their relative solubility and impact to the RPP mission. The solubility of each group of components is characterized by sub-models of varying levels of complexity, ranging from simplified correlations to a set of Pitzer equations used for the minimization of Gibbs Energy.

  10. Application of intelligence based uncertainty analysis for HLW disposal

    International Nuclear Information System (INIS)

    Kato, Kazuyuki

    2003-01-01

    Safety assessment for geological disposal of high level radioactive waste inevitably involves factors that cannot be specified in a deterministic manner. These are namely: (1) 'variability' that arises from stochastic nature of the processes and features considered, e.g., distribution of canister corrosion times and spatial heterogeneity of a host geological formation; (2) 'ignorance' due to incomplete or imprecise knowledge of the processes and conditions expected in the future, e.g., uncertainty in the estimation of solubilities and sorption coefficients for important nuclides. In many cases, a decision in assessment, e.g., selection among model options or determination of a parameter value, is subjected to both variability and ignorance in a combined form. It is clearly important to evaluate both influences of variability and ignorance on the result of a safety assessment in a consistent manner. We developed a unified methodology to handle variability and ignorance by using probabilistic and possibilistic techniques respectively. The methodology has been applied to safety assessment of geological disposal of high level radioactive waste. Uncertainties associated with scenarios, models and parameters were defined in terms of fuzzy membership functions derived through a series of interviews to the experts while variability was formulated by means of probability density functions (pdfs) based on available data set. The exercise demonstrated applicability of the new methodology and, in particular, its advantage in quantifying uncertainties based on expert's opinion and in providing information on dependence of assessment result on the level of conservatism. In addition, it was also shown that sensitivity analysis could identify key parameters in reducing uncertainties associated with the overall assessment. The above information can be used to support the judgment process and guide the process of disposal system development in optimization of protection against

  11. Current status and future plans of R and D on geological disposal of HLW in Japan

    International Nuclear Information System (INIS)

    Sasaki, Noriaki

    1994-01-01

    As to the final disposal of HLW, it is considered highly important to provide a clear distinction between implementation of disposal and the research and development as independent processes, and to increase the transparency of the overall disposal program by defining concrete schedules and the roles and responsibilities of the organizations involved. The Power Reactor and Nuclear Fuel Development Corporation (PNC) has being conducted research and development on the geological disposal of HLW, as the leading organization. The responsibility of PNC is to ensure smooth progress of research and development project and to carry out studies of geological environment. The role of the Japanese government is to take overall responsibilities for appropriate and steady implementations of the program, as well as enacting any laws or policies required. On the other hand, electricity supply utilities are responsible to secure necessary funds for disposal, and in accordance with their role as waste producers, they are expected to cooperate even at the stage of research and development. Fundamental features of research and development of PNC carried out at this stage are as follows; (1) Generic research and development, (2) To establish scientific and technical bases of geological isolation of HLW in Japan, (3) About 15 years program from 1989 with documentation of progress reports, (4) Approach from near-field to far-field. PNC summarized the findings obtained by 1991, and submitted a document (H3 Report) in September 1992 as the first progress report. H3 Report is the first and comprehensive technical report on geological disposal of HLW in Japan, and provides information for the public to find out the current status of the research and development. This paper reviews the conclusions of H3 Report, overall procedures and schedule for implementing geological disposal, and future plans of R and D in PNC. (J.P.N.)

  12. Grouping in partitioning of HLW for burning and/or transmutation with nuclear reactors

    International Nuclear Information System (INIS)

    Kitamoto, Asashi; Mulyanto.

    1995-01-01

    A basic concept on partitioning and transmutation treatment by neutron reaction was developed in order to improve the waste management and the disposal scenario of high level waste (HLW). The grouping in partitioning was important factor and closely linked with the characteristics of B/T (burning and/or transmutation) treatment. The selecting and grouping concept in partitioning of HLW was proposed herein, such as Group MA1 (Np, Am, and unrecovered U and Pu), Group MA2 (Cm, Cf etc.), Group A (Tc and I), Group B (Cs and Sr) and Group R (the partitioned remain of HLW), judging from the three criteria for B/T treatment proposed in this study, which is related to (1) the value of hazard index for long-term tendency based on ALI, (2) the relative dose factor related to the mobility or retardation in ground water penetrated through geologic layer, and (3) burning and/or transmutation characteristics for recycle B/T treatment and the decay acceleration ratio by neutron reaction. Group MA1 and Group A could be burned effectively by thermal B/T reactor. Group MA2 could be burned effectively by fast B/T reactor. Transmutation of Group B by neutron reaction is difficult, therefore the development of radiation application of Group B (Cs and Sr) in industrial scale may be an interesting option in the future. Group R, i.e. the partitioned remains of HLW, and also a part of Group B should be immobilized and solidified by the glass matrix. HI ALI , the hazard index based on ALI, due to radiotoxicity of Group R can be lower than HI ALI due to standard mill tailing (smt) or uranium ore after about 300 years. (author)

  13. Development of a Korean Reference disposal System(A-KRS) for the HLW from Advanced Fuel Cycles

    International Nuclear Information System (INIS)

    Choi, Heui Joo; Choi, J. W.; Lee, J. Y.

    2010-04-01

    A database program for analyzing the characteristics of spent fuels was developed, and A-SOURCE program for characterizing the source term of HLW from advanced fuel cycles. A new technique for developing a copper canister by introducing a cold spray technique was developed, which could reduce the amount of copper. Also, to enhance the performance of A-KRS, two kinds of properties, thermal performance and iodine adsorption, were studied successfully. A complex geological disposal system which can accommodate all the HLW (CANDU and HANARO spent fuels, HLW from pyro-processing of PWR spent fuels, decommissioning wastes) was developed, and a conceptual design was carried out. Operational safety assessment system was constructed for the long-term management of A-KRS. Three representative accidental cases were analyzed, and the probabilistic safety assessment was adopted as a methodology for the safety evaluation of A-KRS operation. A national program was proposed to support the HLW national policy on the HLW management. A roadmap for HLW management was proposed based on the optimum timing of disposal

  14. HLW Melter Control Strategy Without Visual Feedback VSL-12R2500-1 Rev 0

    Energy Technology Data Exchange (ETDEWEB)

    Kruger, A A. [Department of Energy, Office of River Protection, Richland, Washington (United States); Joseph, Innocent [The Catholic University of America, Washington, DC (United States); Matlack, Keith S. [The Catholic University of America, Washington, DC (United States); Callow, Richard A. [The Catholic University of America, Washington, DC (United States); Abramowitz, Howard [The Catholic University of America, Washington, DC (United States); Pegg, Ian L. [The Catholic University of America, Washington, DC (United States); Brandys, Marek [The Catholic University of America, Washington, DC (United States); Kot, Wing K. [The Catholic University of America, Washington, DC (United States)

    2012-11-13

    Plans for the treatment of high level waste (HL W) at the Hanford Tank Waste Treatment and Immobilization Plant (WTP) are based upon the inventory of the tank wastes, the anticipated performance of the pretreatment processes, and current understanding of the capability of the borosilicate glass waste form [I]. The WTP HLW melter design, unlike earlier DOE melter designs, incorporates an active glass bubbler system. The bubblers create active glass pool convection and thereby improve heat and mass transfer and increase glass melting rates. The WTP HLW melter has a glass surface area of 3.75 m{sup 2} and depth of ~ 1.1 m. The two melters in the HLW facility together are designed to produce up to 7.5 MT of glass per day at 100% availability. Further increases in HL W waste processing rates can potentially be achieved by increasing the melter operating temperature above 1150°C and by increasing the waste loading in the glass product. Increasing the waste loading also has the added benefit of decreasing the number of canisters for storage.

  15. The use of mineral-like matrices for hlw solidification and spent fuel immobilization

    International Nuclear Information System (INIS)

    Pokhitonov, J.A.; Starchenko, V.A.; Strelnikov, A.V.; Sorokin, V.T.; Shvedov, A.A.

    2000-01-01

    The conception of radioactive waste management is based upon the multi-barrier protection principle stating that the long-lived radionuclides safety isolation is ensured by a system of engineering and natural geological barriers. One of the effective ways of the long-lived radionuclides immobilization is the integration of these materials within a mineral-like matrice. This technique may be used both for isolation of separated groups of nuclides (Cs, Sr, TUE, TRE) and for immobilization of spent fuel which for some reason can't be processed at the radiochemical plant. In this paper two variants of flowsheets HLW management are discussed. The following ways of HLW reprocessing are considered: - The first cycle raffinate solidification (without partitioning); - The individual solidification of two separated radionuclide groups (Sr+Cs+FP fraction and TPE+TRE fraction). The calcination of some characteristics (annual and total amounts, specific activity, radiochemical composition and radiogenic heat) of HLW integrated within a mineral-like matrix are performed for both options. The matrix compositions may be also used for spent fuel immobilization by means of the hot isostatic pressing technique. (authors)

  16. Integrated HLW Conceptual Process Flowsheet(s) for the Crystalline Silicotitanate Process SRDF-98-04

    International Nuclear Information System (INIS)

    Jacobs, R.A.

    1998-01-01

    The Strategic Research and Development Fund (SRDF) provided funds to develop integrated conceptual flowsheets and material balances for a CST process as a potential replacement for, or second generation to, the ITP process. This task directly supports another SRDF task: Glass Form for HLW Sludge with CST, SRDF-98-01, by M. K. Andrews which seeks to further develop sludge/CST glasses that could be used if the ITP process were replaced by CST ion exchange. The objective of the proposal was to provide flowsheet support for development and evaluation of a High Level Waste Division process to replace ITP. The flowsheets would provide a conceptual integrated material balance showing the impact on the HLW division. The evaluation would incorporate information to be developed by Andrews and Harbour on CST/DWPF glass formulations and provide the bases for evaluating the economic impact of the proposed replacement process. Coincident with this study, the Salt Disposition Team began its evaluation of alternatives for disposition of the HLW salts in the SRS waste tanks. During that time, the CST IX process was selected as one of four alternatives (of eighteen Phase II alternatives) for further evaluation during Phase III

  17. HLW Melter Control Strategy Without Visual Feedback VSL-12R2500-1 Rev 0

    International Nuclear Information System (INIS)

    Kruger, A A.; Joseph, Innocent; Matlack, Keith S.; Callow, Richard A.; Abramowitz, Howard; Pegg, Ian L.; Brandys, Marek; Kot, Wing K.

    2012-01-01

    Plans for the treatment of high level waste (HL W) at the Hanford Tank Waste Treatment and Immobilization Plant (WTP) are based upon the inventory of the tank wastes, the anticipated performance of the pretreatment processes, and current understanding of the capability of the borosilicate glass waste form [I]. The WTP HLW melter design, unlike earlier DOE melter designs, incorporates an active glass bubbler system. The bubblers create active glass pool convection and thereby improve heat and mass transfer and increase glass melting rates. The WTP HLW melter has a glass surface area of 3.75 m 2 and depth of ∼ 1.1 m. The two melters in the HLW facility together are designed to produce up to 7.5 MT of glass per day at 100% availability. Further increases in HL W waste processing rates can potentially be achieved by increasing the melter operating temperature above 1150°C and by increasing the waste loading in the glass product. Increasing the waste loading also has the added benefit of decreasing the number of canisters for storage

  18. Researches on tectonic uplift and denudation with relation to geological disposal of HLW in Japan

    International Nuclear Information System (INIS)

    Fujiwara, Osamu; Sanga, Tomoji; Moriya, Toshifumi

    2005-01-01

    This paper reviews the present state of researches on tectonic uplift and denudation, and shows perspective goals and direction of future researches from the viewpoint of geological disposal of HLW in Japan. Detailed history of tectonics and denudation in geologic time scale, including the rates, temporal and spatial distributions and processes, reconstructed from geologic and geomorphologic evidences will enable us to make the geological predictions. Improvements of the analytic methods for the geological histories, e.g. identification of the tectonic and denudational imprints and age determinations, are indispensable for the accurate prediction. Developments of the tools and methodologies for assessments of the degree and extension of influences by the tectonic uplift, subsidence and denudation on the geological environments such as ground water flows are also fundamental problem in the study field of the geological disposal of HLW. Collaboration of scientific researches using the geological and geomorphological methods and applied technology, such as numerical simulations of ground water flows, is important in improving the safety and accuracy of the geological disposal of HLW. (author)

  19. The senate working party on HLW management in Spain - historical perspective

    International Nuclear Information System (INIS)

    Lang-Lenton, J.

    2007-01-01

    As the first case history Jorge Lang Lenton, Corporate Director of ENRESA, recounted the failed attempt to establish an underground disposal facility for HLW. The site selection process, which was planned by ENRESA in the 1980's, was aimed at finding the 'technically best' site. The process was conducted by technical experts without public involvement. When 40 candidate siting areas were identified in the mid-1990's, information leaked out, creating vigorous public opposition in all of these locations. In 1998 the siting process was halted. The Senate proposed to continue R and D on geological disposal and on P and T, to reduce waste production, and to develop an energy policy that relies more on renewable energy sources. They also suggested that public participation be promoted. The 5. General Radioactive Waste Management Plan, which was developed in 1999, took these proposals into consideration. Regarding underground disposal, the government postponed any decision until 2010. At the end of 2004 a decision was made by Parliament to establish a centralized storage facility for HLW. Mr. Lang-Lenton highlighted the main lessons of the failed siting attempt. First, it has to be acknowledged that HLW management is a societal rather than a technical problem. Second, for any radioactive waste management facility a socially feasible rather than a technically optimal site should be selected, i.e., 'the best site is the possible site'. Finally, transparency and openness are needed for building confidence in the decision-making process. (author)

  20. Gas solubilities widespread applications

    CERN Document Server

    Gerrard, William

    1980-01-01

    Gas Solubilities: Widespread Applications discusses several topics concerning the various applications of gas solubilities. The first chapter of the book reviews Henr's law, while the second chapter covers the effect of temperature on gas solubility. The third chapter discusses the various gases used by Horiuti, and the following chapters evaluate the data on sulfur dioxide, chlorine data, and solubility data for hydrogen sulfide. Chapter 7 concerns itself with solubility of radon, thoron, and actinon. Chapter 8 tackles the solubilities of diborane and the gaseous hydrides of groups IV, V, and

  1. Advances in Glass Formulations for Hanford High-Aluminum, High-Iron and Enhanced Sulphate Management in HLW Streams - 13000

    Energy Technology Data Exchange (ETDEWEB)

    Kruger, Albert A. [WTP Engineering Division, United States Department of Energy, Office of River Protection, Post Office Box 450, Richland, Washington 99352 (United States)

    2013-07-01

    The current estimates and glass formulation efforts have been conservative in terms of achievable waste loadings. These formulations have been specified to ensure that the glasses are homogenous, contain essentially no crystalline phases, are processable in joule-heated, ceramic-lined melters and meet Hanford Tank Waste Treatment and Immobilization Plant (WTP) Contract terms. The WTP's overall mission will require the immobilization of tank waste compositions that are dominated by mixtures of aluminum (Al), chromium (Cr), bismuth (Bi), iron (Fe), phosphorous (P), zirconium (Zr), and sulphur (S) compounds as waste-limiting components. Glass compositions for these waste mixtures have been developed based upon previous experience and current glass property models. Recently, DOE has initiated a testing program to develop and characterize HLW glasses with higher waste loadings and higher throughput efficiencies. Results of this work have demonstrated the feasibility of increases in waste loading from about 25 wt% to 33-50 wt% (based on oxide loading) in the glass depending on the waste stream. In view of the importance of aluminum limited waste streams at Hanford (and also Savannah River), the ability to achieve high waste loadings without adversely impacting melt rates has the potential for enormous cost savings from reductions in canister count and the potential for schedule acceleration. Consequently, the potential return on the investment made in the development of these enhancements is extremely favorable. Glass composition development for one of the latest Hanford HLW projected compositions with sulphate concentrations high enough to limit waste loading have been successfully tested and show tolerance for previously unreported tolerance for sulphate. Though a significant increase in waste loading for high-iron wastes has been achieved, the magnitude of the increase is not as substantial as those achieved for high-aluminum, high-chromium, high-bismuth or

  2. Corrosion behaviour of the WAK-HLW glass

    International Nuclear Information System (INIS)

    Grambow, B.; Luckscheiter, B.; Nesovic, M.

    1997-01-01

    Sorption studies were performed on corrosion products from the glass GP WAK1 formed over a period of 40 days in deionized water at 80 C and S/V=1000 m -1 . After 40 days the pH of the solution was adjusted to various preselected values in the pH range 2-10. The pH was kept constant during the experiments by daily addition of either HNO 3 or NaOH. The sorption experiments were run at ambient temperature and 80 C for up to 10 days using various starting concentrations of Eu, Th and U. Sorption isotherms of Eu, Th and U(VI) on corrosion products were determined in deionized water, in NaCl-rich and MgCl 2 -rich solution. Presently, data of the sorption studies in deionized water are available.Furthermore the investigations of the pH dependence of saturation concentration of silica and of the release of various glass constituent of the glass GP WAK1 were continued with studies in the MgCl 2 -rich solution 1 at 80 C. Results of these studies (30 days) are given in terms of normalized elemental mass losses. (MM)

  3. Pure Phase Solubility Limits: LANL

    International Nuclear Information System (INIS)

    C. Stockman

    2001-01-01

    , complex stability constants, and redox potentials for radionuclides in different oxidation states, form the underlying database to be used for those calculations. The potentially low solubilities of many radionuclides in natural waters constitute the first barrier for their migration from the repository into the environment. Evaluation of this effect requires a knowledge of the site-specific water chemistry and the expected spatial and temporal ranges of its variability. Quantitative determinations of radionuclide solubility in waters within the range of chemistry must be made. Speciation and molecular complexation must be ascertained to interpret and apply solubility results. The solubilities thus determined can be used to assess the effectiveness of solubility in limiting radionuclide migration. These solubilities can also be used to evaluate the effectiveness of other retardation processes expected to occur once dissolution of the source material and migration begin. Understanding the solubility behavior of radionuclides will assist in designing valuable sorption experiments that must be conducted below the solubility limit since only soluble species participate in surface reactions and sorption processes. The present strategy for radionuclide solubility tasks has been to provide a solubility model from bulk-experiments that attempt to bracket the estimate made for this Analysis and Modeling Report (AMR) of water conditions on site. The long-term goal must be to develop a thermodynamic database for solution speciation and solid-state determination as a prerequisite for transport calculations and interpretation of empirical solubility data. The model has to be self-consistent and tested against known solubility studies in order to predict radionuclide solubilities over the continuous distribution ranges of potential water compositions for performance assessment of the site. Solubility studies upper limits for radionuclide concentrations in natural waters. The

  4. Pure Phase Solubility Limits: LANL

    Energy Technology Data Exchange (ETDEWEB)

    C. Stockman

    2001-01-26

    products, complex stability constants, and redox potentials for radionuclides in different oxidation states, form the underlying database to be used for those calculations. The potentially low solubilities of many radionuclides in natural waters constitute the first barrier for their migration from the repository into the environment. Evaluation of this effect requires a knowledge of the site-specific water chemistry and the expected spatial and temporal ranges of its variability. Quantitative determinations of radionuclide solubility in waters within the range of chemistry must be made. Speciation and molecular complexation must be ascertained to interpret and apply solubility results. The solubilities thus determined can be used to assess the effectiveness of solubility in limiting radionuclide migration. These solubilities can also be used to evaluate the effectiveness of other retardation processes expected to occur once dissolution of the source material and migration begin. Understanding the solubility behavior of radionuclides will assist in designing valuable sorption experiments that must be conducted below the solubility limit since only soluble species participate in surface reactions and sorption processes. The present strategy for radionuclide solubility tasks has been to provide a solubility model from bulk-experiments that attempt to bracket the estimate made for this Analysis and Modeling Report (AMR) of water conditions on site. The long-term goal must be to develop a thermodynamic database for solution speciation and solid-state determination as a prerequisite for transport calculations and interpretation of empirical solubility data. The model has to be self-consistent and tested against known solubility studies in order to predict radionuclide solubilities over the continuous distribution ranges of potential water compositions for performance assessment of the site. Solubility studies upper limits for radionuclide concentrations in natural waters. The

  5. Optimization of Deep Borehole Systems for HLW Disposal

    International Nuclear Information System (INIS)

    Driscoll, Michael; Baglietto, Emilio; Buongiorno, Jacopo; Lester, Richard; Brady, Patrick; Arnold, B. W.

    2015-01-01

    This is the final report on a project to update and improve the conceptual design of deep boreholes for high level nuclear waste disposal. The effort was concentrated on application to intact US legacy LWR fuel assemblies, but conducted in a way in which straightforward extension to other waste forms, host rock types and countries was preserved. The reference fuel design version consists of a vertical borehole drilled into granitic bedrock, with the uppermost kilometer serving as a caprock zone containing a diverse and redundant series of plugs. There follows a one to two kilometer waste canister emplacement zone having a hole diameter of approximately 40-50 cm. Individual holes are spaced 200-300 m apart to form a repository field. The choice of verticality and the use of a graphite based mud as filler between the waste canisters and the borehole wall liner was strongly influenced by the expectation that retrievability would continue to be emphasized in US and worldwide repository regulatory criteria. An advanced version was scoped out using zinc alloy cast in place to fill void space inside a disposal canister and its encapsulated fuel assembly. This excludes water and greatly improves both crush resistance and thermal conductivity. However the simpler option of using a sand fill was found adequate and is recommended for near-term use. Thermal-hydraulic modeling of the low permeability and porosity host rock and its small (@@@ 1%) saline water content showed that vertical convection induced by the waste's decay heat should not transport nuclides from the emplacement zone up to the biosphere atop the caprock. First order economic analysis indicated that borehole repositories should be cost-competitive with shallower mined repositories. It is concluded that proceeding with plans to drill a demonstration borehole to confirm expectations, and to carry out priority experiments, such as retention and replenishment of in-hole water is in order.

  6. Optimization of Deep Borehole Systems for HLW Disposal

    Energy Technology Data Exchange (ETDEWEB)

    Driscoll, Michael [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Baglietto, Emilio [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Buongiorno, Jacopo [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Lester, Richard [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Brady, Patrick [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Arnold, B. W. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-09-09

    This is the final report on a project to update and improve the conceptual design of deep boreholes for high level nuclear waste disposal. The effort was concentrated on application to intact US legacy LWR fuel assemblies, but conducted in a way in which straightforward extension to other waste forms, host rock types and countries was preserved. The reference fuel design version consists of a vertical borehole drilled into granitic bedrock, with the uppermost kilometer serving as a caprock zone containing a diverse and redundant series of plugs. There follows a one to two kilometer waste canister emplacement zone having a hole diameter of approximately 40-50 cm. Individual holes are spaced 200-300 m apart to form a repository field. The choice of verticality and the use of a graphite based mud as filler between the waste canisters and the borehole wall liner was strongly influenced by the expectation that retrievability would continue to be emphasized in US and worldwide repository regulatory criteria. An advanced version was scoped out using zinc alloy cast in place to fill void space inside a disposal canister and its encapsulated fuel assembly. This excludes water and greatly improves both crush resistance and thermal conductivity. However the simpler option of using a sand fill was found adequate and is recommended for near-term use. Thermal-hydraulic modeling of the low permeability and porosity host rock and its small (≤ 1%) saline water content showed that vertical convection induced by the waste’s decay heat should not transport nuclides from the emplacement zone up to the biosphere atop the caprock. First order economic analysis indicated that borehole repositories should be cost-competitive with shallower mined repositories. It is concluded that proceeding with plans to drill a demonstration borehole to confirm expectations, and to carry out priority experiments, such as retention and replenishment of in-hole water is in order.

  7. HIGH ALUMINUM HLW (HIGH LEVEL WASTE) GLASSES FOR HANFORD'S WTP (WASTE TREATMENT PROJECT)

    International Nuclear Information System (INIS)

    Kruger, A.A.; Bowan, B.W.; Joseph, I.; Gan, H.; Kot, W.K.; Matlack, K.S.; Pegg, I.L.

    2010-01-01

    This paper presents the results of glass formulation development and melter testing to identify high waste loading glasses to treat high-Al high level waste (HLW) at Hanford. Previous glass formulations developed for this HLW had high waste loadings but their processing rates were lower that desired. The present work was aimed at improving the glass processing rate while maintaining high waste loadings. Glass formulations were designed, prepared at crucible-scale and characterized to determine their properties relevant to processing and product quality. Glass formulations that met these requirements were screened for melt rates using small-scale tests. The small-scale melt rate screening included vertical gradient furnace (VGF) and direct feed consumption (DFC) melter tests. Based on the results of these tests, modified glass formulations were developed and selected for larger scale melter tests to determine their processing rate. Melter tests were conducted on the DuraMelter 100 (DMIOO) with a melt surface area of 0.11 m 2 and the DuraMelter 1200 (DMI200) HLW Pilot Melter with a melt surface area of 1.2 m 2 . The newly developed glass formulations had waste loadings as high as 50 wt%, with corresponding Al 2 O 3 concentration in the glass of 26.63 wt%. The new glass formulations showed glass production rates as high as 1900 kg/(m 2 .day) under nominal melter operating conditions. The demonstrated glass production rates are much higher than the current requirement of 800 kg/(m 2 .day) and anticipated future enhanced Hanford Tank Waste Treatment and Immobilization Plant (WTP) requirement of 1000 kg/(m 2 .day).

  8. Cost effects of Cu powder and bentonite on the disposal costs of an HLW repository in

    International Nuclear Information System (INIS)

    Kim, Sung Ki; Lee, Min Soo; Lee, Jong Youl; Choi, Heui Joo; Choi, Jong Won

    2008-01-01

    This paper provides the cost effect results of Cu powder and bentonite on the disposal cost for an HLW repository in Korea. In the cost analysis for both of these cost drivers, the price of Cu powder and the bentonite can affect the canister cost and the bentonite cost of the disposal holes as well as backfilling cost of the tunnels, respectively. Finally, we found that the unit cost of Cu and bentonite was the dominant cost drivers for the surface and underground facilities of an HLW repository. Therefore, an optimization of a canister and the layout of a disposal hole and disposal tunnels are essential to decrease the direct disposal cost of spent fuels. The disposal costs can be largely divided into two parts such as a surface facilities' cost and an underground facilities' cost. According to the KRS' cost analysis, the encapsulation material as well as the buffering and backfilling cost were the significant costs. Especially, a canister's cost was approximately estimated to be more than one fourth of the overall disposal costs. So it can be estimated that the unit cost of Cu powder is an important cost diver. Because the outer shell of the canister was made of Cu powder by a cold spray coating method. In addition, the unit cost of bentonite can also affect the buffering and the backfilling costs of the disposal holes and the disposal tunnels. But, these material costs will be highly expensive and unstable due to the modernization of the developing countries. So the studies for a material cost should be continued to identify the actual cost of an HLW repository

  9. An analytical overview of the consequences of microbial activity in a Swiss HLW repository

    International Nuclear Information System (INIS)

    McKinley, I.G.; West, J.M.; Grogan, H.A.

    1985-04-01

    Microorganisms are known to be important factors in many geochemical processes and their presence can be assured throughout the envisaged Swiss type C repository for HLW. It is likely that both introduced and resident microbes will colonise the near-field even at times when ambient temperature and radiation fields are relatively high. A simple quantitative model has been developed which indicates that microbial growth in the near-field is limited by the rate of supply of chemical energy from corrosion of the canister. Microbial processes examined include biodegradation of structural and packaging materials, alteration of groundwater chemistry (Eh, pH, organic complexant concentration) and direct nuclide uptake by microorganisms. The most important effects of such organisms are likely to be enhancement of release and mobility of key nuclides due to their complexation by microbial by-product. Resident micro-organisms in the far-field could potentially act as 9 living colloids' thus enhancing nuclide transport. In the case of flow paths through shear zones (kakirites), however, any microbes capable of penetrating the surrounding weathered rock matrix would be extensively retarded. It is concluded that microbial processes are unlikely to be of significance for HLW but will be more important for low/intermediate waste types. As data requirements are similar for all waste types, results from such studies would also resolve the main uncertainties remaining for the HLW case. Key research areas are identified as characterisation of a) nutrient availability in the near-field, b) the bioenergetics of iron corrosion, c) production of organic by-products, d) nuclide sorption by organisms and e) microbial mobility in the near-and far-field

  10. Application of QA to R ampersand D support of HLW programs

    International Nuclear Information System (INIS)

    Ryder, D.E.

    1988-01-01

    Quality has always been of primary importance in the research and development (R ampersand D) environment. An organization's ability to attract funds for new or continued research is largely dependent on the quality of past performance. However, with the possible exceptions of peer reviews for fund allocation and the referee process prior to publication, past quality assurance (QA) activities were primarily informal good practices. This resulted in standards of acceptable practice that varied from organization to organization. The increasing complexity of R ampersand D projects and the increasing need for project results to be upheld outside the scientific community (i.e., lawsuits and licensing hearings) are encouraging R ampersand D organizations and their clients to adopt more formalized methods for the scientific process and to increase control over support organizations (i.e., suppliers and subcontractors). This has become especially true for R ampersand D organizations involved in the high-level (HLW) projects for a number of years. The PNL began to implement QA program requirements within a few HLW repository preliminary studies in 1978. In 1985, PNL developed a comprehensive QA program for R ampersand D activities in support of two of the proposed repository projects. This QA program was developed by the PNL QA department with a significant amount of support assistance and guidance from PNL upper management, the Basalt Waste Isolation Project (BWIP), and the Salt Repository Program Office (SPRO). The QA program has been revised to add a three-level feature and is currently being implemented on projects sponsored by the Office of Geologic Repositories (DOE/OGR), Repository Technology Program (DOE-CH), Nevada Nuclear Waste Storage Investigation (NNWSI) Project, and other HLW projects

  11. Sensitivity of Nuclide Release Behavior to Groundwater Flow in an HLW Repository

    International Nuclear Information System (INIS)

    Lee, Youn-Myoung; Hwang, Yong-Soo

    2008-01-01

    Evaluation of the dose exposure rate to human being due to long-term nuclide releases from a high-level waste repository (HLW) is of importance to meet the dose limit presented by the regulatory bodies in order to ensure the performance of a repository. During the last few years, tools by which such a dose rate to an individual can be evaluated have been developed and implemented for a practical calculation to demonstrate the suitability of an HLW repository, with the aid of commercial tools such as AMBER and GoldSim, both of which are capable of probabilistic and deterministic calculations with their convenient user interface. Recently a migration from AMBER based models to GoldSim based ones has been made in accordance with a better feature of GoldSim, which is designed to facilitate the object-oriented modules to address any specialized programs, similar to solving jig saw puzzles and shows more advantage in a detailed complex modeling over AMBER. Recently a compartment modeling approach both for a geosphere and biosphere has been mainly carried out with AMBER in KAERI, which causes a necessity for a newly devised system performance evaluation model in which geosphere and biosphere models could be coupled organically together with less conservatism in the frame of the development of a total system performance assessment modeling tool, which could be successfully done with the aid of GoldSim. Therefore, through the current study, some probabilistic results of the GoldSim approach for a normal situation that could take place in a typical HLW repository are introduced

  12. The AGP-Project conceptual design for a Spanish HLW final disposal facility

    International Nuclear Information System (INIS)

    Biurrun, E.; Engelmann, H.-J.; Huertas, F.; Ulibarri, A.

    1992-01-01

    Within the framework of the AGP Project a Conceptual Design for a HLW Final Disposal Facility to be eventually built in an underground salt formation in Spain has been developed. The AGP Project has the character of a system analysis. In the current project phase I several alternatives has been considered for different subsystems and/or components of the repository. The system variants, developed to such extent as to allow a comparison of their advantages and disadvantages, will allow the selection of a reference concept, which will be further developed to technical maturity in subsequent project phases. (author)

  13. Studies on the long-term characteristics of HLW glass under ultimate storage conditions

    International Nuclear Information System (INIS)

    Roggendorf, H.; Conradt, R.; Ostertag, R.

    1987-01-01

    This interim report deals with first results of corrosion investigations of HLW simulation glass (COGEMA glass SON 68) in quinary salt solutions of different concentrations; the aim of these investigations was to find out about the corrosion mechanism at the surface of the glass and the quantitative registration of the corrosion products. It became obvious that the surface layers developed can be easily removed and that a determination of weight losses becomes possible thereby. The corrosion rates for a test period of 30 days were determined. (RB) [de

  14. Collaboration, Automation, and Information Management at Hanford High Level Radioactive Waste (HLW) Tank Farms

    International Nuclear Information System (INIS)

    Aurah, Mirwaise Y.; Roberts, Mark A.

    2013-01-01

    Washington River Protection Solutions (WRPS), operator of High Level Radioactive Waste (HLW) Tank Farms at the Hanford Site, is taking an over 20-year leap in technology, replacing systems that were monitored with clipboards and obsolete computer systems, as well as solving major operations and maintenance hurdles in the area of process automation and information management. While WRPS is fully compliant with procedures and regulations, the current systems are not integrated and do not share data efficiently, hampering how information is obtained and managed

  15. Neptunium (IV) oxalate solubility

    International Nuclear Information System (INIS)

    Luerkens, D.W.

    1983-07-01

    The equilibrium solubility of neptunium (IV) oxalate in nitric/oxalic acid solutions was determined at 22 0 C, 45 0 C, and 60 0 C. The concentrations of nitric/oxalic acid solutions represented a wide range of free oxalate ion concentration. A mathematical solubility model was developed which is based on the formation of the known complexes of neptunium (IV) oxalate. the solubility model uses a simplified concentration parameter which is proportional to the free oxalate ion concentration. The solubility model can be used to estimate the equilibrium solubility of neptunium (IV) oxalate over a wide range of oxalic and nitric acid concentrations at each temperature

  16. Transuranium elements leaching from simulated HLW glasses in synthetic interstitial claywater

    International Nuclear Information System (INIS)

    Wang, L.

    1992-08-01

    The main objective of this Master Thesis is to measure the steady-state concentrations of Pu, Np, and Am upon the leaching of High-Level Waste Glass in two types of synthetic claywater: humic acid free and humic acid containing synthetic claywater. The synthetic claywater has a composition that is representative for the in-situ interstitial groundwater of the Boom clay formation, a potential geological repository of radioactive waste in Belgium. The steady-state concentrations of transuranium elements were measured by leaching experiments with a typical duration of 400 days. Five main conclusions are drawn from the experimental data. (1) The transuranium elements that are released from simulated High Level Waste Glass are dominantly present in the synthetic claywater solutions as colloids. These colloids are smaller than 2 nm in absence of humic acids. In the presence of humic acids however, the colloids interact with actinides (adsorb or coagulate) and form particles larger than 2 nm. Np and Am are associated with inorganic and organic colloids in the synthetic interstitial claywater solution whereas Pu forms only inorganic colloids. (2) The steady-state concentration of Pu is in good agreement with the solubility of the Pu compound PuO 2 .xH 2 O. It is therefore concluded that PuO 2 .xH 2 O is the solubility controlling phase. (3) The Pu(IV)-species are dominant in the leaching solutions. Carbonate and humic acid complexes are negligible. (4) The steady-state concentrations of Np and Am in leaching solutions were much lower than the values calculated on the basis of known thermodynamic data. This indicates that the solubility controlling phases for Np and Am were not correctly identified or that the measured Np and Am concentrations were not steady-state values. (5) Non-active glass leaching tests have indicated that no organic colloids were formed as a result of glass dissolution. (A.S.)

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

    International Nuclear Information System (INIS)

    Ebert, W. L.; Lewis, M. A.

    1999-01-01

    A glass-bonded sodalite ceramic waste form is being developed for the long-term immobilization of salt wastes that are generated during spent nuclear fuel conditioning activities. A durable waste form is prepared by hot isostatic pressing (HIP) a mixture of salt-loaded zeolite powders and glass frit. A mechanistic description of the corrosion processes is being developed to support qualification of the CWF for disposal. The initial set of characterization tests included two standard tests that have been used extensively to study the corrosion behavior of high level waste (HLW) glasses: the Material Characterization Center-1 (MCC-1) Test and the Product Consistency Test (PCT). Direct comparison of the results of tests with the reference CWF and HLW glasses indicate that the corrosion behaviors of the CWF and HLW glasses are very similar

  18. Studies on the immobilization of simulated HLW in NaTi2(PO4)3 (NTP) matrix

    International Nuclear Information System (INIS)

    Raja Madhavan, R.; Govindan Kutty, K.V.; Gandhi, A.S.

    2015-01-01

    Immobilization of high level nuclear waste (HLW) is a big challenge faced by the nuclear industry today. The HLW has to be contained and isolated from the biosphere for geological timescales. NZP family of compounds is very versatile monophasic hosts for HLW immobilization. Their crystal structure can accommodate nearly all the cations known to be present in HLW due to its open structure with voids of different size. In the present study a systematic investigation on NaTi 2 (PO 4 ) 3 belonging to the NZP family; as a potential host for HLW immobilization was carried out. A simulated HLW expected from Fast Breeder Test Reactor, India (FBTR) (150Gwd/T burnup, 1 year cooling) was used. Simulated NTP waste forms with 5, 10, 15 wt. % waste loading were prepared by employing a wet chemical method and characterized. Single phase simulated NTP waste forms with up to 5 wt.% waste loading could be prepared for samples sintered in air and above 5 wt.% waste loading, monazite phase is observed as a minor secondary phase. It was found that when sintering is done in Ar/10%H 2 , NTP matrix accepts up to 10 wt.% waste loading without formation of any second phase. From the SEM studies, it was observed that samples sintered in air as well as Ar/10%H 2 palladium segregated as a metal phase and uniformly distributed throughout the waste matrix. The elemental mapping revealed retention of some of the fission products like Ru, Mo, Cs that are volatile during sintering above 1173 K and are homogenously distributed in the matrix. (author)

  19. Phase equilibrium of (CO2 + 1-aminopropyl-3-methylimidazolium bromide + water) electrolyte system and effects of aqueous medium on CO2 solubility: Experiment and modeling

    International Nuclear Information System (INIS)

    Chen, Ying; Guo, Kaihua; Bi, Yin; Zhou, Lan

    2017-01-01

    Highlights: • Phase and chemical equilibrium data for (CO 2 + [APMIm]Br + H 2 O) electrolyte system. • A modified eNRTL model for CO 2 solubility in the amino-based IL aqueous solution. • Effects of aqueous medium on both chemical and physical dissolution of CO 2 . • The correlative coefficient, R s ∗ , for the Henry’s constant of the solution. • New parameters for the segments interaction and the chemical equilibrium constants. - Abstract: New experimental data for solubility of carbon dioxide (CO 2 ) in the aqueous solution of 1-aminopropyl-3-methylimidazolium bromide ([APMIm]Br) with four different water mass fractions (0.559, 0.645, 0.765 and 0.858) at T = (278.15–348.15) K with an interval of T = 10 K and p = (0.1237–6.9143) MPa were presented. The electrolyte nonrandom two-liquid (eNRTL) model was modified to be applicable for an ionic liquid (IL) aqueous solution system, by introducing an idle factor β to illustrate the association effect of IL molecules. A solution Henry’s constant for CO 2 solubility in the IL aqueous solution was defined by introducing a correlative coefficient R s ∗ . The vapor-liquid phase equilibrium of the [APMIm]Br-H 2 O-CO 2 ternary system was successfully calculated with the modified eNRTL model. The chemical and physical mechanisms for the ionized CO 2 formation and the molecular CO 2 dissolved in the solution were identified. The effects of aqueous medium on both chemical and physical dissolution of CO 2 in the [APMIm]Br aqueous solution were studied, and a considerable enhancement of the solubility of CO 2 with increase of the water content in the solution was observed.

  20. An Assessment of Using Vibrational Compaction of Calcined HLW and LLW in DWPF Canisters

    International Nuclear Information System (INIS)

    Yi, Yun-Bo; Amme, Robert C.; Shayer, Zeev

    2008-01-01

    Since 1963, the INEL has calcined almost 8 million gallons of liquid mixed waste and liquid high-level waste, converting it to some 1.1 million gallons of dry calcine (about 4275.0 m3), which consists of alumina-and zirconia-based calcine and zirconia-sodium blend calcine. In addition, if all existing and projected future liquid wastes are solidified, approximately 2,000 m3 of additional calcine will be produced primarily from sodium-bearing waste. Calcine is a more desirable material to store than liquid radioactive waste because it reduces volume, is much less corrosive, less chemically reactive, less mobile under most conditions, easier to monitor and more protective of human health and the environment. This paper describes the technical issue involved in the development of a feasible solution for further volume reduction of calcined nuclear waste for transportation and long term storage, using a standard DWPF canister. This will be accomplished by developing a process wherein the canisters are transported into a vibrational machine, for further volume reduction by about 35%. The random compaction experiments show that this volume reduction is achievable. The main goal of this paper is to demonstrate through computer modeling that it is feasible to use volume reduction vibrational machine without developing stress/strain forces that will weaken the canister integrity. Specifically, the paper presents preliminary results of the stress/strain analysis of the DWPF canister as a function of granular calcined height during the compaction and verifying that the integrity of the canister is not compromised. This preliminary study will lead to the development of better technology for safe compactions of nuclear waste that will have significant economical impact on nuclear waste storage and treatment. The preliminary results will guide us to find better solutions to the following questions: 1) What are the optimum locations and directions (vertical versus horizontal or

  1. Risk and uncertainty assessment for a potential HLW repository in Korea: TSPA 2006

    International Nuclear Information System (INIS)

    Hwang, Y.S.; Kang, C.H.

    2004-01-01

    KAERI has worked on the concept development on permanent disposal of HLW and its total system performance assessment since 1997. More than 36 000 MT of spent nuclear fuel from PWR and CANDU reactors is planned to be disposed of in crystalline bed-rocks. The total system performance assessment (TSPA) tools are under development. The KAERI FEP encyclopedia is actively developed to include all potential FEP suitable for Korean geo- and socio conditions. The FEPs are prioritized and then categorized to the intermediate level FEP groups. These groups become elements of the rock engineering system (RES) matrix. Then the sub-scenarios such as a container failure, groundwater migration, solute transport, etc are developed by connecting interactions between diagonal elements of the RES matrix. The full scenarios are developed from the combination of sub-scenarios. For each specific scenario, the assessment contexts and associated assessment method flow charts are developed. All information on these studies is recorded into the web based programme, FEAS (FEP to Assessment through Scenarios.) KAERI applies three basic programmes for the post closure radionuclide transport calculations; MASCOT-K, AMBER, and the new MDPSA under development. The MASCOT-K originally developed by Serco for a LLW repository has been extended extensively by KAERI to simulate release reactions such as congruent and gap releases in spent nuclear fuel. The new MDPSA code is dedicated for the probabilistic assessment of radio-nuclides in multi-dimensions of a fractured porous medium. To acquire input data for TSPA domestic experiment programmes as well as literature survey are performed. The data are stored in the Performance Assessment Input Data system (PAID.) To assure the transparency, traceability, retrievability, reproducibility, and review (T2R3) the web based KAERI QA system is developed. All tasks in TSPA are recorded under the concept of a 'Project' in this web system. Currently, FEAS, PAID

  2. Capacity of burning and transmutation reactor and grouping in partitioning of HLW in self-consistent fuel recycle

    International Nuclear Information System (INIS)

    Kitamoto, A.; Mulyanto

    1993-01-01

    The concept of capacity of B/T reactor and grouping for partitioning of HLW has been developed in order to perform self-consistent fuel recycle. The concept of grouping of radionuclides is proposed herein, such as Group MA1 (MA below Cm), Group MA2 (Cm and higher MA), Group A ( 99 Te, 129 I, and 135 Cs), Group B ( 137 Cs and 90 Sr) and Group R (the partitioned remain of HLW). In this study P-T treatment were optimized for the in-core and out-core system, respectively. (author). 7 refs., 10 figs

  3. Enhanced sludge processing of HLW: Hydrothermal oxidation of chromium, technetium, and complexants by nitrate. 1998 annual progress report

    International Nuclear Information System (INIS)

    Buelow, S.J.; Robinson, J.M.

    1998-01-01

    'The objective of this project is to develop the scientific basis for hydrothermal separation of chromium from High Level Waste (HLW) sludges. The worked is aimed at attaining a fundamental understanding of chromium speciation, oxidation/reduction and dissolution kinetics, reaction mechanisms, and transport properties under hydrothermal conditions in both simple and complex salt solutions that will ultimately lead to an efficient chromium leaching process. This report summarizes the research over the first 1.5 years of a 3 year project. The authors have examined the dissolution of chromium hydroxide using different oxidants as a function of temperature and alkalinity. The results and possible applications to HLW sludges are discussed'

  4. The Production of Advanced Glass Ceramic HLW Forms using Cold Crucible Induction Melter

    Energy Technology Data Exchange (ETDEWEB)

    Veronica J Rutledge; Vince Maio

    2013-10-01

    Cold Crucible Induction Melters (CCIMs) will favorably change how High-Level radioactive Waste (from nuclear fuel recovery) is treated in the 21st century. Unlike the existing Joule-Heated Melters (JHMs) currently in operation for the glass-based immobilization of High-Level Waste (HLW), CCIMs offer unique material features that will increase melt temperatures, increase throughput, increase mixing, increase loading in the waste form, lower melter foot prints, eliminate melter corrosion and lower costs. These features not only enhance the technology for producing HLW forms, but also provide advantageous attributes to the waste form by allowing more durable alternatives to glass. This paper discusses advantageous features of the CCIM, with emphasis on features that overcome the historical issues with the JHMs presently utilized, as well as the benefits of glass ceramic waste forms over borosilicate glass waste forms. These advantages are then validated based on recent INL testing to demonstrate a first-of-a-kind formulation of a non-radioactive ceramic-based waste form utilizing a CCIM.

  5. Time evolution of the Clay Barrier Chemistry in a HLW deep geological disposal in granite

    International Nuclear Information System (INIS)

    Font, I.; Miguel, M. J.; Juncosa, R.

    2000-01-01

    The main goal of a high level waste geological disposal is to guarantee the waste isolation from the biosphere, locking them away into very deep geological formations. The best way to assure the isolation is by means of a multiple barrier system. These barriers, in a serial disposition, should assure the confinement function of the disposal system. Two kinds of barriers are considered: natural barriers (geological formations) and engineered barriers (waste form, container and backfilling and sealing materials). Bentonite is selected as backfilling and sealing materials for HLW disposal into granite formations, due to its very low permeability and its ability to fill the remaining spaces. bentonite has also other interesting properties, such as, the radionuclide retention capacity by sorption processes. Once the clay barrier has been placed, the saturation process starts. The granite groundwater fills up the voids of the bentonite and because of the chemical interactions, the groundwater chemical composition varies. Near field processes, such as canister corrosion, waste leaching and radionuclide release, strongly depends on the water chemical composition. Bentonite pore water composition is such a very important feature of the disposal system and its determination and its evolution have great relevance in the HLW deep geological disposal performance assessment. The process used for the determination of the clay barrier pore water chemistry temporal evolution, and its influence on the performance assessment, are presented in this paper. (Author)

  6. Technology for the long-term management of defense HLW at the Idaho Chemical Processing Plant

    International Nuclear Information System (INIS)

    Staples, B.A.; Berreth, J.R.; Knecht, D.A.

    1986-01-01

    The Defense Waste Management Plan of June 1983 includes a reference plan for the long-term management of Idaho Chemical Processing Plant (ICPP) high-level waste (HLW), with a goal of disposing of the annual output in 500 canisters a year by FY-2008. Based on the current vitrification technology, the ICPP base-glass case would produce 1700 canisters per year after FY-2007. Thus, to meet the DWMP goal processing steps including fuel dissolution, waste treatment, and waste immobilization are being studied as areas where potential modifications could result in HLW volume reductions for repository disposal. It has been demonstrated that ICPP calcined wastes can be densified by hot isostatic pressing to multiphase ceramic forms of high loading and density. Conversion of waste by hot isostatic pressing to these forms has the potential of reducing the annual ICPP waste production to volumes near those of the goal of the DWMP. This report summarizes the laboratory-scale information currently available on the development of these forms

  7. Proposals of geological sites for L/ILW and HLW repositories. Geological background. Text volume

    International Nuclear Information System (INIS)

    2008-01-01

    On April 2008, the Swiss Federal Council approved the conceptual part of the Sectoral Plan for Deep Geological Repositories. The Plan sets out the details of the site selection procedure for geological repositories for low- and intermediate-level waste (L/ILW) and high-level waste (HLW). It specifies that selection of geological siting regions and sites for repositories in Switzerland will be conducted in three stages, the first one (the subject of this report) being the definition of geological siting regions within which the repository projects will be elaborated in more detail in the later stages of the Sectoral Plan. The geoscientific background is based on the one hand on an evaluation of the geological investigations previously carried out by Nagra on deep geological disposal of HLW and L/ILW in Switzerland (investigation programmes in the crystalline basement and Opalinus Clay in Northern Switzerland, investigations of L/ILW sites in the Alps, research in rock laboratories in crystalline rock and clay); on the other hand, new geoscientific studies have also been carried out in connection with the site selection process. Formulation of the siting proposals is conducted in five steps: A) In a first step, the waste inventory is allocated to the L/ILW and HLW repositories; B) The second step involves defining the barrier and safety concepts for the two repositories. With a view to evaluating the geological siting possibilities, quantitative and qualitative guidelines and requirements on the geology are derived on the basis of these concepts. These relate to the time period to be considered, the space requirements for the repository, the properties of the host rock (depth, thickness, lateral extent, hydraulic conductivity), long-term stability, reliability of geological findings and engineering suitability; C) In the third step, the large-scale geological-tectonic situation is assessed and large-scale areas that remain under consideration are defined. For the L

  8. Threshold Assessment: Definition of Acceptable Sites as Part of Site Selection for the Japanese HLW Program

    International Nuclear Information System (INIS)

    McKenna, S.A.; Wakasugi, Keiichiro; Webb, E.K.; Makino, Hitoshi; Ishihara, Yoshinao; Ijiri, Yuji; Sawada, Atsushi; Baba, Tomoko; Ishiguro, Katsuhiko; Umeki, Hiroyuki

    2000-01-01

    For the last ten years, the Japanese High-Level Nuclear Waste (HLW) repository program has focused on assessing the feasibility of a basic repository concept, which resulted in the recently published H12 Report. As Japan enters the implementation phase, a new organization must identify, screen and choose potential repository sites. Thus, a rapid mechanism for determining the likelihood of site suitability is critical. The threshold approach, described here, is a simple mechanism for defining the likelihood that a site is suitable given estimates of several critical parameters. We rely on the results of a companion paper, which described a probabilistic performance assessment simulation of the HLW reference case in the H12 report. The most critical two or three input parameters are plotted against each other and treated as spatial variables. Geostatistics is used to interpret the spatial correlation, which in turn is used to simulate multiple realizations of the parameter value maps. By combining an array of realizations, we can look at the probability that a given site, as represented by estimates of this combination of parameters, would be good host for a repository site

  9. Natural analogues for containment-providing barriers for a HLW repository in salt

    Energy Technology Data Exchange (ETDEWEB)

    Wolf, J.; Noseck, U.

    2015-06-15

    In 2005, a German research project was started to develop a novel approach to prove safety for a HLW repository in a salt formation, to refine the safety concept, to identify open scientific issues and to define necessary R&D work. This project aimed at identifying the key information for a HLW repository in salt. One important question is how this information may be best fulfilled by natural analogue studies. This question is answered by starting a review of the required key information needs of the safety case (post-closure phase) in order to assess whether or not these requirements can be supported by natural analogues information. In order to structure the review and to address the key elements of the safety concepts, three types of natural analogues are distinguished: (i) natural analogues for the integrity of the geological barrier, (ii) natural analogues for the integrity of the geotechnical barriers and (iii) natural analogues for release scenarios. For the safety case in salt type (i) and (ii) are of highest importance and are treated in this paper. The assessment documented in this paper on the one hand indicates the high potential benefit of natural analogues for a safety case in salt and on the other hand helps to focus the available human and financial resources for the safety case on the most safety-relevant aspects. (authors)

  10. Status of the safety concept and safety demonstration for an HLW repository in salt. Summary report

    Energy Technology Data Exchange (ETDEWEB)

    Bollingerfehr, W.; Buhmann, D.; Filbert, W.; and others

    2013-12-15

    Salt formations have been the preferred option as host rocks for the disposal of high level radioactive waste in Germany for more than 40 years. During this period comprehensive geological investigations have been carried out together with a broad spectrum of concept and safety related R and D work. The behaviour of an HLW repository in salt formations, particularly in salt domes, has been analysed in terms of assessment of the total system performance. This was first carried out for concepts of generic waste repositories in salt and, since 1998, for a repository concept with specific boundary conditions, taking the geology of the Gorleben salt dome as an example. Suitable repository concepts and designs were developed, the technical feasibility has been proven and operational and long-term safety evaluated. Numerical modelling is an important input into the development of a comprehensive safety case for a waste repository. Significant progress in the development of numerical tools and their application for long-term safety assessment has been made in the last two decades. An integrated approach has been used in which the repository concept and relevant scientific and engineering data are combined with the results from iterative safety assessments to increase the clarity and the traceability of the evaluation. A safety concept that takes full credit of the favourable properties of salt formations was developed in the course of the R and D project ISIBEL, which started in 2005. This concept is based on the safe containment of radioactive waste in a specific part of the host rock formation, termed the containment providing rock zone, which comprises the geological barrier, the geotechnical barriers and the compacted backfill. The future evolution of the repository system will be analysed using a catalogue of Features, Events and Processes (FEP), scenario development and numerical analysis, all of which are adapted to suit the safety concept. Key elements of the

  11. Development of quality assurance for HLW disposal R and D in KAERI

    International Nuclear Information System (INIS)

    Hwang, Y. S.; Lee, J. O.; Lee, Y. M.; Kim, S. K.; Kang, C. H.

    2001-01-01

    To assure the credibility of R and D results and to systematically and effectively perform experiments and computations for the performance assessment of high-level radioactive disposal in Korea, the total quality assurance(QA) program is under development. To effectively manage the R and D's and perform decision makings so called WEB based AQ system is proposed based on the U.S.N.R.C. 10CFR50. The current proto-type QA system shall be extended to accommodate functionalities such as QA procedures, forms, and decision-making pathways. In parallel with the QA system, the technical data management (TDM) system is also applied to get probabilistic density functions (PDF's) required for probabilistic safety assessment (PSA). So-called SNL-NRC protocol was applied to construct the PDF for solubility limits of two nuclides

  12. Solubility database for TILA-99

    Energy Technology Data Exchange (ETDEWEB)

    Vuorinen, U.; Carlsson, T. [VTT Chemical Technology, Espoo (Finland); Kulmala, S.; Hakanen, M. [Helsinki Univ. (Finland). Lab. of Radiochemistry; Ahonen, L. [Geological Survey of Finland, Espoo (Finland)

    1998-11-01

    The safety assessment of spent fuel disposal requires solubility values for several elements estimated in Finnish disposal conditions. In Finland four sites (Haestholmen, Kivetty, Olkiluoto and Romuvaara) are investigated for the disposal of spent fuel. Haestholmen and OLkiluoto are onshore sites, while Kivetty and Romuvaara are inland sites. Based on groundwater analysis and classification according to salinity at the planned disposal depth mainly fresh groundwater is encountered at Kivetty and Romuvaara, while brackish and saline water-types are met at Haestholmen and Olkiluoto. Very saline, almost brine-type water ({approx}70 g/l) has been found in the deepest parts of the investigated bedrock at one of the sites (Olkiluoto). The reference waters and conditions were chosen according to the water-types. The considered reference conditions incorporated both the near- and far-field, and both oxidizing and reducing conditions were considered. In the reference conditions, the changes in solubilities were also estimated as caused by possible variations in the pH, carbonate content and redox conditions. Uranium, which is the main component of spent fuel is dealt with in a separate report presenting the solubility of uranium and spent fuel dissolution. In this work the solubilities of all the other elements of concern (Am, Cu, Nb, Np, Pa, Pd, Pu, Ra, Se, Sn, Tc, Zr, Cm, Ni, Sr, Th, C, Cl, Cs, Fe, Ho, I, and Sm) in the safety assessment are considered. Some discussion on the corrosion of the spent fuel canister is also presented. For the estimation of solubilities of the elements in question, literature data was collected that mainly comprised experimentally measured concentrations. The sources used were spent fuel experiments, concentrations measured in solubility measurements, natural concentrations and concentrations from natural analogue sites (especially Palmottu and Hyrkkoelae in Finland) as well as the concentrations measured at the Finnish investigation sites

  13. Solubility database for TILA-99

    International Nuclear Information System (INIS)

    Vuorinen, U.; Carlsson, T.; Kulmala, S.; Hakanen, M.

    1998-11-01

    The safety assessment of spent fuel disposal requires solubility values for several elements estimated in Finnish disposal conditions. In Finland four sites (Haestholmen, Kivetty, Olkiluoto and Romuvaara) are investigated for the disposal of spent fuel. Haestholmen and OLkiluoto are onshore sites, while Kivetty and Romuvaara are inland sites. Based on groundwater analysis and classification according to salinity at the planned disposal depth mainly fresh groundwater is encountered at Kivetty and Romuvaara, while brackish and saline water-types are met at Haestholmen and Olkiluoto. Very saline, almost brine-type water (∼70 g/l) has been found in the deepest parts of the investigated bedrock at one of the sites (Olkiluoto). The reference waters and conditions were chosen according to the water-types. The considered reference conditions incorporated both the near- and far-field, and both oxidizing and reducing conditions were considered. In the reference conditions, the changes in solubilities were also estimated as caused by possible variations in the pH, carbonate content and redox conditions. Uranium, which is the main component of spent fuel is dealt with in a separate report presenting the solubility of uranium and spent fuel dissolution. In this work the solubilities of all the other elements of concern (Am, Cu, Nb, Np, Pa, Pd, Pu, Ra, Se, Sn, Tc, Zr, Cm, Ni, Sr, Th, C, Cl, Cs, Fe, Ho, I, and Sm) in the safety assessment are considered. Some discussion on the corrosion of the spent fuel canister is also presented. For the estimation of solubilities of the elements in question, literature data was collected that mainly comprised experimentally measured concentrations. The sources used were spent fuel experiments, concentrations measured in solubility measurements, natural concentrations and concentrations from natural analogue sites (especially Palmottu and Hyrkkoelae in Finland) as well as the concentrations measured at the Finnish investigation sites. The

  14. Summary of International Waste Management Programs (LLNL Input to SNL L3 MS: System-Wide Integration and Site Selection Concepts for Future Disposition Options for HLW)

    Energy Technology Data Exchange (ETDEWEB)

    Greenberg, Harris R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Blink, James A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Halsey, William G. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Sutton, Mark [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2011-08-11

    The Used Fuel Disposition Campaign (UFDC) within the Department of Energy’s Office of Nuclear Energy (DOE-NE) Fuel Cycle Technology (FCT) program has been tasked with investigating the disposal of the nation’s spent nuclear fuel (SNF) and high-level nuclear waste (HLW) for a range of potential waste forms and geologic environments. This Lessons Learned task is part of a multi-laboratory effort, with this LLNL report providing input to a Level 3 SNL milestone (System-Wide Integration and Site Selection Concepts for Future Disposition Options for HLW). The work package number is: FTLL11UF0328; the work package title is: Technical Bases / Lessons Learned; the milestone number is: M41UF032802; and the milestone title is: “LLNL Input to SNL L3 MS: System-Wide Integration and Site Selection Concepts for Future Disposition Options for HLW”. The system-wide integration effort will integrate all aspects of waste management and disposal, integrating the waste generators, interim storage, transportation, and ultimate disposal at a repository site. The review of international experience in these areas is required to support future studies that address all of these components in an integrated manner. Note that this report is a snapshot of nuclear power infrastructure and international waste management programs that is current as of August 2011, with one notable exception. No attempt has been made to discuss the currently evolving world-wide response to the tragic consequences of the earthquake and tsunami that devastated Japan on March 11, 2011, leaving more than 15,000 people dead and more than 8,000 people missing, and severely damaging the Fukushima Daiichi nuclear power complex. Continuing efforts in FY 2012 will update the data, and summarize it in an Excel spreadsheet for easy comparison and assist in the knowledge management of the study cases.

  15. Small Scale Mixing Demonstration Batch Transfer and Sampling Performance of Simulated HLW - 12307

    Energy Technology Data Exchange (ETDEWEB)

    Jensen, Jesse; Townson, Paul; Vanatta, Matt [EnergySolutions, Engineering and Technology Group, Richland, WA, 99354 (United States)

    2012-07-01

    The ability to effectively mix, sample, certify, and deliver consistent batches of High Level Waste (HLW) feed from the Hanford Double Shell Tanks (DST) to the Waste treatment Plant (WTP) has been recognized as a significant mission risk with potential to impact mission length and the quantity of HLW glass produced. At the end of 2009 DOE's Tank Operations Contractor, Washington River Protection Solutions (WRPS), awarded a contract to EnergySolutions to design, fabricate and operate a demonstration platform called the Small Scale Mixing Demonstration (SSMD) to establish pre-transfer sampling capacity, and batch transfer performance data at two different scales. This data will be used to examine the baseline capacity for a tank mixed via rotational jet mixers to transfer consistent or bounding batches, and provide scale up information to predict full scale operational performance. This information will then in turn be used to define the baseline capacity of such a system to transfer and sample batches sent to WTP. The Small Scale Mixing Demonstration (SSMD) platform consists of 43'' and 120'' diameter clear acrylic test vessels, each equipped with two scaled jet mixer pump assemblies, and all supporting vessels, controls, services, and simulant make up facilities. All tank internals have been modeled including the air lift circulators (ALCs), the steam heating coil, and the radius between the wall and floor. The test vessels are set up to simulate the transfer of HLW out of a mixed tank, and collect a pre-transfer sample in a manner similar to the proposed baseline configuration. The collected material is submitted to an NQA-1 laboratory for chemical analysis. Previous work has been done to assess tank mixing performance at both scales. This work involved a combination of unique instruments to understand the three dimensional distribution of solids using a combination of Coriolis meter measurements, in situ chord length distribution

  16. Effects of a Capital Investment and a Discount Rate on the Optimal Operational Duration of an HLW Repository

    International Nuclear Information System (INIS)

    Kim, Sung Ki; Lee, Min Soo; Choi, Heui Joo; Choi, Jong Won

    2008-01-01

    This study aims to estimate the effects of a capital investment and a discount rate on the optimal operational duration of an HLW repository. According to the previous researches of the KRS(Korea Reference System) for an HLW repository, the amounts of 7,068,200 C$K and 2,636.2 MEUR are necessary to construct and operate surface and underground facilities. Since these huge costs can be a burden to some national economies, a study for a cost optimization should be performed. So we aim to drive the dominant cost driver for an optimal operational duration. A longer operational duration may be needed to dispose of more spent fuels continuously from a nuclear power plant, or to attain a retrievability of an HLW repository at a depth of 500 m below the ground level in a stable plutonic rock body. In this sense, an extended operational duration for an HLW repository affects the overall disposal costs of a repository. In this paper, only the influence of a capital investment and a discount rate was estimated from the view of optimized economics. Because these effects must be significant factors to minimize the overall disposal costs based on minimizing the sum of operational costs and capital investments

  17. Study on systematic integration technology of design and safety assessment for HLW geological disposal. 2. Research document

    International Nuclear Information System (INIS)

    Ishihara, Yoshinao; Fukui, Hiroshi; Sagawa, Hiroshi; Matsunaga, Kenichi; Ito, Takaya; Kohanawa, Osamu; Kuwayama, Yuki

    2003-02-01

    The present study was carried out relating to basic design of the Geological Disposal Technology Integration System' that will be systematized as knowledge base for design analysis and safety assessment of HLW geological disposal system by integrating organically and hierarchically various technical information in three study field. The key conclusions are summarized as follows: (1) As referring to the current performance assessment report, the technical information for R and D program of HLW geological disposal system was systematized hierarchically based on summarized information in a suitable form between the work flow (work item) and processes/characteristic flow (process item). (2) As the result of the systematized technical information, database structure and system functions necessary for development and construction to the computer system were clarified in order to secure the relation between technical information and data set for assessment of HLW geological disposal system. (3) The control procedure for execution of various analysis code used by design and safety assessment in HLW geological disposal study was arranged possibility in construction of 'Geological Disposal Technology Integration System' after investigating the distributed computing technology. (author)

  18. Safety case development in the Japanese programme for geological disposal of HLW: Evolution in the generic stage

    International Nuclear Information System (INIS)

    Ueda, Hiroyoshi; Ishiguro, Katsuhiko; Takeuchi, Mitsuo; Fujihara, Hiroshi; Takeda, Seietsu

    2014-01-01

    In the Japanese programme for nuclear power generation, the safe management of the resulting radioactive waste, particularly vitrified high-level waste (HLW) from fuel reprocessing, has been a major concern and a focus of R and D since the late 70's. According to the specifications in a report issued by an advisory committee of the Japan Atomic Energy Commission (JAEC, 1997), the Second Progress Report on R and D for the Geological Disposal of HLW (H12 report) (JNC, 2000) was published after two decades of R and D activities and showed that disposal of HLW in Japan is feasible and can be practically implemented at sites which meet certain geological stability requirements. The H12 report supported government decisions that formed the basis of the 'Act on Final Disposal of Specified Radioactive Waste' (Final Disposal Act), which came into force in 2000. The Act specifies deep geological disposal of HLW at depths greater than 300 metres, together with a stepwise site selection process in three stages. Following the Final Disposal Act, the supporting 'Basic Policy for Final Disposal' and the 'Final Disposal Plan' were authorised in the same year. (authors)

  19. Proceedings: EPRI Workshop 2 -- Technical basis for EPA HLW disposal criteria

    International Nuclear Information System (INIS)

    Rogers, V.

    1993-03-01

    The Electric Power Research Institute (EPRI) sponsored this workshop to address the scientific and technical issues underlying the regulatory criteria, or standard, for the disposal of spent nuclear fuel, high-level radioactive waste, and transuranic waste, commonly referred to collectively as high-level waste (HLW). These regulatory criteria were originally promulgated by the US Environmental Protection Agency (EPA) in 40 CFR Part 191 in 1985. However, significant portions of the regulation were remanded by the Ninth Circuit Court of Appeals in 1987. This is the second of two workshops. Topics discussed include: gas pathway; individual and groundwater protection; human intrusion; population protection; performance; TRU conversion factors and discussions. Individual projects re processed separately for the databases

  20. KAERI Underground Research Facility (KURF) for the Demonstration of HLW Disposal Technology

    International Nuclear Information System (INIS)

    Hahn, P. S.; Cho, W. J.; Kwon, S.

    2006-01-01

    In order to dispose of high-level radioactive waste(HLW) safely in geological formations, it is necessary to assess the feasibility, safety, appropriateness, and stability of the disposal concept at an underground research site, which is constructed in the same geological formation as the host rock. In this paper, the current status of the conceptual design and the construction of a small scale URL, which is named as KURF, were described. To confirm the validity of the conceptual design of the underground facility, a geological survey including a seismic refraction survey, an electronic resistivity survey, a borehole drilling, and in situ and laboratory tests had been carried out. Based on the site characterization results, it was possible to effectively design the KURF. The construction of the KURF was started in May 2005 and the access tunnel was successfully completed in March 2006. Now the construction of the research modules is under way

  1. Interference of different ionic species on the analysis of phosphate in HLW using spectrophotometer

    International Nuclear Information System (INIS)

    Mishra, P.K.; Ghongane, D.E.; Valsala, T.P.; Sonavane, M.S.; Kulkarni, Y.; Changrani, R.D.

    2010-01-01

    During reprocessing of spent nuclear fuel by PUREX process different categories of radioactive liquid wastes like High Level (HL), Intermediate Level (IL) and Low Level (LL) are generated. Different methodologies are adopted for management of these wastes. Since PUREX solvent (30% Tri butyl phosphate-70% Normal Paraffin Hydrocarbon) undergoes chemical degradation in the highly acidic medium of dissolver solution, presence of phosphate in the waste streams is inevitable. Since higher concentrations of phosphate in the HLW streams will affect its management by vitrification, knowledge about the concentration of phosphate in the waste is essential before finalising the glass composition. Since a large number of anionic and cationic species are present in the waste, these species may interfere phosphate analysis using spectrophotometer. In the present work, the interference of different anionic and cationic species on the analysis of phosphate in waste solutions using spectrophotometer was studied

  2. Depth optimization for the Korean HLW repository System within a discontinuous and saturated granitic rock mass

    International Nuclear Information System (INIS)

    Kim, Jhin Wung; Bae, Dae Seok; Choi, Jong Won

    2005-12-01

    The present study is to evaluate the material properties of the compacted bentonite, backfill material, canister cast iron insert, and the rock mass for the Korean HLW repository system. These material properties are either measured, or taken from other countries, through the evaluation of the thermal, hydraulic, and mechanical interaction behavior of a repository. After the evaluation of the material properties, the most appropriate and economical depth as well as the layout of a single layer repository is to be recommended. Material properties used for the granitic rock mass, rock joints, PWR spent fuel, disposal canister, compacted bentonite, backfill material, and ground water are the data collected domestically, and foreign data are used for some of the data not available domestically. The repository model includes a saturated granitic rock mass with joints, PWR spent fuel in a disposal canister surrounded by compacted bentonite inside a deposition hole, and backfill material in the rest of the space within a repository cavern

  3. Technical Standards on the Safety Assessment of a HLW Repository in Other Countries

    International Nuclear Information System (INIS)

    Lee, Sung Ho; Hwang, Yong Soo

    2009-01-01

    The basic function of HLW disposal system is to prevent excessive radio-nuclides being leaked from the repository in a short time. To do this, many technical standards should be developed and established on the components of disposal system. Safety assessment of a repository is considered as one of technical standards, because it produces quantitative results of the future evolution of a repository based on a reasonably simplified model. In this paper, we investigated other countries' regulations related to safely assessment focused on the assessment period, radiation dose limits and uncertainties of the assessment. Especially, in the investigation process of the USA regulations, the USA regulatory bodies' approach to assessment period and peak dose is worth taking into account in case of a conflict between peak dose from safety assessment and limited value in regulation.

  4. HLW disposal by fission reactors; calculation of trans-mutation rate and recycle

    International Nuclear Information System (INIS)

    Mulyanto

    1997-01-01

    Transmutation of MA (Minor actinide) and LLFPS (long-lived fission products) into stable nuclide or short-lived isotopes by fission reactors seem to become an alternative technology for HLW disposal. in this study, transmutation rate and recycle calculation were developed in order to evaluate transmutation characteristics of MA and LLFPs in the fission reactors. inventory of MA and LLFPs in the transmutation reactors were determined by solving of criticality equation with 1-D cylindrical geometry of multigroup diffusion equations at the beginning of cycle (BOC). transmutation rate and burn-up was determined by solving of depletion equation. inventory of MA and LLFPs was calculated for 40 years recycle. From this study, it was concluded that characteristics of MA and LLFPs in the transmutation reactors can be evaluated by recycle calculation. by calculation of transmutation rate, performance of fission reactor for transmutation of MA or LLFPs can be discussed

  5. An Ilustrative Nuclide Release Behavior from an HLW Repository due to an Earthquake Event

    International Nuclear Information System (INIS)

    Lee, Youn-Myoung; Hwang, Yong-Soo; Choi, Jong-Won

    2008-01-01

    Program for the evaluation of a high-level waste repository which is conceptually modeled. During the last few years, programs developed with the aid of AMBER and GoldSim by which nuclide transports in the near- and far-field of a repository as well as transport through the biosphere under various normal and disruptive release scenarios could be modeled and evaluated, have been continuously demonstrated. To show its usability, as similarly done for the natural groundwater flow scheme, influence of a possible disruptive event on a nuclide release behavior from an HLW repository system caused naturally due to an earthquake has been investigated and illustrated with the newly developed GoldSim program

  6. Assessment of dose conversion factors in a generic biosphere of a Korea HLW repository

    International Nuclear Information System (INIS)

    Hwang, Y. S.; Park, J. B.; Kang, C. H.

    2002-01-01

    Radioactive species released from a waste repository migrate through engineered and natural barriers and eventually reach the biosphere. Once entered the biosphere, contaminants transport various exposure pathways and finally reach a human. In this study the full RES matrix explaining the key compartments in the biosphere and their interactions is introduced considering the characteristics of the Korean biosphere. Then the three exposure groups are identified based on the compartments of interest. The full exposure pathways and corresponding mathematical expression for mass transfer coefficients and etc are developed and applied to assess the dose conversion factors of nuclides for a specific exposure group. Dose conversion factors assessed in this study will be used for total system performance assessment of a potential Korean HLW repository

  7. Current status of preparing buffer/backfill block in HLW disposal abroad

    International Nuclear Information System (INIS)

    Yan Ming; Wang Xuewen; Zhang Huyuan

    2014-01-01

    There is an urgent need for China to commence the full-scale compaction test, resolving the preparation problem for buffer/backfill blocks when underground research laboratory project is planned for High Level Radioactive Waste (HLW) disposal. The foreign countries have some research about the preparation of buffer/backfill blocks in engineered barrier systems. The foreign research shows that installation of clay blocks with sector shape at waste pollution area is a feasible engineering method. Compacted clay blocks need to be cured in a cabinet with controlled temperature and humidity to avoid desiccation and surface powdering. A freeze mixing method, mixing powdered-ice and cooled bentonite, can be operated more easily and obtain more uniform hydration than the traditional mixing of water and bentonite. It is helpful to review and adsorb the foreign research results for the design of full-scale test of bentonite compaction. (authors)

  8. Drop Calculations of HLW Canister and Pu Can-in-Canister

    International Nuclear Information System (INIS)

    Sreten Mastilovic

    2001-01-01

    The objective of this calculation is to determine the structural response of the standard high-level waste (HLW) canister and the canister containing the cans of immobilized plutonium (Pu) (''can-in-canister'' [CIC] throughout this document) subjected to drop DBEs (design basis events) during the handling operation. The evaluated DBE in the former case is 7-m (23-ft) vertical (flat-bottom) drop. In the latter case, two 2-ft (0.61-m) corner (oblique) drops are evaluated in addition to the 7-m vertical drop. These Pu CIC calculations are performed at three different temperatures: room temperature (RT) (20 C), T = 200 F = 93.3 C , and T = 400 F = 204 C ; in addition to these the calculation characterized by the highest maximum stress intensity is performed at T = 750 F = 399 C as well. The scope of the HLW canister calculation is limited to reporting the calculation results in terms of: stress intensity and effective plastic strain in the canister, directional residual strains at the canister outer surface, and change of canister dimensions. The scope of Pu CIC calculation is limited to reporting the calculation results in terms of stress intensity, and effective plastic strain in the canister. The information provided by the sketches from Reference 26 (Attachments 5.3,5.5,5.8, and 5.9) is that of the potential CIC design considered in this calculation, and all obtained results are valid for this design only. This calculation is associated with the Plutonium Immobilization Project and is performed by the Waste Package Design Section in accordance with Reference 24. It should be noted that the 9-m vertical drop DBE, included in Reference 24, is not included in the objective of this calculation since it did not become a waste acceptance requirement. AP-3.124, ''Calculations'', is used to perform the calculation and develop the document

  9. DETERMINATION OF HLW GLASS MELT RATE USING X-RAY COMPUTED TOMOGRAPHY

    Energy Technology Data Exchange (ETDEWEB)

    Choi, A.; Miller, D.; Immel, D.

    2011-10-06

    The purpose of the high-level waste (HLW) glass melt rate study is two-fold: (1) to gain a better understanding of the impact of feed chemistry on melt rate through bench-scale testing, and (2) to develop a predictive tool for melt rate in support of the on-going frit development efforts for the Defense Waste Processing Facility (DWPF). In particular, the focus is on predicting relative melt rates, not the absolute melt rates, of various HLW glass formulations solely based on feed chemistry, i.e., the chemistry of both waste and glass-forming frit for DWPF. Critical to the successful melt rate modeling is the accurate determination of the melting rates of various HLW glass formulations. The baseline procedure being used at the Savannah River National Laboratory (SRNL) is to; (1) heat a 4 inch-diameter stainless steel beaker containing a mixture of dried sludge and frit in a furnace for a preset period of time, (2) section the cooled beaker along its diameter, and (3) measure the average glass height across the sectioned face using a ruler. As illustrated in Figure 1-1, the glass height is measured for each of the 16 horizontal segments up to the red lines where relatively large-sized bubbles begin to appear. The linear melt rate (LMR) is determined as the average of all 16 glass height readings divided by the time during which the sample was kept in the furnace. This 'visual' method has proved useful in identifying melting accelerants such as alkalis and sulfate and further ranking the relative melt rates of candidate frits for a given sludge batch. However, one of the inherent technical difficulties of this method is to determine the glass height in the presence of numerous gas bubbles of varying sizes, which is prevalent especially for the higher-waste-loading glasses. That is, how the red lines are drawn in Figure 1-1 can be subjective and, therefore, may influence the resulting melt rates significantly. For example, if the red lines are drawn too low

  10. INTEGRATED DM 1200 MELTER TESTING OF HLW C-106/AY-102 COMPOSITION USING BUBBLERS VSL-03R3800-1 REV 0 9/15/03

    Energy Technology Data Exchange (ETDEWEB)

    KRUGER AA; MATLACK KS; GONG W; BARDAKCI T; D' ANGELO NA; KOT WK; PEGG IL

    2011-12-29

    This report documents melter and off-gas performance results obtained on the DM1200 HLW Pilot Melter during processing of simulated HLW C-106/AY-102 feed. The principal objectives of the DM1200 melter testing were to determine the achievable glass production rates for simulated HLW C-106/AY-102 feed; determine the effect of bubbling rate on production rate; characterize melter off-gas emissions; characterize the performance of the prototypical off-gas system components as well as their integrated performance; characterize the feed, glass product, and off-gas effluents; and to perform pre- and post test inspections of system components.

  11. Integrated DM 1200 Melter Testing Of HLW C-106/AY-102 Composition Using Bubblers VSL-03R3800-1, Rev. 0, 9/15/03

    International Nuclear Information System (INIS)

    Kruger, A.A.; Matlack, K.S.; Kot, W.K.; Bardakci, T.; Gong, W.; D'Angelo, N.A.; Pegg, I.L.

    2011-01-01

    This report documents melter and off-gas performance results obtained on the DM1200 HLW Pilot Melter during processing of simulated HLW C-106/AY-102 feed. The principal objectives of the DM1200 melter testing were to determine the achievable glass production rates for simulated HLW C-106/AY-102 feed; determine the effect of bubbling rate on production rate; characterize melter off-gas emissions; characterize the performance of the prototypical off-gas system components as well as their integrated performance; characterize the feed, glass product, and off-gas effluents; and to perform pre- and post test inspections of system components.

  12. Evaluated and estimated solubility of some elements for performance assessment of geological disposal of high-level radioactive waste using updated version of thermodynamic database

    International Nuclear Information System (INIS)

    Kitamura, Akira; Doi, Reisuke; Yoshida, Yasushi

    2011-01-01

    Japan Atomic Energy Agency (JAEA) established the thermodynamic database (JAEA-TDB) for performance assessment of geological disposal of high-level radioactive waste (HLW) and TRU waste. Twenty-five elements which were important for the performance assessment of geological disposal were selected for the database. JAEA-TDB enhanced reliability of evaluation and estimation of their solubility through selecting the latest and the most reliable thermodynamic data at present. We evaluated and estimated solubility of the 25 elements in the simulated porewaters established in the 'Second Progress Report for Safety Assessment of Geological Disposal of HLW in Japan' using the JAEA-TDB and compared with those using the previous thermodynamic database (JNC-TDB). It was found that most of the evaluated and estimated solubility values were not changed drastically, but the solubility and speciation of dominant aqueous species for some elements using the JAEA-TDB were different from those using the JNC-TDB. We discussed about how to provide reliable solubility values for the performance assessment. (author)

  13. Solubilities of uranium for TILA-99

    International Nuclear Information System (INIS)

    Ollila, K.; Ahonen, L.

    1998-11-01

    This report presents the evaluation of the uranium solubilities in the reference waters of TILA-99. The behaviour of uranium has been discussed separately in the near-field and far-field conditions. The bentonite/groundwater interactions have been considered in the compositions of the fresh and saline near-field reference waters. The far-field groundwaters' compositions include fresh, brackish, saline and very saline, almost brine-type compositions. The pH and redox conditions, as the main parameters affecting the solubilities, are considered. A literature study was made in order to obtain information on the recent dissolution and leaching experiments of UO 2 and spent fuel. The latest literature includes studies on UO 2 solubility under anoxic conditions, in which the methods for simulating the reducing conditions of deep groundwater have been improved. Studies on natural uraninite and its alteration products give a valuable insight into the long-term behaviour of spent fuel. Also the solubility equilibria for some relevant poorly known uranium minerals have been determined. The solubilities of the selected solubility-limiting phases were calculated using the geochemical code, EQ3/6. The NEA database for uranium was the basis for the modelling. The recently extended and updated SR '97 database was used for comparison. The solubility products for uranophane were taken from the latest literature. The recommended values for solubilities were given after a comparison between the calculated solubilities, experimental information and measured concentrations in natural groundwaters. The experiments include several UO 2 dissolution studies in synthetic groundwaters with compositions close to the reference groundwaters. (author)

  14. Solubilities of uranium for TILA-99

    Energy Technology Data Exchange (ETDEWEB)

    Ollila, K. [VTT Chemical Technology, Espoo (Finland); Ahonen, L. [Geological Survey of Finland, Espoo (Finland)

    1998-11-01

    This report presents the evaluation of the uranium solubilities in the reference waters of TILA-99. The behaviour of uranium has been discussed separately in the near-field and far-field conditions. The bentonite/groundwater interactions have been considered in the compositions of the fresh and saline near-field reference waters. The far-field groundwaters` compositions include fresh, brackish, saline and very saline, almost brine-type compositions. The pH and redox conditions, as the main parameters affecting the solubilities, are considered. A literature study was made in order to obtain information on the recent dissolution and leaching experiments of UO{sub 2} and spent fuel. The latest literature includes studies on UO{sub 2} solubility under anoxic conditions, in which the methods for simulating the reducing conditions of deep groundwater have been improved. Studies on natural uraninite and its alteration products give a valuable insight into the long-term behaviour of spent fuel. Also the solubility equilibria for some relevant poorly known uranium minerals have been determined. The solubilities of the selected solubility-limiting phases were calculated using the geochemical code, EQ3/6. The NEA database for uranium was the basis for the modelling. The recently extended and updated SR `97 database was used for comparison. The solubility products for uranophane were taken from the latest literature. The recommended values for solubilities were given after a comparison between the calculated solubilities, experimental information and measured concentrations in natural groundwaters. The experiments include several UO{sub 2} dissolution studies in synthetic groundwaters with compositions close to the reference groundwaters. (author) 81 refs.

  15. Experimental programme to demonstrate the viability of the supercontainer concept for HLW

    International Nuclear Information System (INIS)

    Van Humbeeck, Hughes; De Bock, Chris; Bastiaens, Wim; Van Cotthem, Alain

    2008-01-01

    The EIG EURIDICE (a joint venture between the Belgian Organisation for Radioactive Waste Management - ONDRAF/NIRAS - and the Belgian Nuclear Research Centre - SCKoCEN) is responsible for performing large-scale tests, technical demonstrations and experiments to assess the feasibility of a final disposal of vitrified radioactive waste in deep clay layers. This is part of the Belgian Research and Development programme managed by ONDRAF/NIRAS. The current Belgian reference design for vitrified HLW and spent fuel assemblies is the so-called Supercontainer design. The vitrified waste canisters or spent fuel assemblies are enclosed in a carbon steel overpack which has to prevent contact between water from the host formation and the waste during the thermal phase. In order to maintain favourable chemical conditions to avoid corrosion during this period (several hundred or even thousand of years), the overpack is surrounded by a high alkaline concrete buffer of about 70 cm thick. The buffer also provides permanent radiological shielding for the workers, simplifying handling and other operations. All the components of the Supercontainer are constructed in above ground installations, thus creating favourable QA/QC conditions. After the emplacement of the Supercontainers in the disposal galleries, the remaining space will be backfilled. Tests to demonstrate the viability and the construction feasibility of the supercontainer design have been initiated. The viability programme includes Tests to verify the feasibility to construct and emplace the components of the supercontainers, and tests to verify the feasibility to backfill the disposal galleries once the supercontainers are placed. Supercontainer construction: Tests in column to verify the construction feasibility (risk of cracking) of the buffer with two different types of concrete (a self-compacting concrete - SCC - and a rheoplastic concrete RPC) were performed in collaboration with the Belgian concrete factory Socea. A

  16. Experience-dependent reduction of soluble β-amyloid oligomers and rescue of cognitive abilities in middle-age Ts65Dn mice, a model of Down syndrome.

    Science.gov (United States)

    Sansevero, Gabriele; Begenisic, Tatjana; Mainardi, Marco; Sale, Alessandro

    2016-09-01

    Down syndrome (DS) is the most diffused genetic cause of intellectual disability and, after the age of forty, is invariantly associated with Alzheimer's disease (AD). In the last years, the prolongation of life expectancy in people with DS renders the need for intervention paradigms aimed at improving mental disability and counteracting AD pathology particularly urgent. At present, however, there are no effective therapeutic strategies for DS and concomitant AD in mid-life people. The most intensively studied mouse model of DS is the Ts65Dn line, which summarizes the main hallmarks of the DS phenotype, included severe learning and memory deficits and age-dependent AD-like pathology. Here we report for the first time that middle-age Ts65Dn mice display a marked increase in soluble Aβ oligomer levels in their hippocampus. Moreover, we found that long-term exposure to environmental enrichment (EE), a widely used paradigm that increases sensory-motor stimulation, reduces Aβ oligomers and rescues spatial memory abilities in trisomic mice. Our findings underscore the potential of EE procedures as a non-invasive paradigm for counteracting brain aging processes in DS subjects. Copyright © 2016 Elsevier Inc. All rights reserved.

  17. Solubility of ethylene in methyl propionate

    NARCIS (Netherlands)

    Shariati - Sarabi, A.; Florusse, L.J.; Peters, C.J.

    2015-01-01

    In this work, the solubility of ethylene in methyl propionate was measured within a temperature range of 283.5–464.8 K and pressures up to 10.7 MPa. Experiments were carried out using the Cailletet apparatus, which uses a synthetic method for the experiments. The critical points of several isopleths

  18. Final Report Start-Up And Commissioning Tests On The Duramelter 1200 HLW Pilot Melter System Using AZ-101 HLW Simulants VSL-01R0100-2, Rev. 0, 1/20/03

    International Nuclear Information System (INIS)

    Kruger, A.A.; Matlack, K.S.; Kot, W.K.; Brandys, M.; Wilson, C.N.; Schatz, T.R.; Gong, W.; Pegg, I.L.

    2011-01-01

    This document provides the final report on data and results obtained from commissioning tests performed on the one-third scale DuraMelter(trademark) 1200 (DM 1200) HLW Pilot Melter system that has been installed at VSL with an integrated prototypical off-gas treatment system. That system has replaced the DM1000 system that was used for HLW throughput testing during Part BI (1). Both melters have similar melt surface areas (1.2 m 2 ) but the DM1200 is prototypical of the present RPP-WTP HLW melter design whereas the DM1000 was not. These tests were performed under a corresponding RPP-WTP Test Specification and associated Test Plan. This report is a followup to the previously issued Preliminary Data Summary Report. The DM1200 system will be used for testing and confirmation of basic design, operability, flow sheet, and process control assumptions as well as for support of waste form qualification and permitting. This will include data on processing rates, off-gas treatment system performance, recycle stream compositions, as well as process operability and reliability. Consequently, this system is a key component of the overall HLW vitrification development strategy. The results presented in this report are from the initial series of short-duration tests that were conducted to support the start-up and commissioning of this system prior to conducting the main body of development tests that have been planned for this system. These tests were directed primarily at system 'debugging,' operator training, and procedure refinement. The AZ-101 waste simulant and glass composition that was used for previous testing was selected for these tests.

  19. FINAL REPORT START-UP AND COMMISSIONING TESTS ON THE DURAMELTER 1200 HLW PILOT MELTER SYSTEM USING AZ-101 HLW SIMULANTS VSL-01R0100-2 REV 0 1/20/03

    Energy Technology Data Exchange (ETDEWEB)

    KRUGER AA; MATLACK KS; KOT WK; BRANDYS M; WILSON CN; SCHATZ TR; GONG W; PEGG IL

    2011-12-29

    This document provides the final report on data and results obtained from commissioning tests performed on the one-third scale DuraMelter{trademark} 1200 (DM 1200) HLW Pilot Melter system that has been installed at VSL with an integrated prototypical off-gas treatment system. That system has replaced the DM1000 system that was used for HLW throughput testing during Part BI [1]. Both melters have similar melt surface areas (1.2 m{sup 2}) but the DM1200 is prototypical of the present RPP-WTP HLW melter design whereas the DM1000 was not. These tests were performed under a corresponding RPP-WTP Test Specification and associated Test Plan. This report is a followup to the previously issued Preliminary Data Summary Report. The DM1200 system will be used for testing and confirmation of basic design, operability, flow sheet, and process control assumptions as well as for support of waste form qualification and permitting. This will include data on processing rates, off-gas treatment system performance, recycle stream compositions, as well as process operability and reliability. Consequently, this system is a key component of the overall HLW vitrification development strategy. The results presented in this report are from the initial series of short-duration tests that were conducted to support the start-up and commissioning of this system prior to conducting the main body of development tests that have been planned for this system. These tests were directed primarily at system 'debugging,' operator training, and procedure refinement. The AZ-101 waste simulant and glass composition that was used for previous testing was selected for these tests.

  20. Discussing compliance. Summary report from discussions with Robert Bernero and Chris Whipple regarding compliance with the Swedish HLW Regulations from meetings in Stockholm May 3 and 4, 1999

    Energy Technology Data Exchange (ETDEWEB)

    Jensen, Mikael

    1999-06-01

    Summary report from discussions with Robert Bernero and Chris Whipple regarding compliance with the Swedish HLW Regulations from meetings in Stockholm. The report also contains bibliographical information and preliminary observations made by Robert Bernero and Chris Whipple.

  1. Discussing compliance. Summary report from discussions with Robert Bernero and Chris Whipple regarding compliance with the Swedish HLW Regulations from meetings in Stockholm May 3 and 4, 1999

    International Nuclear Information System (INIS)

    Jensen, Mikael

    1999-06-01

    Summary report from discussions with Robert Bernero and Chris Whipple regarding compliance with the Swedish HLW Regulations from meetings in Stockholm. The report also contains bibliographical information and preliminary observations made by Robert Bernero and Chris Whipple

  2. Final Report - Crystal Settling, Redox, and High Temperature Properties of ORP HLW and LAW Glasses, VSL-09R1510-1, Rev. 0, dated 6/18/09

    Energy Technology Data Exchange (ETDEWEB)

    Kruger, Albert A.; Wang, C.; Gan, H.; Pegg, I. L.; Chaudhuri, M.; Kot, W.; Feng, Z.; Viragh, C.; McKeown, D. A.; Joseph, I.; Muller, I. S.; Cecil, R.; Zhao, W.

    2013-11-13

    The radioactive tank waste treatment programs at the U. S. Department of Energy (DOE) have featured joule heated ceramic melter technology for the vitrification of high level waste (HLW). The Hanford Tank Waste Treatment and Immobilization Plant (WTP) employs this same basic technology not only for the vitrification of HLW streams but also for the vitrification of Low Activity Waste (LAW) streams. Because of the much greater throughput rates required of the WTP as compared to the vitrification facilities at the West Valley Demonstration Project (WVDP) or the Defense Waste Processing Facility (DWPF), the WTP employs advanced joule heated melters with forced mixing of the glass pool (bubblers) to improve heat and mass transport and increase melting rates. However, for both HLW and LAW treatment, the ability to increase waste loadings offers the potential to significantly reduce the amount of glass that must be produced and disposed and, therefore, the overall project costs. This report presents the results from a study to investigate several glass property issues related to WTP HLW and LAW vitrification: crystal formation and settling in selected HLW glasses; redox behavior of vanadium and chromium in selected LAW glasses; and key high temperature thermal properties of representative HLW and LAW glasses. The work was conducted according to Test Plans that were prepared for the HLW and LAW scope, respectively. One part of this work thus addresses some of the possible detrimental effects due to considerably higher crystal content in waste glass melts and, in particular, the impact of high crystal contents on the flow property of the glass melt and the settling rate of representative crystalline phases in an environment similar to that of an idling glass melter. Characterization of vanadium redox shifts in representative WTP LAW glasses is the second focal point of this work. The third part of this work focused on key high temperature thermal properties of

  3. HLW Feed Delivery AZ101 Batch Transfer to the Private Contractor Transfer and Mixing Process Improvements [Initial Release at Rev 2

    Energy Technology Data Exchange (ETDEWEB)

    DUNCAN, G.P.

    2000-02-28

    The primary purpose of this business case is to provide Operations and Maintenance with a detailed transfer process review for the first High Level Waste (HLW) feed delivery to the Privatization Contractor (PC), AZ-101 batch transfer to PC. The Team was chartered to identify improvements that could be implemented in the field. A significant penalty can be invoked for not providing the quality, quantity, or timely delivery of HLW feed to the PC.

  4. Final Report - Testing of Optimized Bubbler Configuration for HLW Melter VSL-13R2950-1, Rev. 0, dated 6/12/2013

    Energy Technology Data Exchange (ETDEWEB)

    Kruger, Albert A.; Pegg, I. L.; Callow, R. A.; Joseph, I.; Matlack, K. S.; Kot, W. K.

    2013-11-13

    The principal objective of this work was to determine the glass production rate increase and ancillary effects of adding more bubbler outlets to the current WTP HLW melter baseline. This was accomplished through testing on the HLW Pilot Melter (DM1200) at VSL. The DM1200 unit was selected for these tests since it was used previously with several HLW waste streams including the four tank wastes proposed for initial processing at Hanford. This melter system was also used for the development and optimization of the present baseline WTP HLW bubbler configuration for the WTP HLW melter, as well as for MACT testing for both HLW and LAW. Specific objectives of these tests were to: Conduct DM1200 melter testing with the baseline WTP bubbling configuration and as augmented with additional bubblers. Conduct DM1200 melter testing to differentiate the effects of total bubbler air flow and bubbler distribution on glass production rate and cold cap formation. Collect melter operating data including processing rate, temperatures at a variety of locations within the melter plenum space, melt pool temperature, glass melt density, and melter pressure with the baseline WTP bubbling configuration and as augmented with additional bubblers. Collect melter exhaust samples to compare particulate carryover for different bubbler configurations. Analyze all collected data to determine the effects of adding more bubblers to the WTP HLW melter to inform decisions regarding future lid re-designs. The work used a high aluminum HLW stream composition defined by ORP, for which an appropriate simulant and high waste loading glass formulation were developed and have been previously processed on the DM1200.

  5. Safety studies of HLW-disposal in the Mors salt dome - Support to the salt option of the Pagis project

    International Nuclear Information System (INIS)

    Lindstroem Jensen, K.E.

    1987-01-01

    The study, which is a support to the Pagis project, covers three tasks concerning the evaluation of the Danish salt dome Mors (variant disposal site): evaluation of the human intrusion scenario where a cavern is excavated near the HLW-repository by solution mining technique. The waste is supposed to be leached during the operation period until the abandoned cavern is closed by convergence and the contaminated brine is pressed up into the overburden. Evaluation of the brine intrusion scenario, where the HLW-repository is inadvertently located close to a major brine pocket which subsequently releases its brine content through defects in the repository to the discharge stream for the catchment area. Collection and description of hydrological data of surface and deep layers (down to circa 700 metres) in the repository region. The data will be used by GSF to calculate the radionuclide migration in the geosphere

  6. Regional Geologic Evaluations for Disposal of HLW and SNF: The Pierre Shale of the Northern Great Plains

    Energy Technology Data Exchange (ETDEWEB)

    Perry, Frank Vinton [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Kelley, Richard E. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-09-14

    The DOE Spent Fuel and Waste Technology (SWFT) R&D Campaign is supporting research on crystalline rock, shale (argillite) and salt as potential host rocks for disposal of HLW and SNF in a mined geologic repository. The distribution of these three potential repository host rocks is limited to specific regions of the US and to different geologic and hydrologic environments (Perry et al., 2014), many of which may be technically suitable as a site for mined geologic disposal. This report documents a regional geologic evaluation of the Pierre Shale, as an example of evaluating a potentially suitable shale for siting a geologic HLW repository. This report follows a similar report competed in 2016 on a regional evaluation of crystalline rock that focused on the Superior Province of the north-central US (Perry et al., 2016).

  7. Determination of alpha dose rate profile at the HLW nuclear glass/water interface

    Energy Technology Data Exchange (ETDEWEB)

    Mougnaud, S., E-mail: sarah.mougnaud@cea.fr [CEA Marcoule, DEN/DTCD/SECM, BP 17171, 30207 Bagnols-sur-Cèze cedex (France); Tribet, M.; Rolland, S. [CEA Marcoule, DEN/DTCD/SECM, BP 17171, 30207 Bagnols-sur-Cèze cedex (France); Renault, J.-P. [CEA Saclay, NIMBE UMR 3685 CEA/CNRS, 91191 Gif-sur-Yvette cedex (France); Jégou, C. [CEA Marcoule, DEN/DTCD/SECM, BP 17171, 30207 Bagnols-sur-Cèze cedex (France)

    2015-07-15

    Highlights: • The nuclear glass/water interface is studied. • The way the energy of alpha particles is deposited is modeled using MCNPX code. • A model giving dose rate profiles at the interface using intrinsic data is proposed. • Bulk dose rate is a majoring estimation in alteration layer and in surrounding water. • Dose rate is high in small cracks; in larger ones irradiated volume is negligible. - Abstract: Alpha irradiation and radiolysis can affect the alteration behavior of High Level Waste (HLW) nuclear glasses. In this study, the way the energy of alpha particles, emitted by a typical HLW glass, is deposited in water at the glass/water interface is investigated, with the aim of better characterizing the dose deposition at the glass/water interface during water-induced leaching mechanisms. A simplified chemical composition was considered for the nuclear glass under study, wherein the dose rate is about 140 Gy/h. The MCNPX calculation code was used to calculate alpha dose rate and alpha particle flux profiles at the glass/water interface in different systems: a single glass grain in water, a glass powder in water and a water-filled ideal crack in a glass package. Dose rate decreases within glass and in water as distance to the center of the grain increases. A general model has been proposed to fit a dose rate profile in water and in glass from values for dose rate in glass bulk, alpha range in water and linear energy transfer considerations. The glass powder simulation showed that there was systematic overlapping of radiation fields for neighboring glass grains, but the water dose rate always remained lower than the bulk value. Finally, for typical ideal cracks in a glass matrix, an overlapping of irradiation fields was observed while the crack aperture was lower than twice the alpha range in water. This led to significant values for the alpha dose rate within the crack volume, as long as the aperture remained lower than 60 μm.

  8. Cavern disposal concepts for HLW/SF: assuring operational practicality and safety with maximum programme flexibility

    International Nuclear Information System (INIS)

    McKinley, Ian G.; Apted, Mick; Umeki, Hiroyuki; Kawamura, Hideki

    2008-01-01

    Most conventional engineered barrier system (EBS) designs for HLW/SF repositories are based on concepts developed in the 1970s and 1980s that assured feasibility with high margins of safety, in order to convince national decision makers to proceed with geological disposal despite technological uncertainties. In the interval since the advent of such 'feasibility designs', significant progress has been made in reducing technological uncertainties, which has lead to a growing awareness of other, equally important uncertainties in operational implementation and challenges regarding social acceptance in many new, emerging national repository programs. As indicated by the NUMO repository concept catalogue study (NUMO, 2004), there are advantages in reassessing how previous designs can be modified and optimised in the light of improved system understanding, allowing a robust EBS to be flexibly implemented to meet nation-specific and site-specific conditions. Full-scale emplacement demonstrations, particularly those carried out underground, have highlighted many of the practical issues to be addressed; e.g., handling of compacted bentonite in humid conditions, use of concrete for support infrastructure, remote handling of heavy radioactive packages in confined conditions, quality inspection, monitoring / ease of retrieval of emplaced packages and institutional control. The CAvern REtrievable (CARE) concept reduces or avoids such issues by emplacement of HLW or SF within multi-purpose transportation / storage / disposal casks in large ventilated caverns at a depth of several hundred metres. The facility allows the caverns to serve as inspectable stores for an extended period of time (up to a few hundred years) until a decision is made to close them. At this point the caverns are backfilled and sealed as a final repository, effectively with the same safety case components as conventional 'feasibility designs'. In terms of operational practicality an d safety, the CARE

  9. Development Of High Waste-Loading HLW Glasses For High Bismuth Phosphate Wastes, VSL-12R2550-1, Rev 0

    International Nuclear Information System (INIS)

    Kruger, A. A.; Pegg, Ian L.; Gan, Hao; Kot, Wing K.

    2012-01-01

    This report presents results from tests with new glass formulations that have been developed for several high Bi-P HLW compositions that are expected to be processed at the WTP that have not been tested previously. WTP HLW feed compositions were reviewed to select waste batches that are high in Bi-P and that are reasonably distinct from the Bi-limited waste that has been tested previously. Three such high Bi-P HLW compositions were selected for this work. The focus of the present work was to determine whether the same type of issues as seen in previous work with high-Bi HLW will be seen in HLW with different concentrations of Bi, P and Cr and also whether similar glass formulation development approaches would be successful in mitigating these issues. New glass compositions were developed for each of the three representative Bi-P HLW wastes and characterized with respect to key processing and product quality properties and, in particular, those relating to crystallization and foaming tendency

  10. Development Of High Waste-Loading HLW Glasses For High Bismuth Phosphate Wastes, VSL-12R2550-1, Rev 0

    Energy Technology Data Exchange (ETDEWEB)

    Kruger, A. A. [Department of Energy, Office of River Protection, Richland, Washington (United States); Pegg, Ian L. [The Catholic University of America, Washington, DC (United States); Gan, Hao [The Catholic University of America, Washington, DC (United States); Kot, Wing K. [The Catholic University of America, Washington, DC (United States)

    2012-12-13

    This report presents results from tests with new glass formulations that have been developed for several high Bi-P HLW compositions that are expected to be processed at the WTP that have not been tested previously. WTP HLW feed compositions were reviewed to select waste batches that are high in Bi-P and that are reasonably distinct from the Bi-limited waste that has been tested previously. Three such high Bi-P HLW compositions were selected for this work. The focus of the present work was to determine whether the same type of issues as seen in previous work with high-Bi HLW will be seen in HLW with different concentrations of Bi, P and Cr and also whether similar glass formulation development approaches would be successful in mitigating these issues. New glass compositions were developed for each of the three representative Bi-P HLW wastes and characterized with respect to key processing and product quality properties and, in particular, those relating to crystallization and foaming tendency.

  11. Final Report - Management of High Sulfur HLW, VSL-13R2920-1, Rev. 0, dated 10/31/2013

    Energy Technology Data Exchange (ETDEWEB)

    Kruger, Albert A.; Gan, H.; Pegg, I. L.; Feng, Z.; Gan, H; Joseph, I.; Matlack, K. S.

    2013-11-13

    The present report describes results from a series of small-scale crucible tests to determine the extent of corrosion associated with sulfur containing HLW glasses and to develop a glass composition for a sulfur-rich HLW waste stream, which was then subjected to small-scale melter testing to determine the maximum acceptable sulfate loadings. In the present work, a new glass formulation was developed and tested for a projected Hanford HLW composition with sulfate concentrations high enough to limit waste loading. Testing was then performed on the DM10 melter system at successively higher waste loadings to determine the maximum waste loading without the formation of a separate sulfate salt phase. Small scale corrosion testing was also conducted using the glass developed in the present work, the glass developed in the initial phase of this work [26], and a high iron composition, all at maximum sulfur concentrations determined from melter testing, in order to assess the extent of Inconel 690 and MA758 corrosion at elevated sulfate contents.

  12. Study on a Preliminary Survey and Analysis of HLW Management Technology Suitable for Nuclear Industrial Environment in Korea

    International Nuclear Information System (INIS)

    Kim, Eun Ka; Suh, In Suk; Ro, Seong Gy; Yoo, Kun Joong; Yoo, Jae Hyung; Cho, Sung Soo

    2010-12-01

    The purpose of this study is to suggest development direction of related technologies to analyze patented technology filed as a leading technology and to identify the technology trend for developing HLW management technology suitable for atomic industrial environment in Korea. For patent analysis of HLW management technology, international patent data were collected. And international application number, patent share of applicant and nationality, annual number of applications, application trends of assignees and detail technology, and frequency of patent citations / citations-to were analyzed by statistical analysis. Technical level and competitiveness through quantitative analysis by indicators of patent analysis were confirmed. And technology developments of blank technology, similarity analysis, the point of the main patent and a range of patent rights were analyzed through in-depth analysis. Trends of the patented technology of our country and world patent technology in such results have been identified, and statistical data on patents were secured. Especially in HLW management technology, patent application in Korea compared ti United States, Japan and European Union was began much later for the '90s, and are showing the annual increase on trend of patent application. Patent trend in Korea corresponds to development generation, while declining in foreign patent. The result of this study will be usefully applied to setting a development direction and blank technology of patent technology to pursue future in Korea

  13. Final Report Integrated DM1200 Melter Testing Using AZ-102 And C-106/AY-102 HLW Simulants: HLW Simulant Verification VSL-05R5800-1, Rev. 0, 6/27/05

    International Nuclear Information System (INIS)

    Kruger, A.A.; Matlack, K.S.; Gong, W.; Bardakci, T.; D'Angelo, N.A.; Brandys, M.; Kot, W.K.; Pegg, I.L.

    2011-01-01

    The principal objectives of the DM1200 melter tests were to determine the effects of feed rheology, feed solid content, and bubbler configuration on glass production rate and off-gas system performance while processing the HLW AZ-101 and C-106/AY-102 feed compositions; characterize melter off-gas emissions; characterize the performance of the prototypical off-gas system components, as well as their integrated performance; characterize the feed, glass product, and off-gas effluents; and perform pre- and post test inspections of system components. The specific objectives (including test success criteria) of this testing, along with how each objective was met, are outlined in a table. The data provided in this Final Report address the impacts of HLW melter feed rheology on melter throughput and validation of the simulated HLW melter feeds. The primary purpose of this testing is to further validate/verify the HLW melter simulants that have been used for previous melter testing and to support their continued use in developing melter and off-gas related processing information for the Project. The primary simulant property in question is rheology. Simulants and melter feeds used in all previous melter tests were produced by direct addition of chemicals; these feed tend to be less viscous than rheological the upper-bound feeds made from actual wastes. Data provided here compare melter processing for the melter feed used in all previous DM100 and DM1200 tests (nominal melter feed) with feed adjusted by the feed vendor (NOAH Technologies) to be more viscous, thereby simulating more closely the upperbounding feed produced from actual waste. This report provides results of tests that are described in the Test Plan for this work. The Test Plan is responsive to one of several test objectives covered in the WTP Test Specification for this work; consequently, only part of the scope described in the Test Specification was addressed in this particular Test Plan. For the purpose of

  14. FINAL REPORT INTEGRATED DM1200 MELTER TESTING USING AZ 102 AND C 106/AY-102 HLW SIMULANTS: HLW SIMULANT VERIFICATION VSL-05R5800-1 REV 0 6/27/05

    Energy Technology Data Exchange (ETDEWEB)

    KRUGER AA; MATLACK KS; GONG W; BARDAKCI T; D' ANGELO NA; BRANDYS M; KOT WK; PEGG IL

    2011-12-29

    The principal objectives of the DM1200 melter tests were to determine the effects of feed rheology, feed solid content, and bubbler configuration on glass production rate and off-gas system performance while processing the HLW AZ-101 and C-106/AY-102 feed compositions; characterize melter off-gas emissions; characterize the performance of the prototypical off-gas system components, as well as their integrated performance; characterize the feed, glass product, and off-gas effluents; and perform pre- and post test inspections of system components. The specific objectives (including test success criteria) of this testing, along with how each objective was met, are outlined in a table. The data provided in this Final Report address the impacts of HLW melter feed rheology on melter throughput and validation of the simulated HLW melter feeds. The primary purpose of this testing is to further validate/verify the HLW melter simulants that have been used for previous melter testing and to support their continued use in developing melter and off-gas related processing information for the Project. The primary simulant property in question is rheology. Simulants and melter feeds used in all previous melter tests were produced by direct addition of chemicals; these feed tend to be less viscous than rheological the upper-bound feeds made from actual wastes. Data provided here compare melter processing for the melter feed used in all previous DM100 and DM1200 tests (nominal melter feed) with feed adjusted by the feed vendor (NOAH Technologies) to be more viscous, thereby simulating more closely the upperbounding feed produced from actual waste. This report provides results of tests that are described in the Test Plan for this work. The Test Plan is responsive to one of several test objectives covered in the WTP Test Specification for this work; consequently, only part of the scope described in the Test Specification was addressed in this particular Test Plan. For the purpose of

  15. Time evolution of dissolved oxygen and redox conditions in a HLW repository

    International Nuclear Information System (INIS)

    Wersin, P.; Spahiu, K.; Bruno, J.

    1994-02-01

    The evolution of oxygen in a HLW repository has been studied using presently available geochemical background information. The important processes affecting oxygen migration in the near-field include diffusion and oxidation of pyrite and dissolved Fe(II). The evaluation of time scales of oxygen decrease is carried out with 1. an analytical approach involving the coupling of diffusion and chemical reaction, 2. a numerical geochemical approach involving the application of a newly developed diffusion-extended version of the STEADYQL code. Both approaches yield consistent rates of oxygen decrease and indicate that oxidation of pyrite impurities in the clay is the dominant process. The results obtained fRom geochemical modelling are interpreted in terms of evolution of redox conditions. Moreover, a sensitivity analysis of the major geochemical and physical parameters is performed. These results indicate that the uncertainties associated with reactive pyrite surface area impose the overall uncertainties of prediction of time scales. Thus, the obtained time of decrease to 1% of initial O 2 concentrations range between 7 and 290 years. The elapsed time at which the transition to anoxic conditions occurs is estimated to be within the same time range. Additional experimental information on redox sensitive impurities in the envisioned buffer and backfill material would further constrain the evaluated time scales. 41 refs

  16. Use of Gap-fills in the Buffer and Backfill of an HLW Repository

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Owan; Lee, Min Soo; Choi, Heui Joo [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    The buffer and backfill are significant barrier components of the repository. They play the roles of preventing the inflow of groundwater from the surrounding rock, retarding the release of radionuclides from the waste, supporting disposal container against external impacts, and discharging decay heat from the waste. When the buffer and backfill are installed for the HLW repository, there may be gaps between the container and buffer and between the backfill and the wall of disposal tunnels, respectively. These gaps occur because spaces are allowed for ease of the installation of the buffer and backfill in excavated deposition boreholes and disposal tunnels. If the gaps are left without any sealing as they are, however, the buffer and backfill can't accomplish their functions as the barrier components. This paper reviews the gap-fill concepts of the developed foreign countries, and then suggests a gap-fill concept which is applicable for the KRS. The gap-fill is suggested to employ bentonite- based materials with a type of pellet, granule, and pellet-granule mixture. The roller compression method and extrusion-cutting method are applicable for the fabrication of the bentonite pellets which can have the high density and the required amount for use to the buffer and backfill. For the installation of the gap-fill, the pouring and then pressing method and the shotcrete- blowing method are preferable for the gap of the deposition borehole and the gap of the disposal tunnel, respectively.

  17. Preliminary formulation studies for a ''hydroceramic'' alternative waste form for INEEL HLW

    International Nuclear Information System (INIS)

    Siemer, D.D.; Gougar, M.L.D.; Grutzeck, M.W.; Scheetz, B.E.

    1999-01-01

    Herein the authors discuss scoping studies performed to develop an efficient way to prepare the Idaho National Engineering and Environmental Laboratory (INEEL) nominally high-level (∼40 W/m 3 ) calcined radioactive waste (HLW) and liquid metal (sodium) reactor coolants for disposal. The investigated approach implements the chemistry of Hanford's cancrinite-making clay reaction process via Oak Ridge National Laboratory's (ORNL's) formed-under-elevated-temperatures-and-pressures concrete monolith-making technology to make hydroceramics (HCs). The HCs differ from conventional Portland cement/blast furnace slag (PC/BFS) grouts in that the binder minerals formed during the curing process are hydrated alkali-aluminosilicates (feldspathoids-sodalites, cancrinites, and zeolites) rather than hydrated calcium silicates (CSH). This is desirable because (a) US defense-type radioactive wastes generally contain much more sodium and aluminum than calcium; (b) sodalites/cancrinites do a much better job of retaining the anionic components of real radioactive waste (e.g., nitrate) than do calcium silicates; (c) natural feldspathoids form from glasses (and therefore are more stable) in that region of the United States where a repository for this sort of waste could be sited; and (d) if eventually deemed necessary, feldspathoid-type concrete wasteforms could be hot-isostatically-pressed into even more durable materials without removing them from their original canisters

  18. Performance Assessment Uncertainty Analysis for Japan's HLW Program Feasibility Study (H12)

    International Nuclear Information System (INIS)

    BABA, T.; ISHIGURO, K.; ISHIHARA, Y.; SAWADA, A.; UMEKI, H.; WAKASUGI, K.; WEBB, ERIK K.

    1999-01-01

    Most HLW programs in the world recognize that any estimate of long-term radiological performance must be couched in terms of the uncertainties derived from natural variation, changes through time and lack of knowledge about the essential processes. The Japan Nuclear Cycle Development Institute followed a relatively standard procedure to address two major categories of uncertainty. First, a FEatures, Events and Processes (FEPs) listing, screening and grouping activity was pursued in order to define the range of uncertainty in system processes as well as possible variations in engineering design. A reference and many alternative cases representing various groups of FEPs were defined and individual numerical simulations performed for each to quantify the range of conceptual uncertainty. Second, parameter distributions were developed for the reference case to represent the uncertainty in the strength of these processes, the sequencing of activities and geometric variations. Both point estimates using high and low values for individual parameters as well as a probabilistic analysis were performed to estimate parameter uncertainty. A brief description of the conceptual model uncertainty analysis is presented. This paper focuses on presenting the details of the probabilistic parameter uncertainty assessment

  19. Strategy for safety case development: impact of a volunteering approach to siting a japanese HLW repository

    International Nuclear Information System (INIS)

    Kitayama, K.; Ishiguro, K.; Takeuchi, M.; Tsuchi, H.; Kato, T.; Sakabe, Y.; Wakasugi, K.

    2008-01-01

    NUMO strategy for safety case development is constrained by a staged siting approach, which has been initiated by a call for volunteer municipalities to host the HLW repository. For each site, the safety case is an important factor to be considered at the selection steps which narrow down towards the preferred repository location. This is particularly challenging, however, as every site requires a tailored repository concept, with associated performance assessment and an individual site evaluation programme all of which evolve with gradually increasing understanding of the host environment. In order to maintain flexibility without losing focus, NUMO has developed a formalized tailoring procedure, termed the NUMO Structured Approach (NSA). The NSA guides the interaction of the key site characterisation, repository design and performance assessment groups and is facilitated by tools to help the decision making associated with the tailoring process (e.g. a requirements management system) and with comparison of siting and design options (e.g. multi-attribute analysis). Pragmatically, the post-closure safety case will initially emphasize near-field processes and a robust engineering barrier system, considering the limited geological information at early stages. This will be complemented by a more realistic assessment of total system performance, as needed to compare options. In addition, efforts to rigorously assess operational phase safety and the practicality of assuring quality of the constructed engineered barriers are components of the total safety case which are receiving particular attention now, as they may better discriminate between sites while information is still limited. (authors)

  20. Cleanup of a HLW nuclear fuel-reprocessing center using 3-D database modeling technology

    International Nuclear Information System (INIS)

    Sauer, R.C.

    1992-01-01

    A significant challenge in decommissioning any large nuclear facility is how to solidify the large volume of residual high-level radioactive waste (HLW) without structurally interfering with the existing equipment and piping used at the original facility or would require rework due to interferences which were not identified during the design process. This problem is further compounded when the nuclear facility to be decommissioned is a 35 year old nuclear fuel reprocessing center designed to recover usable uranium and plutonium. Facilities of this vintage usually tend to lack full documentation of design changes made over the years and as a result, crude traps or pockets of high-level contamination may not be fully realized. Any miscalculation in the construction or modification sequences could compound the overall dismantling and decontamination of the facility. This paper reports that development of a 3-dimensional (3-D) computer database tool was considered critical in defining the most complex portions of this one-of-a-kind vitrification facility

  1. Nuclide release calculation in the near-field of a reference HLW repository

    International Nuclear Information System (INIS)

    Lee, Youn Myoung; Hwang, Yong Soo; Kang, Chul Hyung

    2004-01-01

    The HLW-relevant R and D program for disposal of high-level radioactive waste has been carried out at Korea Atomic Energy Research Institute (KAERI) since early 1997 in order to develop a conceptual Korea Reference Repository System for direct disposal of nuclear spent fuel by the end of 2007. A preliminary reference geologic repository concept considering such established criteria and requirements as waste and generic site characteristics in Korea was roughly envisaged in 2003 focusing on the near-field components of the repository system. According to above basic repository concept, which is similar to that of Swedish KBS-3 repository, the spent fuel is first encapsulated in corrosion resistant canisters, even though the material has not yet been determined, and then emplaced into the deposition holes surrounded by high density bentonite clay in tunnels constructed at a depth of about 500 m in a stable plutonic rock body. Not only to demonstrate how much a reference repository is safe in the generic point of view with several possible scenarios and cases associated with a preliminary repository concept by conducting calculations for nuclide release and transport in the near-field components of the repository, even though enough information has not been available that much yet, but also to show a methodology by which a generic safety assessment could be performed for further development of Korea reference repository concept, nuclide release calculation study strongly seems to be necessary

  2. Biosphere Modeling for the Dose Assessment of a HLW Repository: Development of ACBIO

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Youn Myoung; Hwang, Yong Soo; Kang, Chul Hyung

    2006-01-15

    For the purpose of evaluating a dose rate to an individual due to a long-term release of nuclides from a HLW repository, a biosphere assessment model and an implemented code, ACBIO, based on the BIOMASS methodology have been developed by utilizing AMBER, a general compartment modeling tool. To demonstrate its practicability and usability as well as to observe the sensitivity of the compartment scheme, the concentration, the activity in the compartments as well as the annual flux between the compartments at their peak values, were calculated and investigated. For each case when changing the structure of the compartments and GBIs as well as varying selected input Kd values, all of which seem very important among the others, the dose rate per nuclide release rate is calculated separately and analyzed. From the maximum dose rates, the flux to dose conversion factors for each nuclide were derived, which are used for converting the nuclide release rate appearing from the geosphere through various GBIs to dose rates (Sv/y) for an individual in a critical group. It has also been observed that the compartment scheme, the identification of a possible exposure group and the GBIs could all be highly sensitive to the final consequences in a biosphere modeling.

  3. Storage of HLW in engineered structures: air-cooled and water-cooled concepts

    International Nuclear Information System (INIS)

    Ahner, S.; Dekais, J.J.; Puttke, B.; Staner, P.

    1981-01-01

    A comparative study on an air-cooled and a water-cooled intermediate storage of vitrified, highly radioactive waste (HLW) in overground installations has been performed by Nukem and Belgonucleaire respectively. In the air-cooled storage concept the decay heat from the storage area will be removed using natural convection. In the water-cooled storage concept the decay heat is carried off by a primary and secondary forced-cooling system with redundant and diverse devices. The safety study carried out by Nukem used a fault tree method. It shows that the reliability of the designed water-cooled system is very high and comparable to the inherent, safe, air-cooled system. The impact for both concepts on the environment is determined by the release route, but even during accident conditions the release is far below permissible limits. The economic analysis carried out by Belgonucleaire shows that the construction costs for both systems do not differ very much, but the operation and maintenance costs for the water-cooled facility are higher than for the air cooled facility. The result of the safety and economic analysis and the discussions with the members of the working group have shown some possible significant modifications for both systems, which are included in this report. The whole study has been carried out using certain national criteria which, in certain Member States at least, would lead to a higher standard of safety than can be justified on any social, political or economic grounds

  4. Development of a computer tool to support scenario analysis for safety assessment of HLW geological disposal

    International Nuclear Information System (INIS)

    Makino, Hitoshi; Kawamura, Makoto; Wakasugi, Keiichiro; Okubo, Hiroo; Takase, Hiroyasu

    2007-02-01

    In 'H12 Project to Establishing Technical Basis for HLW Disposal in Japan' a systematic approach that was based on an international consensus was adopted to develop scenarios to be considered in performance assessment. Adequacy of the approach was, in general term, appreciated through the domestic and international peer review. However it was also suggested that there were issues related to improving transparency and traceability of the procedure. To achieve this, improvement of scenario analysis method has been studied. In this study, based on an improvement method for treatment of FEP interaction a computer tool to support scenario analysis by specialists of performance assessment has been developed. Anticipated effects of this tool are to improve efficiency of complex and time consuming scenario analysis work and to reduce possibility of human errors in this work. This tool also enables to describe interactions among a vast number of FEPs and the related information as interaction matrix, and analysis those interactions from a variety of perspectives. (author)

  5. Impact Of Particle Agglomeration On Accumulation Rates In The Glass Discharge Riser Of HLW Melter

    International Nuclear Information System (INIS)

    Kruger, A. A.; Rodriguez, C. A.; Matyas, J.; Owen, A. T.; Jansik, D. P.; Lang, J. B.

    2012-01-01

    The major factor limiting waste loading in continuous high-level radioactive waste (HLW) melters is an accumulation of particles in the glass discharge riser during a frequent and periodic idling of more than 20 days. An excessive accumulation can produce robust layers a few centimeters thick, which may clog the riser, preventing molten glass from being poured into canisters. Since the accumulation rate is driven by the size of particles we investigated with x-ray microtomography, scanning electron microscopy, and image analysis the impact of spinel forming components, noble metals, and alumina on the size, concentration, and spatial distribution of particles, and on the accumulation rate. Increased concentrations of Fe and Ni in the baseline glass resulted in the formation of large agglomerates that grew over the time to an average size of ∼185+-155 μm, and produced >3 mm thick layer after 120 h at 850 deg C. The noble metals decreased the particle size, and therefore significantly slowed down the accumulation rate. Addition of alumina resulted in the formation of a network of spinel dendrites which prevented accumulation of particles into compact layers

  6. The development of basic glass formulations for solidifying HLW from nuclear fuel reprocessing plant

    International Nuclear Information System (INIS)

    Jiang Yaozhong; Tang Baolong; Zhang Baoshan; Zhou Hui

    1995-01-01

    Basic glass formulations 90U/19, 90U/20, 90Nd/7 and 90Nd/10 applied in electric melting process are developed by using the mathematical model of the viscosity and electric resistance of waste glass. The yellow phase does not occur for basic glass formulations 90U/19 and 90U/20 solidifying HLW from nuclear fuel reprocessing plant when the waste loading is 20%. Under the waste loading is 16%, the process and product properties of glass 90U/19 and 90U/20 come up to or surpass the properties of the same kind of foreign waste glasses, and other properties are about the same to them of foreign waste glasses. The process and product properties of basic glass formulations 90Nd/7 and 90Nd/10 used for the solidification of 'U replaced by Nd' liquid waste are almost similar to them of 90U/19 and 90U/20. These properties fairly meet the requirements of 'joint test' (performed at KfK-INE, Germany). Among these formulations, 90Nd/7 is applied in cold engineering scale electric melting test performed at KfK-INE in Germany. The main process properties of cold test is similar to laboratory results

  7. Biosphere Modeling for the Dose Assessment of a HLW Repository: Development of ACBIO

    International Nuclear Information System (INIS)

    Lee, Youn Myoung; Hwang, Yong Soo; Kang, Chul Hyung

    2006-01-01

    For the purpose of evaluating a dose rate to an individual due to a long-term release of nuclides from a HLW repository, a biosphere assessment model and an implemented code, ACBIO, based on the BIOMASS methodology have been developed by utilizing AMBER, a general compartment modeling tool. To demonstrate its practicability and usability as well as to observe the sensitivity of the compartment scheme, the concentration, the activity in the compartments as well as the annual flux between the compartments at their peak values, were calculated and investigated. For each case when changing the structure of the compartments and GBIs as well as varying selected input Kd values, all of which seem very important among the others, the dose rate per nuclide release rate is calculated separately and analyzed. From the maximum dose rates, the flux to dose conversion factors for each nuclide were derived, which are used for converting the nuclide release rate appearing from the geosphere through various GBIs to dose rates (Sv/y) for an individual in a critical group. It has also been observed that the compartment scheme, the identification of a possible exposure group and the GBIs could all be highly sensitive to the final consequences in a biosphere modeling

  8. Seismic design evaluation guidelines for buried piping for the DOE HLW Facilities

    International Nuclear Information System (INIS)

    Lin, Chi-Wen; Antaki, G.; Bandyopadhyay, K.; Bush, S.H.; Costantino, C.; Kennedy, R.

    1995-01-01

    This paper presents the seismic design and evaluation guidelines for underground piping for the Department of Energy (DOE) High-Level-Waste (HLW) Facilities. The underground piping includes both single and double containment steel pipes and concrete pipes with steel lining, with particular emphasis on the double containment piping. The design and evaluation guidelines presented in this paper follow the generally accepted beam-on-elastic-foundation analysis principle and the inertial response calculation method, respectively, for piping directly in contact with the soil or contained in a jacket. A standard analysis procedure is described along with the discussion of factors deemed to be significant for the design of the underground piping. The following key considerations are addressed: the design feature and safety requirements for the inner (core) pipe and the outer pipe; the effect of soil strain and wave passage; assimilation of the necessary seismic and soil data; inertial response calculation for the inner pipe; determination of support anchor movement loads; combination of design loads; and code comparison. Specifications and justifications of the key parameters used, stress components to be calculated and the allowable stress and strain limits for code evaluation are presented

  9. A Safety Case Approach for Deep Geologic Disposal of DOE HLW and DOE SNF in Bedded Salt - 13350

    Energy Technology Data Exchange (ETDEWEB)

    Sevougian, S. David [Advanced Nuclear Energy Programs Group, Sandia National Laboratories, P.O. Box 5800, Albuquerque, NM 87185 (United States); MacKinnon, Robert J. [Advanced Nuclear Energy Programs Group, Sandia National Laboratories, P.O. Box 5800, Albuquerque, NM 87185 (United States); Leigh, Christi D. [Defense Waste Management Programs Group, Sandia National Laboratories, P.O. Box 5800, Albuquerque, NM 87185 (United States); Hansen, Frank D. [Geoscience Research and Applications Group, Sandia National Laboratories, P.O. Box 5800, Albuquerque, NM 87185 (United States)

    2013-07-01

    The primary objective of this study is to investigate the feasibility and utility of developing a defensible safety case for disposal of United States Department of Energy (U.S. DOE) high-level waste (HLW) and DOE spent nuclear fuel (SNF) in a conceptual deep geologic repository that is assumed to be located in a bedded salt formation of the Delaware Basin [1]. A safety case is a formal compilation of evidence, analyses, and arguments that substantiate and demonstrate the safety of a proposed or conceptual repository. We conclude that a strong initial safety case for potential licensing can be readily compiled by capitalizing on the extensive technical basis that exists from prior work on the Waste Isolation Pilot Plant (WIPP), other U.S. repository development programs, and the work published through international efforts in salt repository programs such as in Germany. The potential benefits of developing a safety case include leveraging previous investments in WIPP to reduce future new repository costs, enhancing the ability to effectively plan for a repository and its licensing, and possibly expediting a schedule for a repository. A safety case will provide the necessary structure for organizing and synthesizing existing salt repository science and identifying any issues and gaps pertaining to safe disposal of DOE HLW and DOE SNF in bedded salt. The safety case synthesis will help DOE to plan its future R and D activities for investigating salt disposal using a risk-informed approach that prioritizes test activities that include laboratory, field, and underground investigations. It should be emphasized that the DOE has not made any decisions regarding the disposition of DOE HLW and DOE SNF. Furthermore, the safety case discussed herein is not intended to either site a repository in the Delaware Basin or preclude siting in other media at other locations. Rather, this study simply presents an approach for accelerated development of a safety case for a potential

  10. A Safety Case Approach for Deep Geologic Disposal of DOE HLW and DOE SNF in Bedded Salt - 13350

    International Nuclear Information System (INIS)

    Sevougian, S. David; MacKinnon, Robert J.; Leigh, Christi D.; Hansen, Frank D.

    2013-01-01

    The primary objective of this study is to investigate the feasibility and utility of developing a defensible safety case for disposal of United States Department of Energy (U.S. DOE) high-level waste (HLW) and DOE spent nuclear fuel (SNF) in a conceptual deep geologic repository that is assumed to be located in a bedded salt formation of the Delaware Basin [1]. A safety case is a formal compilation of evidence, analyses, and arguments that substantiate and demonstrate the safety of a proposed or conceptual repository. We conclude that a strong initial safety case for potential licensing can be readily compiled by capitalizing on the extensive technical basis that exists from prior work on the Waste Isolation Pilot Plant (WIPP), other U.S. repository development programs, and the work published through international efforts in salt repository programs such as in Germany. The potential benefits of developing a safety case include leveraging previous investments in WIPP to reduce future new repository costs, enhancing the ability to effectively plan for a repository and its licensing, and possibly expediting a schedule for a repository. A safety case will provide the necessary structure for organizing and synthesizing existing salt repository science and identifying any issues and gaps pertaining to safe disposal of DOE HLW and DOE SNF in bedded salt. The safety case synthesis will help DOE to plan its future R and D activities for investigating salt disposal using a risk-informed approach that prioritizes test activities that include laboratory, field, and underground investigations. It should be emphasized that the DOE has not made any decisions regarding the disposition of DOE HLW and DOE SNF. Furthermore, the safety case discussed herein is not intended to either site a repository in the Delaware Basin or preclude siting in other media at other locations. Rather, this study simply presents an approach for accelerated development of a safety case for a potential

  11. Final Report - Melt Rate Enhancement for High Aluminum HLW Glass Formulation, VSL-08R1360-1, Rev. 0, dated 12/19/08

    Energy Technology Data Exchange (ETDEWEB)

    Kruger, Albert A.; Pegg, I. L.; Chaudhuri, M.; Gong, W.; Gan, H.; Matlack, K. S.; Bardakci, T.; Kot, W.

    2013-11-13

    The principal objective of the work reported here was to develop and identify HLW glass compositions that maximize waste processing rates for the aluminum limted waste composition specified by ORP while maintaining high waste loadings and acceptable glass properties. This was accomplished through a combination of crucible-scale tests, confirmation tests on the DM100 melter system, and demonstration at pilot scale (DM1200). The DM100-BL unit was selected for these tests since it was used previously with the HLW waste streams evaluated in this study, was used for tests on HLW glass compositions to support subsequent tests on the HLW Pilot Melter, conduct tests to determine the effect of various glass properties (viscosity and conductivity) and oxide concentrations on glass production rates with HLW feed streams, and to assess the volatility of cesium and technetium during the vitrification of an HLW AZ-102 composition. The same melter was selected for the present tests in order to maintain comparisons between the previously collected data. These tests provide information on melter processing characteristics and off-gas data, including formation of secondary phases and partitioning. Once DM100 tests were completed, one of the compositions was selected for further testing on the DM1200; the DM1200 system has been used for processing a variety of simulated Hanford waste streams. Tests on the larger melter provide processing data at one third of the scale of the actual WTP HLW melter and, therefore, provide a more accurate and reliable assessment of production rates and potential processing issues. The work focused on maximizing waste processing rates for high aluminum HLW compositions. In view of the diversity of forms of aluminum in the Hanford tanks, tests were also conducted on the DM100 to determine the effect of changes in the form of aluminum on feed properties and production rate. In addition, the work evaluated the effect on production rate of modest increases

  12. Soluble CD163

    DEFF Research Database (Denmark)

    Møller, Holger J

    2012-01-01

    CD163 is an endocytic receptor for haptoglobin-hemoglobin complexes and is expressed solely on macrophages and monocytes. As a result of ectodomain shedding, the extracellular portion of CD163 circulates in blood as a soluble protein (sCD163) at 0.7-3.9 mg/l in healthy individuals. The function o...

  13. Solubility Part 1

    NARCIS (Netherlands)

    Tantra, Ratna; Bolea, Eduardo; Bouwmeester, H.; Rey-Castro, Carlos; David, C.A.A.; Dogné, Jean Michel; Laborda, Francisco; Laloy, Julie; Robinson, Kenneth N.; Undas, A.K.; Zande, van der M.

    2016-01-01

    This chapter gives an overview of different methods that can potentially be used to determine the solubility of nanomaterials. In general, the methods presented can be broadly divided into four categories: separation methods, methods to quantify free ions, methods to quantify total dissolved

  14. The aqueous solubility and speciation analysis for uranium, neptunium and selenium by the geochemical code(EQ3/6)

    International Nuclear Information System (INIS)

    Takeda, Seiji; Shima, Shigeki; Kimura, Hideo; Matsuzuru, Hideo

    1995-11-01

    The geochemical condition of a geologic disposal system of HLW controls the solubility and physicochemical forms of dominant aqueous species for elements, which are one of essential information required for safety assessment. Based on the measured compositions of groundwater, the compositions of groundwater in the disposal system were calculated. The solubility and speciation analyses for the polyvalent elements, uranium, neptunium, and selenium, were performed by the geochemical code EQ3/6. The results obtained were compared with the data appeared in the literatures on the solubilities and speciations. The geochemical behaviors of the elements with respect to the solubility and speciation could quantitatively be elucidated for the compositions of the interstitial waters in an engineered barrier and ground water in a natural barrier. In the pH range of neutral to alkali, the solubilities of U and Np tend to increase with an increase of the carbonate concentration in groundwater. This carbonate concentration dependence of the solubility was also estimated. In the engineered barrier the predominant aqueous species were specified, and in the natural barrier the change of aqueous species was also predicted while the chemical compositions changed from the reducing to oxidizing conditions. The dominant aqueous species for the elements, which migrate in and through the disposal system, were determined by the speciation analysis. (author)

  15. Methodology of fuel cycles long-term safety assessment of SNF/HLW geological disposal

    International Nuclear Information System (INIS)

    Pritrsky, J.

    2008-02-01

    Methodology for the long-term safety assessment of nuclear fuel cycles is given in the presented doctoral thesis. The aim of work was to develop a geological repository model for disposal of spent nuclear fuel (SNF) and high level waste (HLW) using an appropriate software code able to calculate the influence of partitioning and transmutation in advanced fuel cycles. The first step in this process was specifying of indicators which can be used to quantify the radiological impact of each fuel cycle. Indicators such as annual effective dose and radiotoxicity of inventory have been quantitatively analysed to determine the potential risk and radiological consequences associated with production of SNF/HLW. Advanced fuel types bring a number of advantages in comparison to uranium oxide fuel UO 2 used worldwide nowadays in terms of safety improvement due to minor actinides transmutation and non-proliferation aspects as well. Within the scope of work, three different fuel cycles are compared from the point of view of long-term safety of deep geological repository. The first considered fuel cycle is the currently used open fuel cycle (UOX) which uses only U-FA (Uranium Fuel Assembly). The second assessed cycle is a closed fuel cycle (MOX) with MOX-FA (Mixed OXides Fuel Assembly) and the third considered one is a partially closed fuel cycle (IMF) with IMC-FA (Inert Matrix Combined Fuel Assembly). Description and input data of advanced fuel cycles have been gained by participation in the EC project RED-IMPACT. Results were calculated using code AMBER, which is a flexible software tool that allows building dynamic compartmental models to represent the migration and fate of contaminants in a system, for example in the surface and sub-surface environment. Contaminants in solid, liquid and gaseous phases can be considered. AMBER gives the user the flexibility to define any number of compartments; any number of contaminants and associated decays; deterministic, probabilistic and

  16. Methodology of fuel cycles long-term safety assessment of SNF/HLW geological disposal

    International Nuclear Information System (INIS)

    Pritrsky, J.

    2008-01-01

    Methodology for the long-term safety assessment of nuclear fuel cycles is given in the presented doctoral thesis. The aim of work was to develop a geological repository model for disposal of spent nuclear fuel (SNF) and high level waste (HLW) using an appropriate software code able to calculate the influence of partitioning and transmutation in advanced fuel cycles. The first step in this process was specifying of indicators which can be used to quantify the radiological impact of each fuel cycle. Indicators such as annual effective dose and radiotoxicity of inventory have been quantitatively analysed to determine the potential risk and radiological consequences associated with production of SNF/HLW. Advanced fuel types bring a number of advantages in comparison to uranium oxide fuel UO 2 used worldwide nowadays in terms of safety improvement due to minor actinides transmutation and non-proliferation aspects as well. Within the scope of work, three different fuel cycles are compared from the point of view of long-term safety of deep geological repository. The first considered fuel cycle is the currently used open fuel cycle (UOX) which uses only U-FA (Uranium Fuel Assembly). The second assessed cycle is a closed fuel cycle (MOX) with MOX-FA (Mixed OXides Fuel Assembly) and the third considered one is a partially closed fuel cycle (IMF) with IMC-FA (Inert Matrix Combined Fuel Assembly). Description and input data of advanced fuel cycles have been gained by participation in the EC project RED-IMPACT. Results were calculated using code AMBER, which is a flexible software tool that allows building dynamic compartmental models to represent the migration and fate of contaminants in a system, for example in the surface and sub-surface environment. Contaminants in solid, liquid and gaseous phases can be considered. AMBER gives the user the flexibility to define any number of compartments; any number of contaminants and associated decays; deterministic, probabilistic and

  17. Hydro-mechanical behaviour of crushed COx argillite used as backfilling material in HLW repository

    International Nuclear Information System (INIS)

    Tang Chaosheng; Shi Bin; Cui Yujun; Anh-Minh Tang

    2010-01-01

    At present, the crushed Callovo-Oxfordian (COx) argillite powder is proposed as an alternative backfilling material in France, which will be constructed in the engineering barrier of high-level radioactive waste (HLW) repository. In this investigation, the compression behavior of two crushed COx argillite powders (coarser one and finer one) was studied by running l-D compression tests with several loading-unloading cycles. After the final dry density 2.0 g/cm 3 was reached, the specimen was flooding with distilled water and the evolution of axial stress was studied during saturation process. The effects of initial axial stress level and grain size distribution (GSD) on hydro-mechanical behaviour of compacted specimen were analyzed. The results show that the compression curves are significantly influenced by the GSD of the soils. To obtain the same degree of compaction, the axial stress applied to finer soil is much higher than that of coarser soil. In addition, the compression index of the finer soil is bigger than that of coarser soil. The swelling index at initial water content increases with the dry density and seems to be independent of the GSD. During saturation, the initial lower axial stress causes obvious swelling behavior for both the coarser and finer powder samples and the corresponding axial stress increase gradually. At initial higher axial stress condition, monotone collapse behavior is observed for the coarser powder samples. Whereas the axial stress decrease firstly, then increase and finally decrease again for the finer powder samples. After saturation, the equilibrium axial stresses of finer powder samples are higher than that of coarser powder samples. (authors)

  18. Rheology of Savannah River site tank 42 and tank 51 HLW radioactive sludges

    International Nuclear Information System (INIS)

    Ha, B.C.; Bibler, N.E.

    1996-01-01

    Knowledge of the rheology of the radioactive sludge slurries at the Savannah River Site (SRS) is necessary in order to ensure that they can be retrieved from waste tanks and processed for final disposal. The high activity radioactive wastes stored as caustic slurries at SRS result from the neutralization of acid waste generated from production of nuclear defense materials. During storage, the wastes separate into a supernate layer and a sludge layer. In the Defense Waste Processing Facility (DWPF) at SRS, the radionuclides from the sludge and supernate will be immobilized into borosilicate glass for long term storage and eventual disposal. Before transferring the waste from a storage tank to the DWPF, a portion of the aluminum in the waste sludge will be dissolved and the sludge will be extensively washed to remove sodium. Tank 51 and Tank 42 radioactive sludges represent the first batch of HLW sludge to be processed in the DWPF. This paper presents results of rheology measurements of Tank 51 and Tank 42 at various solids concentrations. The rheologies of Tank 51 and Tank 42 radioactive slurries were measured remotely in the Shielded Cells Operations (SCO) at the Savannah River Technology Center (SRTC) using a modified Haake Rotovisco RV-12 with an M150 measuring drive unit and TI sensor system. Rheological properties of the Tank 51 and Tank 42 radioactive sludges were measured as a function of weight percent solids. The weight percent solids of Tank 42 sludge was 27, as received. Tank 51 sludge had already been washed. The weight percent solids were adjusted by dilution with water or by concentration through drying. At 12, 15, and 18 weight percent solids, the yield stresses of Tank 51 sludge were 5, 11, and 14 dynes/cm2, respectively. The apparent viscosities were 6, 10, and 12 centipoises at 300 sec-1 shear rate, respectively

  19. Time-frames and the demonstration of safety for HLW disposal

    International Nuclear Information System (INIS)

    Watkins, B.; Kessler, J.

    1999-01-01

    An important principle which is often embodied in the criteria for the safe disposal of long-lived radioactive wastes is that a similar level of radiation protection should be provided to future generations as that provided for those alive today. This has resulted in the development of performance assessment methodologies to evaluate the potential long term impacts of HLW disposal on humans, usually in terms of individual dose or risk. However, the actual periods of time over which it is expected that there will be full control over high level waste disposals are extremely short in comparison with the times over which radionuclides in the wastes could potentially move from the deep repository and emerge into the surface environment. This leads to problems in setting quantitative dose or risk based standard appropriate for the short and long term, and in setting the time-frames for which calculations should be carried out. This is especially difficult in view of the uncertainty in predicting changes in human behaviour and changes in the biosphere and geosphere over the time-scales involved. Different assessment time-frames and approaches proposed by IAEA, Nordic countries, Britain and US guidance documents are briefly reviewed. Whilst accepting the basic radiation protection objective of protecting future generations, no international consensus bas been agreed on what time-frames should be used in performance assessments. It is recommended that different time-frames should be associated with different quantitative or qualitative performance measures. As a result, a range of indicators of safety may be appropriate in demonstrating compliance with regulatory performance criteria and the consequent overall assessment context. It is argued that what is required is a simple, robust yet defensible approach to time-frames and performance indicators which can be accepted by the public, regulators and the nuclear industry

  20. DESIGN OF A CONCRETE SLAB FOR STORAGE OF SNF AND HLW CASKS

    International Nuclear Information System (INIS)

    J. Bisset

    2005-01-01

    This calculation documents the design of the Spent Nuclear Fuel (SNF) and High-Level Waste (HLW) Cask storage slab for the Aging Area. The design is based on the weights of casks that may be stored on the slab, the weights of vehicles that may be used to move the casks, and the layout shown on the sketch for a 1000 Metric Ton of Heavy Metal (MTHM) storage pad on Attachment 2, Sht.1 of the calculation 170-C0C-C000-00100-000-00A (BSC 2004a). The analytical model used herein is based on the storage area for 8 vertical casks. To simplify the model, the storage area of the horizontal concrete modules and their related shield walls is not included. The heavy weights of the vertical storage casks and the tensile forces due to pullout at the anchorages will produce design moments and shear forces that will envelope those that would occur in the storage area of the horizontal modules. The design loadings will also include snow and live loads. In addition, the design will also reflect pertinent geotechnical data. This calculation will document the preliminary thickness and general reinforcing steel requirements for the slab. This calculation also documents the initial design of the cask anchorage. Other slab details are not developed in this calculation. They will be developed during the final design process. The calculation also does not include the evaluation of the effects of cask drop loads. These will be evaluated in this or another calculation when the exact cask geometry is known

  1. Uranyl Oxalate Solubility

    Energy Technology Data Exchange (ETDEWEB)

    Leturcq, G.; Costenoble, S.; Grandjean, S. [CEA Marcoule DEN/DRCP/SCPS/LCA - BP17171 - 30207 Bagnols sur Ceze cedex (France)

    2008-07-01

    The solubility of uranyl oxalate was determined at ambient temperature by precipitation in oxalic-nitric solutions, using an initial uranyl concentration of 0.1 mol/L. Oxalic concentration varied from 0.075 to 0.3 mol/L while nitric concentration ranged between 0.75 and 3 mol/L. Dissolution tests, using complementary oxalic-nitric media, were carried out for 550 hours in order to study the kinetic to reach thermodynamic equilibrium. Similar solubility values were reached by dissolution and precipitation. Using the results, it was possible to draw the solubility surface versus oxalic and nitric concentrations and to determine both the apparent solubility constant of UO{sub 2}C{sub 2}O{sub 4}, 3H{sub 2}O (Ks) and the apparent formation constant of the first uranyl-oxalate complex UO{sub 2}C{sub 2}O{sub 4} (log {beta}1), for ionic strengths varying between 1 and 3 mol/L. Ks and log {beta}1 values were found to vary from 1.9 10{sup -8} to 9.2 10{sup -9} and from 5.95 to 6.06, respectively, when ionic strength varied from 1 to 3 mol/L. A second model may fit our data obtained at an ionic strength of 3 mol/L suggesting as reported by Moskvin et al. (1959) that no complexes are formed for [H{sup +}] at 3 M. The Ks value would then be 1.3 10{sup -8}. (authors)

  2. Issues at stake when considering long term storage of HLW. A comprehensive approach to designing the facility

    International Nuclear Information System (INIS)

    Marvy, A.; Ochem, D.

    2002-01-01

    CEA has been conducting a comprehensive R and D program to identify and study key HLW storage design criteria to possibly meet the lifetime goal of a century and beyond. A novel approach is being used since such installations must be understood as a global system comprised of various materials and hardware components, canisters, concrete and steel structures and specific procedures covering engineering steps from construction to operation including monitoring, care and maintenance as well as licensing. The challenge set by such a lifetime design goal made the R and D people focus on issues at stake and relevant to long term HLW storage in particular heat management, the effect of time on materials and the sustainability of care and maintenance. This opened up the R and D field from fundamental research areas to more conventional and technical aspects. Two major guiding principles have been devised as key design goals for the storage concepts under consideration. One is the paramount function of retrievability, which must allow the safe retrieval of any HLW package from the facility at any given time. Next is the passive containment philosophy requiring that a two-barrier system be considered. In the case of spent fuel, CEA's early assessment of the long-term behaviour of cladding shows that it cannot qualify as a reliable barrier over a long period of time. Therefore, the overriding strategy of preventing corrosion and material degradation to achieve canister protection, and therefore containment of radioactive material throughout the time of period envisaged, is at the heart of the R and D program and several design alternatives are being studied to meet that objective. For instance available thermal power from SF is used to establish dry corrosion conditions within the storage facility. The paper reviews all of these different R and D and engineering aspects. (author)

  3. Glass formulation development and testing for the vitrification of DWPF HLW sludge coupled with crystalline silicotitanate (CST)

    International Nuclear Information System (INIS)

    Andrews, M.K.; Workman, P.J.

    1997-01-01

    An alternative to the In Tank Precipitation and sodium titanate processes at the Savannah River Site is the removal of cesium, strontium, and plutonium from the tank supernate by ion exchange using crystalline silicotitanate (CST). This inorganic material has been shown to effectively and selectively sorb these elements from supernate. The loaded CST could then be immobilized with High-Level Waste (HLW) sludge during vitrification. Initial efforts on the development of a glass formulation for a coupled waste stream indicate that reasonable loadings of both sludge and CST can be achieved in glass

  4. Natural analogue of redox front formation in near-field environment at post-closure phase of HLW geological disposal

    International Nuclear Information System (INIS)

    Yoshida, Hidekazu; Yamamoto, Koushi; Amano, Yuki

    2005-01-01

    Redox fronts are created in the near field of rocks, in a range of oxidation environments, by microbial activity in rock groundwater. Such fronts, and the associated oxide formation, are usually unavoidable around high level radioactive waste (HLW) repositories, whatever their design. The long term behaviour of these oxides after repositories have been closed is however little known. Here we introduce an analogue of redox front formation, such as 'iron oxide' deposits, known as takashikozo forming cylindrical nodules, and the long term behaviour of secondarily formed iron oxyhydroxide in subsequent geological environments. (author)

  5. Status of Progress Made Toward Safety Analysis and Technical Site Evaluations for DOE Managed HLW and SNF.

    Energy Technology Data Exchange (ETDEWEB)

    Sevougian, S. David [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Stein, Emily [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Gross, Michael B [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Hammond, Glenn Edward [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Frederick, Jennifer M [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Mariner, Paul [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2016-11-01

    The Spent Fuel and Waste Science and Technology (SFWST) Campaign of the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE) is conducting research and development (R&D) on generic deep geologic disposal systems (i.e., repositories). This report describes specific activities in FY 2016 associated with the development of a Defense Waste Repository (DWR)a for the permanent disposal of a portion of the HLW and SNF derived from national defense and research and development (R&D) activities of the DOE.

  6. Solubility studies of Np(IV)

    International Nuclear Information System (INIS)

    Zhang Yingjie; Yao Jun; Jiao Haiyang; Ren Lihong; Zhou Duo; Fan Xianhua

    2001-01-01

    The solubility of Np(IV) in simulated underground water and redistilled water has been measured with the variations of pH(6-12) and storage time (0-100 d) in the presence of reductant (Na 2 S 2 O 4 , metallic Fe). All experiments are performed in a low oxygen concentration glove box containing high purity Ar(99.99%), with an oxygen content of less than 5 x 10 -6 mol/mol. Experimental results show that the variation of pH in solution has little effect on the solubility of Np(IV) in the two kinds of water; the measured solubility of Np(IV) is affected by the composition of solution; with Na 2 S 2 O 4 as a reductant, the solubility of Np(IV) in simulated underground water is (9.23 +- 0.48) x 10 -10 mol/L, and that in redistilled water is (8.31 +- 0.35) x 10 -10 mol/L; with metallic Fe as a reductant, the solubility of Np(IV) in simulated underground water is (1.85 +- 0.56) x 10 -9 mol/L, and that in redistilled water is (1.48 +- 0.66) x 10 -9 mol/L

  7. FINAL REPORT DM1200 TESTS WITH AZ 101 HLW SIMULANTS VSL-03R3800-4 REV 0 2/17/04

    Energy Technology Data Exchange (ETDEWEB)

    KRUGER AA; MATLACK KS; BARDAKCI T; D' ANGELO NA; GONG W; KOT WK; PEGG IL

    2011-12-29

    This report documents melter and off-gas performance results obtained on the DM 1200 HLW Pilot Melter during processing of simulated HLW AZ-101 feed. The principal objectives of the DM1200 melter testing were to determine the achievable glass production rates for simulated HLW AZ-101 feed; determine the effect of bubbling rate and feed solids content on production rate; characterize melter off-gas emissions; characterize the performance of the prototypical off-gas system components as well as their integrated performance; characterize the feed, glass product, and off-gas effluents; and to perform pre- and post-test inspections of system components. The test objectives (including test success criteria), along with how they were met, are outlined in a table.

  8. Final Report DM1200 Tests With AZ 101 HLW Simulants VSL-03R3800-4, Rev. 0, 2/17/04

    International Nuclear Information System (INIS)

    Kruger, A.A.; Matlack, K.S.; Bardakci, T.; D'Angelo, N.A.; Gong, W.; Kot, W.K.; Pegg, I.L.

    2011-01-01

    This report documents melter and off-gas performance results obtained on the DM 1200 HLW Pilot Melter during processing of simulated HLW AZ-101 feed. The principal objectives of the DM1200 melter testing were to determine the achievable glass production rates for simulated HLW AZ-101 feed; determine the effect of bubbling rate and feed solids content on production rate; characterize melter off-gas emissions; characterize the performance of the prototypical off-gas system components as well as their integrated performance; characterize the feed, glass product, and off-gas effluents; and to perform pre- and post-test inspections of system components. The test objectives (including test success criteria), along with how they were met, are outlined in a table.

  9. Soluble porphyrin polymers

    Science.gov (United States)

    Gust, Jr., John Devens; Liddell, Paul Anthony

    2015-07-07

    Porphyrin polymers of Structure 1, where n is an integer (e.g., 1, 2, 3, 4, 5, or greater) ##STR00001## are synthesized by the method shown in FIGS. 2A and 2B. The porphyrin polymers of Structure 1 are soluble in organic solvents such as 2-MeTHF and the like, and can be synthesized in bulk (i.e., in processes other than electropolymerization). These porphyrin polymers have long excited state lifetimes, making the material suitable as an organic semiconductor for organic electronic devices including transistors and memories, as well as solar cells, sensors, light-emitting devices, and other opto-electronic devices.

  10. Prediction of the solubility in lipidic solvent mixture: Investigation of the modeling approach and thermodynamic analysis of solubility.

    Science.gov (United States)

    Patel, Shruti V; Patel, Sarsvatkumar

    2015-09-18

    Self-micro emulsifying drug delivery system (SMEDDS) is one of the methods to improve solubility and bioavailability of poorly soluble drug(s). The knowledge of the solubility of pharmaceuticals in pure lipidic solvents and solvent mixtures is crucial for designing the SMEDDS of poorly soluble drug substances. Since, experiments are very time consuming, a model, which allows for solubility predictions in solvent mixtures based on less experimental data is desirable for efficiency. Solvents employed were Labrafil® M1944CS and Labrasol® as lipidic solvents; Capryol-90®, Capryol-PGMC® and Tween®-80 as surfactants; Transcutol® and PEG-400 as co-solvents. Solubilities of both drugs were determined in single solvent systems at temperature (T) range of 283-333K. In present study, we investigated the applicability of the thermodynamic model to understand the solubility behavior of drugs in the lipiodic solvents. By using the Van't Hoff and general solubility theory, the thermodynamic functions like Gibbs free energy, enthalpy and entropy of solution, mixing and solvation for drug in single and mixed solvents were understood. The thermodynamic parameters were understood in the framework of drug-solvent interaction based on their chemical similarity and dissimilarity. Clotrimazole and Fluconazole were used as active ingredients whose solubility was measured in single solvent as a function of temperature and the data obtained were used to derive mathematical models which can predict solubility in multi-component solvent mixtures. Model dependent parameters for each drug were calculated at each temperature. The experimental solubility data of solute in mixed solvent system were measured experimentally and further correlated with the calculates values obtained from exponent model and log-linear model of Yalkowsky. The good correlation was observed between experimental solubility and predicted solubility. Copyright © 2015 Elsevier B.V. All rights reserved.

  11. Vitrification operational experiences and lessons learned at the WVDP

    International Nuclear Information System (INIS)

    Hamel, W.F. Jr.; Sheridan, M.J.; Valenti, P.J.

    1997-01-01

    The Vitrification Facility (VF) at the West Valley Demonstration Project (WVDP) commenced full, high-level radioactive waste (HLW) processing activities in July 1996. The HLW consists of a blend of washed plutonium-uranium extraction (PUREX) sludge, neutralized thorium extraction (THOREX) waste, and cesium-loaded zeolite. The waste product is borosilicate glass contained in stainless steel canisters, sealed for eventual disposal in a federal repository. This paper discusses the WVDP vitrification process, focusing on operational experience and lessons learned during the first year of continuous, remote operation

  12. A methodology of uncertainty/sensitivity analysis for PA of HLW repository learned from 1996 WIPP performance assessment

    International Nuclear Information System (INIS)

    Lee, Y. M.; Kim, S. K.; Hwang, Y. S.; Kang, C. H.

    2002-01-01

    The WIPP (Waste Isolation Pilot Plant) is a mined repository constructed by the US DOE for the permanent disposal of transuranic (TRU) wastes generated by activities related to defence of the US since 1970. Its historical disposal operation began in March 1999 following receipt of a final permit from the State of NM after a positive certification decision for the WIPP was issued by the EPA in 1998, as the first licensed facility in the US for the deep geologic disposal of radioactive wastes. The CCA (Compliance Certification Application) for the WIPP that the DOE submitted to the EPA in 1966 was supported by an extensive Performance Assessment (PA) carried out by Sandia National Laboratories (SNL), with so-called 1996 PA. Even though such PA methodologies could be greatly different from the way we consider for HLW disposal in Korea largely due to quite different geologic formations in which repository are likely to be located, a review on lots of works done through the WIPP PA studies could be the most important lessons that we can learn from in view of current situation in Korea where an initial phase of conceptual studies on HLW disposal has been just started. The objective of this work is an overview of the methodology used in the recent WIPP PA to support the US DOE WIPP CCA ans a proposal for Korean case

  13. Thermo-mechanical analysis for multi-level HLW repository concept

    International Nuclear Information System (INIS)

    Kwon, Sang Ki; Choi, Jong Won

    2004-01-01

    This work aims to investigate the influence of design parameters for the underground high-level nuclear waste repository with multi-level concept. B. Necessity o In order to construct an HLW repository in deep underground, it is required to select a site, which is far from major discontinuities. To dispose the whole spent fuels generated from the Korean nuclear power plants in a repository, the underground area of about 4km 2 is required. This would be a constraints for selecting an adequate repository site. It is recommended to dispose the two different spent fuels, PWR and CANDU, in different areas at the operation efficiency point of view. It is necessary to investigate the influence of parameters, which can affect the stability of multi-level repository. It is also needed to consider the influence of heat generated from the HLW and the high in situ stress in deep location. Therefore, thermo-mechanical coupling analysis should be carried out and the results should be compared with the results from single-level repository concept. Three-dimensional analysis is required to model the disposal tunnel and deposition hole. It is recommended to use the Korean geological condition and actually measured rock properties in Korea in order to achieve reliable modeling results. A FISH routine developed for effective modeling of Thermal-Mechanical coupling was implemented in the modeling using FLAC3D, which is a commercial three-dimensional FDM code. The thermal and mechanical properties of rock and rock mass achieved from Yusung drilling site, were used for the computer modeling. Different parameters such as level distance, waste type disposed on different levels, and time interval between the operation on different levels, were considered in the three-dimensional analysis. From the analysis, it was possible to derive adequate multi-level repository concept. Results and recommendations for application From the thermal-mechanical analysis for the multi-level repository

  14. Determination of radionuclide solubility limits to be used in SR 97. Uncertainties associated to calculated solubilities

    Energy Technology Data Exchange (ETDEWEB)

    Bruno, J.; Cera, E.; Duro, L.; Jordana, S. [QuantiSci S.L., Barcelona (Spain); Pablo, J. de [DEQ-UPC, Barcelona (Spain); Savage, D. [QuantiSci Ltd., Henley-on-Thames (United Kingdom)

    1997-12-01

    The thermochemical behaviour of 24 critical radionuclides for the forthcoming SR97 PA exercise is discussed. The available databases are reviewed and updated with new data and an extended database for aqueous and solid species of the radionuclides of interest is proposed. We have calculated solubility limits for the radionuclides of interest under different groundwater compositions. A sensitivity analysis of the calculated solubilities with the composition of the groundwater is presented. Besides selecting the most likely solubility limiting phases, in this work we have used coprecipitation approaches in order to calculate more realistic solubility limits for minor radionuclides, such as Ra, Am and Cm. The comparison between the calculated solubilities and the concentrations measured in relevant natural systems (NA) and in spent fuel leaching experiments helps to assess the validity of the methodology used and to derive source term concentrations for the radionuclides studied. The uncertainties associated to the solubilities of the main radionuclides involved in the spent nuclear fuel have also been discussed in this work. The variability of the groundwater chemistry; redox conditions and temperature of the system have been considered the main factors affecting the solubilities. In this case, a sensitivity analysis has been performed in order to study solubility changes as a function of these parameters. The uncertainties have been calculated by including the values found in a major extent in typical granitic groundwaters. The results obtained from this analysis indicate that there are some radionuclides which are not affected by these parameters, i.e. Ag, Cm, Ho, Nb, Ni, Np, Pu, Se, Sm, Sn, Sr, Tc and U

  15. An estimation of influence of humic acid and organic matter originated from bentonite on samarium solubility

    International Nuclear Information System (INIS)

    Kanaji, Mariko; Sato, Haruo; Sasahira, Akira

    1999-10-01

    Organic acids in groundwater are considered to form complexes and increase the solubility of radionuclides released from vitrified waste in a high-level radioactive waste (HLW) repository. To investigate whether the solubility of samarium (Sm) is influenced by organic substances, we measured Sm solubility in the presence of different organic substances and compared those values with results from thermodynamic predictions. Humic acid (Aldrich) is commercially available and soluble organic matter originated from bentonite were used as organic substances in this study. Consequently, the solubility of Sm showed a tendency to apparently increase with increasing the concentration of humic acid, but in the presence of carbonate, thermodynamic predictions suggested that the dominant species are carbonate complexes and that the effect of organic substances are less than that of carbonate. Based on total organic carbon (TOC), the increase of Sm solubility measured with humic acid (Aldrich) was more significant than that in the case with soluble organic matter originated from bentonite. Since bentonite is presumed to include also simple organic matters of which stability constant for forming complexes is low, the effect of soluble organic matter originated from bentonite on the solubility of Sm is considered to be less effective than that of humic acid (Aldrich). Experimental values were compared with model prediction, proposed by Kim, based on data measured in a low pH region. Tentatively we calculated the increase in Sm solubility assuming complexation with humic acid. Trial calculations were carried out on the premise that the complexation reaction of metal ion with humic acid is based on neutralization process by 1-1 complexation. In this process, it was assumed that one metal ion coordinates with one unit of complexation sites which number of proton exchange sites is equal to ionic charge. Consequently, Kim's model indicated that carbonate complexes should be dominant

  16. Benefits Of Vibration Analysis For Development Of Equipment In HLW Tanks - 12341

    International Nuclear Information System (INIS)

    Stefanko, D.; Herbert, J.

    2012-01-01

    Vibration analyses of equipment intended for use in the Savannah River Site (SRS) radioactive liquid waste storage tanks are performed during pre-deployment testing and has been demonstrated to be effective in reducing the life-cycle costs of the equipment. Benefits of using vibration analysis to identify rotating machinery problems prior to deployment in radioactive service will be presented in this paper. Problems encountered at SRS and actions to correct or lessen the severity of the problem are discussed. In short, multi-million dollar cost saving have been realized at SRS as a direct result of vibration analysis on existing equipment. Vibration analysis of equipment prior to installation can potentially reduce inservice failures, and increases reliability. High-level radioactive waste is currently stored in underground carbon steel waste tanks at the United States Department of Energy (DOE) Savannah River Site and at the Hanford Site, WA. Various types of rotating machinery (pumps and separations equipment) are used to manage and retrieve the tank contents. Installation, maintenance, and repair of these pumps and other equipment are expensive. In fact, costs to remove and replace a single pump can be as high as a half million dollars due to requirements for radioactive containment. Problems that lead to in-service maintenance and/or equipment replacement can quickly exceed the initial investment, increase radiological exposure, generate additional waste, and risk contamination of personnel and the work environment. Several different types of equipment are considered in this paper, but pumps provide an initial example for the use of vibration analysis. Long-shaft (45 foot long) and short-shaft (5-10 feet long) equipment arrangements are used for 25-350 horsepower slurry mixing and transfer pumps in the SRS HLW tanks. Each pump has a unique design, operating characteristics and associated costs, sometimes exceeding a million dollars. Vibration data are routinely

  17. BENEFITS OF VIBRATION ANALYSIS FOR DEVELOPMENT OF EQUIPMENT IN HLW TANKS - 12341

    Energy Technology Data Exchange (ETDEWEB)

    Stefanko, D.; Herbert, J.

    2012-01-10

    Vibration analyses of equipment intended for use in the Savannah River Site (SRS) radioactive liquid waste storage tanks are performed during pre-deployment testing and has been demonstrated to be effective in reducing the life-cycle costs of the equipment. Benefits of using vibration analysis to identify rotating machinery problems prior to deployment in radioactive service will be presented in this paper. Problems encountered at SRS and actions to correct or lessen the severity of the problem are discussed. In short, multi-million dollar cost saving have been realized at SRS as a direct result of vibration analysis on existing equipment. Vibration analysis of equipment prior to installation can potentially reduce inservice failures, and increases reliability. High-level radioactive waste is currently stored in underground carbon steel waste tanks at the United States Department of Energy (DOE) Savannah River Site and at the Hanford Site, WA. Various types of rotating machinery (pumps and separations equipment) are used to manage and retrieve the tank contents. Installation, maintenance, and repair of these pumps and other equipment are expensive. In fact, costs to remove and replace a single pump can be as high as a half million dollars due to requirements for radioactive containment. Problems that lead to in-service maintenance and/or equipment replacement can quickly exceed the initial investment, increase radiological exposure, generate additional waste, and risk contamination of personnel and the work environment. Several different types of equipment are considered in this paper, but pumps provide an initial example for the use of vibration analysis. Long-shaft (45 foot long) and short-shaft (5-10 feet long) equipment arrangements are used for 25-350 horsepower slurry mixing and transfer pumps in the SRS HLW tanks. Each pump has a unique design, operating characteristics and associated costs, sometimes exceeding a million dollars. Vibration data are routinely

  18. Execution techniques and approach for high level radioactive waste disposal in Japan: Demonstration of geological disposal techniques and implementation approach of HLW project

    International Nuclear Information System (INIS)

    Kawanishi, M.; Komada, H.; Kitayama, K.; Akasaka, H.; Tsuchi, H.

    2001-01-01

    In Japan, the high-level radioactive waste (HLW) disposal project is expected to start fully after establishment of the implementing organization, which is planned around the year 2000 and to dispose the wastes in the 2030s to at latest in the middle of 2040s. Considering each step in the implementation of the HLW disposal project in Japan, this paper discusses the execution procedure for HLW disposal project, such as the selection of candidate/planned disposal sites, the construction and operation of the disposal facility, the closure and decommissioning of facilities, and the institutional control and monitoring after the closure of disposal facility, from a technical viewpoint for the rational execution of the project. Furthermore, we investigate and propose some ideas for the concept of the design of geological disposal facility, the validation and demonstration of the reliability on the disposal techniques and performance assessment methods at a candidate/planned site. Based on these investigation results, we made clear a milestone for the execution of the HLW disposal project in Japan. (author)

  19. Thermodynamic data development using the solubility method (Joint research)

    International Nuclear Information System (INIS)

    Rai, Dhanpat; Yui, Mikazu

    2013-05-01

    The solubility method is one of the most powerful tools to obtain reliable thermodynamic data for 1) solubility products of discrete solids and double salts, 2) complexation constants for various ligands, 3) development of data in a wide range of pH values, 4) evaluation of data for metals that form very insoluble solids (e.g. tetravalent actinides), 5) determining solubility-controlling solids in different types of wastes and 6) elevated temperatures for redox sensitive systems. This document is focused on describing various aspects of obtaining thermodynamic data using the solubility method. This manuscript deals with various aspects of conducting solubility studies, including selecting the study topic, modeling to define important variables, selecting the range of variables and experimental parameters, anticipating results, general equipment requirements, conducting experiments, and interpreting experimental data. (author)

  20. Water-soluble vitamins.

    Science.gov (United States)

    Konings, Erik J M

    2006-01-01

    Simultaneous Determination of Vitamins.--Klejdus et al. described a simultaneous determination of 10 water- and 10 fat-soluble vitamins in pharmaceutical preparations by liquid chromatography-diode-array detection (LC-DAD). A combined isocratic and linear gradient allowed separation of vitamins in 3 distinct groups: polar, low-polar, and nonpolar. The method was applied to pharmaceutical preparations, fortified powdered drinks, and food samples, for which results were in good agreement with values claimed. Heudi et al. described a separation of 9 water-soluble vitamins by LC-UV. The method was applied for the quantification of vitamins in polyvitaminated premixes used for the fortification of infant nutrition products. The repeatability of the method was evaluated at different concentration levels and coefficients of variation were based on, for example, LC. Koontz et al. showed results of total folate concentrations measured by microbiological assay in a variety of foods. Samples were submitted in a routine manner to experienced laboratories that regularly perform folate analysis fee-for-service basis in the United States. Each laboratory reported the use of a microbiological method similar to the AOAC Official Method for the determination of folic acid. Striking was, the use of 3 different pH extraction conditions by 4 laboratories. Only one laboratory reported using a tri-enzyme extraction. Results were evaluated. Results for folic acid fortified foods had considerably lower between-laboratory variation, 9-11%, versus >45% for other foods. Mean total folate ranged from 14 to 279 microg/100 g for a mixed vegetable reference material, from 5 to 70 microg/100 g for strawberries, and from 28 to 81 microg/100 g for wholemeal flour. One should realize a large variation in results, which might be caused by slight modifications in the microbiological analysis of total folate in foods or the analysis in various (unfortified) food matrixes. Furthermore, optimal

  1. Students’ misconceptions on solubility equilibrium

    Science.gov (United States)

    Setiowati, H.; Utomo, S. B.; Ashadi

    2018-05-01

    This study investigated the students’ misconceptions of the solubility equilibrium. The participants of the study consisted of 164 students who were in the science class of second year high school. Instrument used is two-tier diagnostic test consisting of 15 items. Responses were marked and coded into four categories: understanding, misconception, understand little without misconception, and not understanding. Semi-structured interviews were carried out with 45 students according to their written responses which reflected different perspectives, to obtain a more elaborated source of data. Data collected from multiple methods were analyzed qualitatively and quantitatively. Based on the data analysis showed that the students misconceptions in all areas in solubility equilibrium. They had more misconceptions such as in the relation of solubility and solubility product, common-ion effect and pH in solubility, and precipitation concept.

  2. On the americium oxalate solubility

    International Nuclear Information System (INIS)

    Zakolupin, S.A.; Korablin, Eh.V.

    1977-01-01

    The americium oxalate solubility at different nitric (0.0-1 M) and oxalic (0.0-0.4 M) acid concentrations was investigated in the temperature range from 14 to 60 deg C. The dependence of americium oxalate solubility on the oxalic acid concentration was determined. Increasing oxalic acid concentration was found to reduce the americium oxalate solubility. The dependence of americium oxalate solubility on the oxalic acid concentration was noted to be a minimum at low acidity (0.1-0.3 M nitric acid). This is most likely due to Am(C 2 O 4 ) + , Am(C 2 O 4 ) 2 - and Am(C 2 O 4 ) 3 3- complex ion formation which have different unstability constants. On the basis of the data obtained, a preliminary estimate was carried out for the product of americium oxalate solubility in nitric acid medium (10 -29 -10 -31 ) and of the one in water (6.4x10 -20 )

  3. Initiating the Validation of CCIM Processability for Multi-phase all Ceramic (SYNROC) HLW Form: Plan for Test BFY14CCIM-C

    Energy Technology Data Exchange (ETDEWEB)

    Maio, Vince [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-08-01

    This plan covers test BFY14CCIM-C which will be a first–of–its-kind demonstration for the complete non-radioactive surrogate production of multi-phase ceramic (SYNROC) High Level Waste Forms (HLW) using Cold Crucible Induction Melting (CCIM) Technology. The test will occur in the Idaho National Laboratory’s (INL) CCIM Pilot Plant and is tentatively scheduled for the week of September 15, 2014. The purpose of the test is to begin collecting qualitative data for validating the ceramic HLW form processability advantages using CCIM technology- as opposed to existing ceramic–lined Joule Heated Melters (JHM) currently producing BSG HLW forms. The major objectives of BFY14CCIM-C are to complete crystalline melt initiation with a new joule-heated resistive starter ring, sustain inductive melting at temperatures between 1600 to 1700°C for two different relatively high conductive materials representative of the SYNROC ceramic formation inclusive of a HLW surrogate, complete melter tapping and pouring of molten ceramic material in to a preheated 4 inch graphite canister and a similar canister at room temperature. Other goals include assessing the performance of a new crucible specially designed to accommodate the tapping and pouring of pure crystalline forms in contrast to less recalcitrant amorphous glass, assessing the overall operational effectiveness of melt initiation using a resistive starter ring with a dedicated power source, and observing the tapped molten flow and subsequent relatively quick crystallization behavior in pans with areas identical to standard HLW disposal canisters. Surrogate waste compositions with ceramic SYNROC forming additives and their measured properties for inductive melting, testing parameters, pre-test conditions and modifications, data collection requirements, and sampling/post-demonstration analysis requirements for the produced forms are provided and defined.

  4. Thermo-hydro-mechanical processes in the nearfield around a HLW repository in argillaceous formations. Vol. I. Laboratory investigations

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Chun-Liang; Czaikowski, Oliver; Rothfuchs, Tilmann; Wieczorek, Klaus

    2013-06-15

    All over the world, clay formations are being investigated as host medium for geologic disposal of radioactive waste because of their favourable properties, such as very low hydraulic conductivity against fluid transport, good sorption capacity for retardation of radionuclides, and high potential of self-sealing of fractures. The construction of a repository, the disposal of heat-emitting high-level radioactive waste (HLW), the backfilling and sealing of the remaining voids, however, will inevitably induce mechanical (M), hydraulic (H), thermal (T) and chemical (C) disturbances to the host formation and the engineered barrier system (EBS) over very long periods of time during the operation and post-closure phases of the repository. The responses and resulting property changes of the clay host rock and engineered barriers are to be well understood, characterized, and predicted for assessing the long-term performance and safety of the repository.

  5. Study on a transportation and emplacement system of pre-assembled EBS module for HLW geological disposal

    International Nuclear Information System (INIS)

    Awano, Toshihiko; Kanno, Takeshi; Katsumata, Syunsuke; Kosuge, Kazuhiro

    2009-01-01

    HLW disposal is one of the largest issue to utilize Nuclear power safely. In the past study, the concept, which buffer materials and Overpacked waste were transported into underground respectively, have shown. The concept of pre-assembled engineered barrier has advantage to simplify the logistics and emplacement procedure, however there are difficulties to support heavy weight of pre-assembled package by equipment under the condition of little clearance between tunnel and package. In this study, Combination of air bearing and two degree-of-freedom wheels were suggested for transportation, and air jack was suggested for unloading and emplacement system. Also, whole system for transportation and emplacement procedure was designed, and Scale model test was examined to evaluate the feasibility of these concept and functions. (author)

  6. Alternative biosphere modeling for safety assessment of HLW disposal taking account of geosphere-biosphere interface of marine environment

    International Nuclear Information System (INIS)

    Kato, Tomoko; Ishiguro, Katsuhiko; Naito, Morimasa; Ikeda, Takao; Little, Richard

    2001-03-01

    In the safety assessment of a high-level radioactive waste (HLW) disposal system, it is required to estimate radiological impacts on future human beings arising from potential radionuclide releases from a deep repository into the surface environment. In order to estimated the impacts, a biosphere model is developed by reasonably assuming radionuclide migration processes in the surface environment and relevant human lifestyles. It is important to modify the present biosphere models or to develop alternative biosphere models applying the biosphere models according to quality and quantify of the information acquired through the siting process for constructing the repository. In this study, alternative biosphere models were developed taking geosphere-biosphere interface of marine environment into account. Moreover, the flux to dose conversion factors calculated by these alternative biosphere models was compared with those by the present basic biosphere models. (author)

  7. A comparison of three methods for determining the amount of nitric acid needed to treat HLW sludge at SRS

    International Nuclear Information System (INIS)

    Siegwald, S.F.; Ferrara, D.M.

    1994-01-01

    A comparison was made of three methods for determining the amount of nitric acid which will be needed to treat a sample of high-level waste (HLW) sludge from the Savannah River Site (SRS) Tank Farm. The treatment must ensure the resulting melter feed will have the necessary rheological and oxidation-reduction properties, reduce mercury and manganese in the sludge, and be performed in a fashion which does not produce a flammable gas mixture. The three methods examined where an empirical method based on pH measurements, a computational method based on known reactions of the species in the sludge and a titration based on neutralization of carbonate in the solution

  8. Microcrack growing and long-term mechanical stability in a HLW deep-borehole repository in granite

    International Nuclear Information System (INIS)

    Biurrun, E.; Hahne, K.

    1989-01-01

    The long-term host rock integrity assessment of a deep borehole emplacement for HLW in granite has been addressed with a detailed new constitutive model considering temperature and pressure effects on microscale phenomena (as microcracking) under repository conditions. The results of these finite element calculations have been compared with results obtained using conventional, state-of-the-art constitutive modelling. While the results of conventional modelling did suggest the existence of an important safety margin before failure, the improved calculations with the new model predict a thin but very long region of degradated host rock along the waste canister column. The results obtained up to now may well be considered as safety relevant, because they suggest that the actual long-term granite strength lies well below the conventionally determined failure limits, thus challenging the barrier properties of this host rock if the actual strength is not properly considered in the repository design

  9. Molybdenum solubility in aluminium nitrate solutions

    Energy Technology Data Exchange (ETDEWEB)

    Heres, X.; Sans, D.; Bertrand, M.; Eysseric, C. [CEA, Centre de Marcoule, Nuclear Energy Division, DRCP, BP 17171, 30207 Bagnols-sur-Ceze Cedex (France); Brackx, E.; Domenger, R.; Excoffier, E. [CEA, Centre de Marcoule, Nuclear Energy Division, DTEC, BP 17171, 30207 Bagnols-sur-Ceze Cedex (France); Valery, J.F. [AREVA-NC, DOR/RDP, Paris - La Defense (France)

    2016-07-01

    For over 60 years, research reactors (RR or RTR for research testing reactors) have been used as neutron sources for research, radioisotope production ({sup 99}Mo/{sup 99m}Tc), nuclear medicine, materials characterization, etc... Currently, over 240 of these reactors are in operation in 56 countries. They are simpler than power reactors and operate at lower temperature (cooled to below 100 C. degrees). The fuel assemblies are typically plates or cylinders of uranium alloy and aluminium (U-Al) coated with pure aluminium. These fuels can be processed in AREVA La Hague plant after batch dissolution in concentrated nitric acid and mixing with UOX fuel streams. The aim of this study is to accurately measure the solubility of molybdenum in nitric acid solution containing high concentrations of aluminium. The higher the molybdenum solubility is, the more flexible reprocessing operations are, especially when the spent fuels contain high amounts of molybdenum. To be most representative of the dissolution process, uranium-molybdenum alloy and molybdenum metal powder were dissolved in solutions of aluminium nitrate at the nominal dissolution temperature. The experiments showed complete dissolution of metallic elements after 30 minutes long stirring, even if molybdenum metal was added in excess. After an induction period, a slow precipitation of molybdic acid occurs for about 15 hours. The data obtained show the molybdenum solubility decreases with increasing aluminium concentration. The solubility law follows an exponential relation around 40 g/L of aluminium with a high determination coefficient. Molybdenum solubility is not impacted by the presence of gadolinium, or by an increasing concentration of uranium. (authors)

  10. Operational safety and radiation protection considerations in designing an HLW repository in Germany

    International Nuclear Information System (INIS)

    Filbert, W.; Kreienmeyer, M.; Poehler, M.; Niehues, N.

    2008-01-01

    In Germany the reference concept for disposal of heat generating radioactive waste considers emplacing canisters with vitrified waste in deep vertical boreholes drilled from the drifts of a repository mine in salt at a depth of 870 m. Spent fuel is to be disposed of in self-shielding POLLUX casks in horizontal drifts. An optimized disposal concept anticipates emplacing unshielded canisters with vitrified HLW and canisters containing the fuel rods of 3 PWR or 9 BWR fuel assemblies in boreholes with a diameter of 60 cm and a depth of up to 300 m.. In all cases the void space between POLLUX cask and drifts and canisters and borehole wall will be backfilled with crushed salt. (1) Operational Safety: Based on a detailed description of all underground disposal operation steps, the possible impacts on the disposal operations were analysed and the need for further studies determined. The disposal operation steps comprise e.g. rail bound transport from the shaft to the emplacement drift and emplacement process itself. As possible impacts the following occurrences were considered: ventilation failure, power supply failure, rock mechanics impact including cross-section convergence, irregular floor uplift and rock fall, brine and natural gas intrusion, derailing of transport carts and finally internal fire. (2) Radiation Protection: According to the German Atomic Energy Act (AtG), the design, construction and operation of a nuclear site like a final repository has to be licensed by the responsible authority. The Radiological Protection Ordinance and further guidelines i.e. concerning the emission and immission of released radioactive nuclides or the risk analysis of possible failure, build the basis for the licensing procedures. To ensure adequate protection against undue radiation exposure the repository is divided into different radiological protection areas. Generally, the handling of shielded waste packages above und under ground (including all the pathway of transport and

  11. Uranium solubility and solubility controls in selected Needle's Eye groundwaters

    International Nuclear Information System (INIS)

    Falck, W.E.; Hooker, P.J.

    1991-01-01

    The solubility control of uranium in selected groundwater samples from the cliff and sediments at the Needle's Eye natural analogue site is investigated using the speciation code PHREEQE and the CHEMVAL thermodynamic database (release 3). Alkali-earth bearing uranyl carbonate secondary minerals are likely to exert influence on the solubility . Other candidates are UO 2 and arsenates, depending on the prevailing redox conditions. In the absence of literature data, solubility products for important arsenates have been estimated from analogy with other arsenates and phosphates. Phosphates themselves are unlikely to exert control owing to their comparatively high solubilities. The influence of seawater flooding into the sediments is also discussed. The importance of uranyl arsenates in the retardation of uranium in shallow sediments has been demonstrated in theory, but there are some significant gaps in the thermodynamic databases used. (author)

  12. Noble gases solubility in water

    International Nuclear Information System (INIS)

    Crovetto, Rosa; Fernandez Prini, Roberto.

    1980-07-01

    The available experimental data of solubility of noble gases in water for temperatures smaller than 330 0 C have been critically surveyed. Due to the unique structure of the solvent, the solubility of noble gases in water decreases with temperature passing through a temperature of minimum solubility which is different for each gas, and then increases at higher temperatures. As aresult of the analysis of the experimental data and of the features of the solute-solvent interaction, a generalized equation is proposed which enables thecalculation of Henry's coefficient at different temperatures for all noble gases. (author) [es

  13. The effects of disordered structure on the solubility and dissolution rates of some hydrophilic, sparingly soluble drugs.

    Science.gov (United States)

    Mosharraf, M; Sebhatu, T; Nyström, C

    1999-01-15

    The effects of experimental design on the apparent solubility of two sparingly soluble hydrophilic compounds (barium sulphate and calcium carbonate) were studied in this paper. The apparent solubility appeared to be primarily dependent on the amount of solute added to the solvent in each experiment, increasing with increased amounts. This effect seems to be due to the existence of a peripheral disordered layer. However physico-chemical methods used in the present study were not able to unambiguously verify the existence of any disorder in the solid state structure of the drugs. At higher proportions of solute to solvent, the solubility reached a plateau corresponding to the solubility of the disordered or amorphous molecular form of the material. Milling the powders caused the plateau to be reached at lower proportions of solute to solvent, since this further disordered the surface of the drug particles. It was also found that the apparent solubility of the drugs tested decreased after storage at high relative humidities. A model for describing the effects of a disordered surface layer of varying thickness and continuity on the solubility of a substance is presented. This model may be used as a method for detection of minute amount of disorder, where no other technique is capable of detecting the disordered structure. It is suggested that recrystallisation of the material occurs via slow solid-state transition at the surface of the drug particle; this would slowly reduce the apparent solubility of the substance at the plateau level to the thermodynamically stable value. A biphasic dissolution rate profile was obtained. The solubility of the disordered surface of the particles appeared to be the rate-determining factor during the initial dissolution phase, while the solubility of the crystalline core was the rate-determining factor during the final slower phase.

  14. The role of quality management in safety case development - Nagra's experience

    International Nuclear Information System (INIS)

    Schneider, Juerg W.; Zuidema, Piet

    2014-01-01

    This paper discusses the role of quality management (QM) in safety case development based on Nagra's experience from a broad range of projects. These include Project Gewahr (L/ILW and HLW, Nagra, 1985), the Wellenberg Project (L/ILW, Nagra, 1994), Project Opalinus Clay (HLW, Nagra, 2002a, 2002b), and recent project work needed in the context of the Swiss site selection process (L/ILW and HLW, Nagra, 2008a, 2008b, 2008c, 2010). Broadly speaking, Nagra's Quality Management policy is focused on ensuring: i) the quality of the disposal system (siting, design and implementation); ii) the quality of the underlying scientific understanding, which are seen as key elements of a credible safety case, along with the quality of the safety calculations themselves and of compiling the safety case, including the drawing of conclusions (Nagra, 2002a). All aspects of QM discussed in this paper should be seen in this context. (authors)

  15. Operating experience during high-level waste vitrification at the West Valley Demonstration Project

    International Nuclear Information System (INIS)

    Valenti, P.J.; Elliott, D.I.

    1999-01-01

    This report provides a summary of operational experiences, component and system performance, and lessons learned associated with the operation of the Vitrification Facility (VF) at the West Valley Demonstration Project (WVDP). The VF was designed to convert stored high-level radioactive waste (HLW) into a stable waste form (borosilicate glass) suitable for disposal in a federal repository. Following successful completion on nonradioactive test, HLW processing began in July 1995. Completion of Phase 1 of HLW processing was reached on 10 June 1998 and represented the processing of 9.32 million curies of cesium-137 (Cs-137) and strontium-90 (Sr-90) to fill 211 canisters with over 436,000 kilograms of glass. With approximately 85% of the total estimated curie content removed from underground waste storage tanks during Phase 1, subsequent operations will focus on removal of tank heel wastes

  16. Properties and solubility of chrome in iron alumina phosphate glasses containing high level nuclear waste

    International Nuclear Information System (INIS)

    Huang, W.; Day, D.E.; Ray, C.S.; Kim, C.W.; Reis, S.T.D.

    2004-01-01

    Chemical durability, glass formation tendency, and other properties of iron alumina phosphate glasses containing 70 wt% of a simulated high level nuclear waste (HLW), doped with different amounts of Cr 2 O 3 , have been investigated. All of the iron alumina phosphate glasses had an outstanding chemical durability as measured by their small dissolution rate (1 . 10 -9 g/(cm 2 . min)) in deionized water at 90 C for 128 d, their low normalized mass release as determined by the product consistency test (PCT) and a barely measurable corrosion rate of 2 . d) after 7 d at 200 C by the vapor hydration test (VHT). The solubility limit for Cr 2 O 3 in the iron phosphate melts was estimated at 4.1 wt%, but all of the as-annealed melts contained a few percent of crystalline Cr 2 O 3 that had no apparent effect on the chemical durability. The chemical durability was unchanged after deliberate crystallization, 48 h at 650 C. These iron phosphate waste forms, with a waste loading of at least 70 wt%, can be readily melted in commercial refractory crucibles at 1250 C for 2 to 4 h, are resistant to crystallization, meet all current US Department of Energy requirements for chemical durability, and have a solubility limit for Cr 2 O 3 which is at least three times larger than that for borosilicate glasses. (orig.)

  17. Final Report - Effects of High Spinel and Chromium Oxide Crystal Contents on Simulated HLW Vitrification in DM100 Melter Tests, VSL-09R1520-1, Rev. 0, dated 6/22/09

    Energy Technology Data Exchange (ETDEWEB)

    Kruger, Albert A.; Matlack, K. S.; Kot, W.; Pegg, I. L.; Chaudhuri, M.; Lutze, W.

    2013-11-13

    The principal objective of the work was to evaluate the effects of spinel and chromium oxide particles on WTP HLW melter operations and potential impacts on melter life. This was accomplished through a combination of crucible-scale tests, settling and rheological tests, and tests on the DM100 melter system. Crucible testing was designed to develop and identify HLW glass compositions with high waste loadings that exhibit formation of crystalline spinel and/or chromium oxide phases up to relatively high crystal contents (i.e., > 1 vol%). Characterization of crystal settling and the effects on melt rheology was performed on the HLW glass formulations. Appropriate candidate HLW glass formulations were selected, based on characterization results, to support subsequent melter tests. In the present work, crucible melts were formulated that exhibit up to about 4.4 vol% crystallization.

  18. Hydrogen adsorption on and solubility in graphites

    International Nuclear Information System (INIS)

    Kanashenko, S.L.; Wampler, W.R.

    1996-01-01

    The experimental data on adsorption and solubility of hydrogen isotopes in graphite over a wide range of temperatures and pressures are reviewed. Langmuir adsorption isotherms are proposed for the hydrogen-graphite interaction. The entropy and enthalpy of adsorption are estimated, allowing for effects of relaxation of dangling sp 2 bonds. Three kinds of traps are proposed: edge carbon atoms of interstitial loops with an adsorption enthalpy relative to H 2 gas of -4.4 eV/H 2 (unrelaxed, Trap 1), edge carbon atoms at grain surfaces with an adsorption enthalpy of -2.3 eV/H 2 (relaxed, Trap 2), and basal plane adsorption sites with an enthalpy of +2.43 eV/H 2 (Trap 3). The adsorption capacity of different types of graphite depends on the concentration of traps which depends on the crystalline microstructure of the material. The number of potential sites for the 'true solubility' (Trap 3) is assumed to be about one site per carbon atom in all types of graphite, but the endothermic character of this solubility leads to a negligible H inventory compared to the concentration of hydrogen in type 1 and type 2 traps for temperatures and gas pressures used in the experiments. Irradiation with neutrons or carbon atoms increases the concentration of type 1 and type 2 traps from about 20 and 200 appm respectively for unirradiated (POCO AXF-5Q) graphite to about 1500 and 5000 appm, respectively, at damage levels above 1 dpa. (orig.)

  19. Grouping of HLW in partitioning for B/T (burning and/or transmutation) treatment with neutron reactors based on three criteria

    International Nuclear Information System (INIS)

    Kitamoto, Mulyanto; Kitamoto, Asashi

    1995-01-01

    A grouping concept of HLW in partitioning for B/T (burning and/or transmutation) treatment by fission reactor was developed in order to improve the disposal in waste management from the safety aspect. The selecting and grouping concept was proposed herein, such as Group MA1 (Np, Am, and unrecovered U and Pu), Group MA2 (Cm, and trace quantity of Cf, etc.), Group A (Tc and I), Group B (Cs and Sr) and Group R (the partitioned remains of HLW), judging from the three criteria for B/T treatment, based on (1) the concept of the potential risk estimated by the hazard index for long-term tendency based on ALI (2) the concept of the relative dose factor related to the adsorbed migration rate transferred through ground water, and (3) the concept of the decay acceleration factor, the burning and/or transmutation characteristics for recycle B/T treatment. (author)

  20. Enhancing the solubility and bioavailability of poorly water-soluble drugs using supercritical antisolvent (SAS) process.

    Science.gov (United States)

    Abuzar, Sharif Md; Hyun, Sang-Min; Kim, Jun-Hee; Park, Hee Jun; Kim, Min-Soo; Park, Jeong-Sook; Hwang, Sung-Joo

    2018-03-01

    Poor water solubility and poor bioavailability are problems with many pharmaceuticals. Increasing surface area by micronization is an effective strategy to overcome these problems, but conventional techniques often utilize solvents and harsh processing, which restricts their use. Newer, green technologies, such as supercritical fluid (SCF)-assisted particle formation, can produce solvent-free products under relatively mild conditions, offering many advantages over conventional methods. The antisolvent properties of the SCFs used for microparticle and nanoparticle formation have generated great interest in recent years, because the kinetics of the precipitation process and morphologies of the particles can be accurately controlled. The characteristics of the supercritical antisolvent (SAS) technique make it an ideal tool for enhancing the solubility and bioavailability of poorly water-soluble drugs. This review article focuses on SCFs and their properties, as well as the fundamentals of overcoming poorly water-soluble drug properties by micronization, crystal morphology control, and formation of composite solid dispersion nanoparticles with polymers and/or surfactants. This article also presents an overview of the main aspects of the SAS-assisted particle precipitation process, its mechanism, and parameters, as well as our own experiences, recent advances, and trends in development. Copyright © 2017 Elsevier B.V. All rights reserved.

  1. Iron solubility related to particle sulfur content in source emission and ambient fine particles.

    Science.gov (United States)

    Oakes, M; Ingall, E D; Lai, B; Shafer, M M; Hays, M D; Liu, Z G; Russell, A G; Weber, R J

    2012-06-19

    The chemical factors influencing iron solubility (soluble iron/total iron) were investigated in source emission (e.g., biomass burning, coal fly ash, mineral dust, and mobile exhaust) and ambient (Atlanta, GA) fine particles (PM2.5). Chemical properties (speciation and mixing state) of iron-containing particles were characterized using X-ray absorption near edge structure (XANES) spectroscopy and micro-X-ray fluorescence measurements. Bulk iron solubility (soluble iron/total iron) of the samples was quantified by leaching experiments. Major differences were observed in iron solubility in source emission samples, ranging from low solubility (iron solubility did not correspond to silicon content or Fe(II) content. However, source emission and ambient samples with high iron solubility corresponded to the sulfur content observed in single particles. A similar correspondence between bulk iron solubility and bulk sulfate content in a series of Atlanta PM2.5 fine particle samples (N = 358) further supported this trend. In addition, results of linear combination fitting experiments show the presence of iron sulfates in several high iron solubility source emission and ambient PM2.5 samples. These results suggest that the sulfate content (related to the presence of iron sulfates and/or acid-processing mechanisms by H(2)SO(4)) of iron-containing particles is an important proxy for iron solubility.

  2. Progress of the research and development on the geological disposal technology of HLW with aid of the industry/university collaboration system and fixed term researcher system

    International Nuclear Information System (INIS)

    Yamada, Fumitaka; Sonobe, Hitoshi; Igarashi, Hiroshi

    2008-02-01

    In Japan Atomic Energy Agency (JAEA), various systems associated with the collaboration with industries and universities on the Nuclear Fuel Cycle and the Postdoctoral Fellow system, etc. are enacted. These systems have been operated considering the needs of JAEA's program, industry and academia, resultantly contributed, for example, to basic research and the project development. The activities under these collaboration systems contain personal exchanges, the publication of the accomplishments and utilization of those, in research and development concerning geological disposal technology of high-level radioactive waste (HLW). These activities have progressed in Power Reactor and Nuclear Fuel Development Corporation (PNC) and Japan Nuclear Cycle Development Institute (JNC), which are the successive predecessors of JAEA, through JAEA. The accomplishments from these systems have been not only published as papers in journals and individual technical reports but also integrated into the project reports, accordingly contributed to the advancement of the national program on the geological disposal of HLW. In this report, the progress of the research and development under these systems was investigated from the beginning of the operation of the systems. The contribution to the research and development on geological disposal technology of HLW was also studied. On the basis of these studies, the future utilization of the systems of the collaboration was also discussed from the view point of the management of research and development program. A CD-ROM is attached as an appendix. (J.P.N.)

  3. Preparation and characterization of an improved borosilicate glass for the solidification of high level radioactive fission product solutions (HLW). Pt. 2

    International Nuclear Information System (INIS)

    Kahl, L.; Ruiz-Lopez, M.C.; Saidl, J.; Dippel, T.

    1982-04-01

    In the 'Institut fuer Nuklare Entsorgungstechnik' the borosilicate glass VG 98/12 has been developed for the solidification of the high level radioactive waste (HLW). This borosilicate glass can be used in a direct heated ceramic melter and forms together with the HLW the borosilicate glass product GP 98/12. This borosilicate glass product has been examined in detail both in liquid and solid state. The elements contained in the HLW can be incorporated without problems. Only in a few exceptions the concentration must be kept below certain limits to exclude the formation of a second phase ('yellow phase') by separation. No spontaneous crystallization and no crystallization over a long time could be observed as long as the temperature of the borosilicate glass product is kept below its transformation area. Simulating accidental conditions in the final storage, samples had been leached at temperatures up to 200 0 C and pressures up to 130 bar with saturated rock salt brine and saturated quinary salt brine. The leaching process seems to be stopped by the formed 'leached layer' on the surface of the borosilicate glass product after a limited leaching time. Detailed investigations have been started to explain this phenomenon. (orig.) [de

  4. Geochemical evaluation of the near-field for future HLW repository at Olkiluoto

    International Nuclear Information System (INIS)

    Idiart, A.; Maia, F.; Arcos, D.

    2013-10-01

    The concept for the final disposal of spent nuclear fuel in Finland considers an engineered and natural (crystalline rock) multi-barrier system surrounding the spent fuel. This work aims at predicting and making a quantitative assessment of the geochemical evolution of the near-field (canister, buffer, backfill and adjacent fractured bedrock) during the unsaturated thermal period and in the long-term, after saturation has been completed. The groundwater/bentonite buffer interaction during the unsaturated thermal period is tackled through a two-dimensional (2D) axisymmetric scheme using the thermo-hydro-geochemical code TOUGHREACT. In turn, the long-term interaction of the fully water-saturated buffer and backfill with groundwater is assessed through 3D numerical models using the reactive transport code PHAST under isothermal conditions. A set of base cases have been set up based on the most plausible set of input data. In addition, a limited number of sensitivity cases have been conducted to analyse the influence of key parameters controlling the system and reduce uncertainty. Predicted mineralogical changes of accessory minerals in the bentonite for the thermal period are controlled by the dependence of mineral solubilities on temperature and on the solute transport by advection during the saturation process, and diffusion during the whole period. The results of the thermal period indicate that a small amount of the primary amorphous silica is redistributed in the buffer: dissolution close to the canister and precipitation close to the buffer - rock interface. Primary calcite dissolution/precipitation is minimal, remaining stable throughout the simulation time in all cases. Anhydrite precipitates near the canister due to the elevated temperature, while it dissolves from the outside of the buffer. The results indicate that there is no significant evaporation of water near the copper canister and thus no chloride salt reaches saturation. The geochemical changes of

  5. Confidence building on the total system performance assessment code, MASCOT-K for permanent disposal of HLW in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Y. S.; Kim, S. G.; Kang, C. H

    2002-12-01

    To perform Total System Performance Assessment(TSPA) of a potential HLW repository, it is necessary to develop the TSPA code. KAERI has developed the one-dimensional PSA code MASCOT-K since 1997 and verified special modules dedicated for the dissolution of spent nuclear fuel. In the second R and D phase, MASCOT-K is once again verified as a part of the confidence building for TSPA. The AMBER code based on the totally different mathematical approach, compartment theory is used together with MASCOT-K to assess the annual individual doses for given K- and Q- scenarios. Results indicate that both AMBER and MASCOT-K simulate the annual individual doses to a potential biosphere. And the MASCOT-K is more flexible to describe the natural barrier such as a fracture for sensitivity studies. In the third R and D phase, MASCOT-K will be actively used to check whether the proposed KAERI reference disposal concept is solid or not.

  6. Confidence building on the total system performance assessment code, MASCOT-K for permanent disposal of HLW in Korea

    International Nuclear Information System (INIS)

    Hwang, Y. S.; Kim, S. G.; Kang, C. H.

    2002-12-01

    To perform Total System Performance Assessment(TSPA) of a potential HLW repository, it is necessary to develop the TSPA code. KAERI has developed the one-dimensional PSA code MASCOT-K since 1997 and verified special modules dedicated for the dissolution of spent nuclear fuel. In the second R and D phase, MASCOT-K is once again verified as a part of the confidence building for TSPA. The AMBER code based on the totally different mathematical approach, compartment theory is used together with MASCOT-K to assess the annual individual doses for given K- and Q- scenarios. Results indicate that both AMBER and MASCOT-K simulate the annual individual doses to a potential biosphere. And the MASCOT-K is more flexible to describe the natural barrier such as a fracture for sensitivity studies. In the third R and D phase, MASCOT-K will be actively used to check whether the proposed KAERI reference disposal concept is solid or not

  7. A compartment model for nuclide release calculation in the near-and far-field of a HLW repository

    International Nuclear Information System (INIS)

    Lee, Youn Myoung; Hwang, Yong Soo; Kang, Chul Hyung; Hahn, Pil Soo

    2004-01-01

    The HLW-relevant R and D program for disposal of high-level radioactive waste has been carried out at Korea Atomic Energy Research Institute (KAERI) since early 1997, from which a conceptual Korea Reference Repository System for direct disposal of nuclear spent fuel is to be introduced by the end of 2007. A preliminary reference geologic repository concept considering such established criteria and requirements as spent fuel and generic site characteristics in Korea was roughly envisaged in 2003. Not only to demonstrate how much a reference repository is safe in the generic point of view with several possible scenarios and cases associated with a preliminary repository concept by conducting calculations for nuclide release and transport in the near - and far - field components of the repository, even though sufficient information has not been available that much yet, but also to show a appropriate methodology by which both a generic and site - specific safety assessment could be performed for further in - depth development of Korea reference repository concept, nuclide release calculation study for various nuclide release cases is mandatory. To this end a similar study done and yet limited for the near - field release case has been extended to the case including far - field system by introducing some more geosphere compartments. Advective and longitudinal dispersive nuclide transports along the fracture with matrix diffusion as well as several retention mechanisms and nuclide ingrowth has been added

  8. Mechanical behavior of host rock close to H.L.W. disposal cavities in a deep granitic formation

    International Nuclear Information System (INIS)

    Hoorelbeke, J.M.; Dourthe, M.

    1986-01-01

    The construction of a H.L.W. repository in a deep granitic formation creates mechanical disturbances in the rock on the scale of the massif and in the nearfield. Amongst all the disturbances noted in the nearfield, this study is concerned with examining the evolution of stresses linked with the excavation of the rock and the rise in temperature in the proximity of the waste packages. Several linear elasticity calculations were made using on the one hand finite element models and on the other simple analytical models. These calculations concern two different storage concepts - in room concept and in floor concept- whose differences in mechanical behavior are analyzed. A study of sensitivity with regard to the characteristics of the rock and to the initial geostatic stresses is presented. The comparison of the calculated stresses with three-dimensional failure criteria gives a clear indication of the satisfactory behavior of granite for final storage. However, the need for experimental study and complementary calculation must be emphasized

  9. On nitrogen solubility in water

    International Nuclear Information System (INIS)

    Kalajda, Yu.A.; Katkov, Yu.D.; Kuznetsov, V.A.; Lastovtsev, A.Yu.; Lastochkin, A.P.; Susoev, V.S.

    1980-01-01

    Presented are the results of experimental investigations on nitrogen solubility in water under 0-15 MPa pressure, at the temperature of 100-340 deg C and nitrogen concentration of 0-5000 n.ml. N 2 /kg H 2 O. Empiric equations are derived and a diagram of nitrogen solubility in water is developed on the basis of the experimental data, as well as critically evaluated published data. The investigation results can be used in analyzing water-gas regime of a primary heat carrier in stream-generating plants with water-water reactors

  10. Preliminary considerations concerning actinide solubilities

    International Nuclear Information System (INIS)

    Newton, T.W.; Bayhurst, B.P.; Daniels, W.R.; Erdal, B.R.; Ogard, A.E.

    1980-01-01

    Work at the Los Alamos Scientific Laboratory on the fundamental solution chemistry of the actinides has thus far been confined to preliminary considerations of the problems involved in developing an understanding of the precipitation and dissolution behavior of actinide compounds under environmental conditions. Attempts have been made to calculate solubility as a function of Eh and pH using the appropriate thermodynamic data; results have been presented in terms of contour maps showing lines of constant solubility as a function of Eh and pH. Possible methods of control of the redox potential of rock-groundwater systems by the use of Eh buffers (redox couples) is presented

  11. Thorium oxalate solubility and morphology

    International Nuclear Information System (INIS)

    Monson, P.R. Jr.; Hall, R.

    1981-10-01

    Thorium was used as a stand-in for studying the solubility and precipitation of neptunium and plutonium oxalates. Thorium oxalate solubility was determined over a range of 0.001 to 10.0 in the concentration parameter [H 2 C 2 O 4 ]/[HNO 3 ] 2 . Morphology of thorium oxide made from the oxalate precipitates was characterized by scanning electron microscopy. The different morphologies found for oxalate-lean and oxalate-rich precipitations were in agreement with predictions based on precipitation theory

  12. The long term geochemical evolution of the nearfield of the HLW repository

    Energy Technology Data Exchange (ETDEWEB)

    Bradbury, M. H.; Berner, U.; Curti, E.; Hummel, W.; Kosakowski, G.; Thoenen, T.

    2014-11-15

    The work presented in this report focuses on the spatial and temporal evolution of the near-field of the high level radioactive waste repository situated in the Opalinus Clay formation. The major components of the near-field of such a repository are spent fuel, vitrified high-level waste, canisters (assumed for the purposes of the present report to be made of carbon steel), compacted bentonite and a concrete liner. Over the one million year time period considered in safety analysis, these components will chemically interact with one another and potentially change their retention characteristics. As a starting reference point the 'initial' (unreacted) states of the Opalinus Clay, bentonite, concrete liner (mineralogies and water chemistries) and the canister are briefly described. The main processes considered to influence the evolution of the repository in time and space, and which often operate over different time scales, are: interactions of the concrete tunnel liner with compacted bentonite and Opalinus Clay, temperature gradients caused by the heat generating high level waste, mineralogical changes to the compacted bentonite through interactions with the corrosion products of the iron canisters, and finally, the dissolution of the spent fuel and vitrified high-level waste. The consequences of these processes (as a function of time) on the long term barrier performance of the near-field have been estimated, particularly with respect to radionuclide solubilities and the sorption, diffusion and swelling characteristics of the bentonite. The main conclusions drawn are as follows: The alteration depth into the bentonite due to the interaction with the concrete liner (assumed to be 15 cm thick) is likely to be much less than 13 cm over a one million year time scale, with the main reaction products being clays (illite), hydroxides, carbonates, calcium silicate hydrates, and aluminosilicates. The swelling pressure and the sorption capacity of the bentonite in

  13. The long term geochemical evolution of the nearfield of the HLW repository

    International Nuclear Information System (INIS)

    Bradbury, M. H.; Berner, U.; Curti, E.; Hummel, W.; Kosakowski, G.; Thoenen, T.

    2014-11-01

    The work presented in this report focuses on the spatial and temporal evolution of the near-field of the high level radioactive waste repository situated in the Opalinus Clay formation. The major components of the near-field of such a repository are spent fuel, vitrified high-level waste, canisters (assumed for the purposes of the present report to be made of carbon steel), compacted bentonite and a concrete liner. Over the one million year time period considered in safety analysis, these components will chemically interact with one another and potentially change their retention characteristics. As a starting reference point the 'initial' (unreacted) states of the Opalinus Clay, bentonite, concrete liner (mineralogies and water chemistries) and the canister are briefly described. The main processes considered to influence the evolution of the repository in time and space, and which often operate over different time scales, are: interactions of the concrete tunnel liner with compacted bentonite and Opalinus Clay, temperature gradients caused by the heat generating high level waste, mineralogical changes to the compacted bentonite through interactions with the corrosion products of the iron canisters, and finally, the dissolution of the spent fuel and vitrified high-level waste. The consequences of these processes (as a function of time) on the long term barrier performance of the near-field have been estimated, particularly with respect to radionuclide solubilities and the sorption, diffusion and swelling characteristics of the bentonite. The main conclusions drawn are as follows: The alteration depth into the bentonite due to the interaction with the concrete liner (assumed to be 15 cm thick) is likely to be much less than 13 cm over a one million year time scale, with the main reaction products being clays (illite), hydroxides, carbonates, calcium silicate hydrates, and aluminosilicates. The swelling pressure and the sorption capacity of the bentonite in this

  14. TESTS WITH HIGH-BISMUTH HLW GLASSES FINAL REPORT VSL-10R1780-1, Rev. 0; 12/13/10

    International Nuclear Information System (INIS)

    Matlack, K.S.; Kruger, A.A.; Joseph, I.; Gan, H.; Kot, W.K.; Chaudhuri, M.; Mohr, R.K.; Mckeown, D.A.; Bardakei, T.; Gong, W.; Buecchele, A.C.; Pegg, I.L.

    2011-01-01

    This Final Report describes the testing of glass formulations developed for Hanford High Level Waste (HLW) containing high concentrations of bismuth. In previous work on high-bismuth HLW streams specified by the Office of River Protection (ORP), fully compliant, high waste loading compositions were developed and subjected to melter testing on the DM100 vitrification system. However, during heat treatment according to the Hanford Tank Waste Treatment and Immobilization Plant (WTP) HLW canister centerline cooling (CCC) curves, crucible melts of the high-bismuth glasses were observed to foam. Clearly, such an occurrence during cooling of actual HLW canisters would be highly undesirable. Accordingly, the present work involves larger-scale testing to determine whether this effect occurs under more prototypical conditions, as well as crucible-scale tests to determine the causes and potentially remediate the observed foaming behavior. The work included preparation and characterization of crucible melts designed to determine the underlying causes of the foaming behavior as well as to assess potential mitigation strategies. Testing was also conducted on the DM1200 HLW Pilot melter with a composition previously tested on the DM100 and shown to foam during crucible-scale CCC heat treatment. The DM1200 tests evaluated foaming of glasses over a range of bismuth concentrations poured into temperature-controlled, 55-gallon drums which have a diameter that is close to that of the full-scale WTP HLW canisters. In addition, the DM1200 tests provided the first large-scale melter test data on high-bismuth WTP HLW compositions, including information on processing rates, cold cap behavior and off-gas characteristics, and data from this waste composition on the prototypical DM1200 off-gas treatment system. This work builds on previous work performed at the Vitreous State Laboratory (VSL) for ORP on the same waste composition. The scope of this study was outlined in a Test Plan that was

  15. TESTS WITH HIGH-BISMUTH HLW GLASSES FINAL REPORT VSL-10R1780-1 REV 0 12/13/10

    Energy Technology Data Exchange (ETDEWEB)

    MATLACK KS; KRUGER AA; JOSEPH I; GAN H; KOT WK; CHAUDHURI M; MOHR RK; MCKEOWN DA; BARDAKEI T; GONG W; BUECCHELE AC; PEGG IL

    2011-01-05

    This Final Report describes the testing of glass formulations developed for Hanford High Level Waste (HLW) containing high concentrations of bismuth. In previous work on high-bismuth HLW streams specified by the Office of River Protection (ORP), fully compliant, high waste loading compositions were developed and subjected to melter testing on the DM100 vitrification system. However, during heat treatment according to the Hanford Tank Waste Treatment and Immobilization Plant (WTP) HLW canister centerline cooling (CCC) curves, crucible melts of the high-bismuth glasses were observed to foam. Clearly, such an occurrence during cooling of actual HLW canisters would be highly undesirable. Accordingly, the present work involves larger-scale testing to determine whether this effect occurs under more prototypical conditions, as well as crucible-scale tests to determine the causes and potentially remediate the observed foaming behavior. The work included preparation and characterization of crucible melts designed to determine the underlying causes of the foaming behavior as well as to assess potential mitigation strategies. Testing was also conducted on the DM1200 HLW Pilot melter with a composition previously tested on the DM100 and shown to foam during crucible-scale CCC heat treatment. The DM1200 tests evaluated foaming of glasses over a range of bismuth concentrations poured into temperature-controlled, 55-gallon drums which have a diameter that is close to that of the full-scale WTP HLW canisters. In addition, the DM1200 tests provided the first large-scale melter test data on high-bismuth WTP HLW compositions, including information on processing rates, cold cap behavior and off-gas characteristics, and data from this waste composition on the prototypical DM1200 off-gas treatment system. This work builds on previous work performed at the Vitreous State Laboratory (VSL) for ORP on the same waste composition. The scope of this study was outlined in a Test Plan that was

  16. Molecular Simulation of Gas Solubility in Nitrile Butadiene Rubber.

    Science.gov (United States)

    Khawaja, M; Sutton, A P; Mostofi, A A

    2017-01-12

    Molecular simulation is used to compute the solubility of small gases in nitrile butadiene rubber (NBR) with a Widom particle-insertion technique biased by local free volume. The convergence of the method is examined as a function of the number of snapshots upon which the insertions are performed and the number of insertions per snapshot and is compared to the convergence of the unbiased Widom insertion technique. The effect of varying the definition of local free volume is also investigated. The acrylonitrile content of the polymer is altered to examine its influence on the solubility of helium, CO 2 , and H 2 O, and the solubilities of polar gases are found to be enhanced relative to those of nonpolar gases, in qualitative agreement with experiment. To probe this phenomenon further, the solubilities are decomposed into contributions from the neighborhoods of different atoms, using a Voronoi cell construction, and a strong bias is found for CO 2 and H 2 O in particular to be situated near nitrogen sites in the elastomer. Temperature is shown to suppress the solubility of CO 2 and H 2 O but to increase that of helium. Increasing pressure is found to suppress the solubility of all gases but at different rates, according to a balance between their molecular sizes and electrostatic interactions with the polymer. These results are relevant to the use of NBR seals at elevated temperatures and pressures, such as in oil and gas wells.

  17. Research on solubility characteristics of gaseous methyl iodide

    International Nuclear Information System (INIS)

    Zhou Yanmin; Sun Zhongning; Gu Haifeng; Wang Junlong

    2014-01-01

    With the deionized water as the absorbent, the solubility characteristics of the gaseous methyl iodide were studied under different temperature and pressure conditions, using a dynamic measuring method. The results show that within the range of experiment parameters, namely temperature is below 80℃ and pressure is lower than 0.3 MPa, the physical dissolution process of gaseous methyl iodide in water obeys Henry's law. The solubility coefficient under different temperature and pressure conditions was calculated based on the measurement results. Further research indicates that at atmospheric pressure, the solubility coefficient of methyl iodide in water decreases exponentially with the increase of temperature. While the pressure changes from 0.1 MPa to 0.3 MPa with equal interval, the solubility coefficient also increases linearly. The variation of the solubility coefficient with temperature under different pressure conditions all decreases exponentially. An equation is given to calculate the solubility coefficient of methyl iodide under different pressure and temperature conditions. (authors)

  18. The study of long-term stability in liquid-solid phases for HLW disposal

    International Nuclear Information System (INIS)

    Wei, Y.Y.; Tseng, C.L.; Yang, J.Y.; Ke, C.H.; Wang, T.H.; Jan, Y.L.; Lee, C.B.; Lan, P.L.; Hsu, C.N.; Tsai, S.C.; Li, M.H.; Teng, S.P.

    2005-01-01

    Full text of publication follows: This study is conducted to observe changes in both chemical properties of buffer materials and liquid phases over an experimental period of 2 years. In our experiments, bentonite powder and crushed granite are separately mixed with synthetic groundwater, synthetic seawater and de-ionised water at a fixed liquid-solid ratio of 30. A mixed set with both bentonite and granite together as solid phase is also investigated. During this study, aliquots of the liquid phases are sampled every two months and pH and Eh values are measured immediately. Concentrations of Na, Mg, K, Al, Ca, Ti, Mn, Ba, Fe, Sr, Li and Th are analyzed in the liquid phase directly by ICP-AES. After separation by centrifugation followed by freeze drying and digestion, the solid phases are analyzed as well for elemental composition. Alteration of solid phases during the experimental period is discussed. The preliminary results show that the pH values of the three solutions vary considerably in the individual experimental systems containing bentonite, granite or the mixed system. In general, higher pH values are found in DI-water for all solid phases. Eh values fluctuate a lot in the range 100 to 300 mV in all experiment sets. Different to the experiments with granite for which similar Eh values are found in all solutions, a significantly different Eh-value is found in the experiment with bentonite in DI-water as compared to the other solutions. The results from element analysis indicate that equilibrium is achieved after only two months and element concentrations change only slightly thereafter. We conclude from our experiments that both bentonite and granite keep their characteristics as radionuclide sorbents in the vicinity of a nuclear waste repository. Reaction equilibria appear to be attained rapidly. Because there are just a few alterations in this study, it would be a huge error source in analyzing from the inhomogeneous solid phase such as granite and losses

  19. Solubility limits on radionuclide dissolution

    Energy Technology Data Exchange (ETDEWEB)

    Kerrisk, J.F.

    1984-12-31

    This paper examines the effects of solubility in limiting dissolution rates of a number of important radionuclides from spent fuel and high-level waste. Two simple dissolution models were used for calculations that would be characteristics of a Yucca Mountain repository. A saturation-limited dissolution model, in which the water flowing through the repository is assumed to be saturated with each waste element, is very conservative in that it overestimates dissolution rates. A diffusion-limited dissolution model, in which element-dissolution rates are limited by diffusion of waste elements into water flowing past the waste, is more realistic, but it is subject to some uncertainty at this time. Dissolution rates of some elements (Pu, Am, Sn, Th, Zr, Sm) are always limited by solubility. Dissolution rates of other elements (Cs, Tc, Np, Sr, C, I) are never solubility limited; their release would be limited by dissolution of the bulk waste form. Still other elements (U, Cm, Ni, Ra) show solubility-limited dissolution under some conditions. 9 references, 3 tables.

  20. Solubility of Nd in brine

    International Nuclear Information System (INIS)

    Khalili, F.I.; Symeopoulos, V.; Chen, J.F.; Choppin, G.R.

    1994-01-01

    The solubility of Nd(III) has been measured at 23±3 C in a synthetic brine at pcH 6.4, 8.4, 10.4 and 12.4. The brine consisted predominantly of (Na+K)Cl and MgCl 2 with an ionic strength of 7.8 M (9.4 m) a solid compound of Nd(III) at each pcH was assigned from X-ray diffraction patterns. The log values of the experimental solubilities decrease fomr -3 at pcH 6.4 to -5.8 at pcH 8.4; at pcH 10.4 and 12.4 the solubility was below the detection limit of -7.5. The experimental solubility does not follow closely the variation with pcH estimated from modeling of the species in solution in equilibrium with the Nd solid using S.I.T. (orig.)

  1. SIERRA Mechanics, an emerging massively parallel HPC capability, for use in coupled THMC analyses of HLW repositories in clay/shale

    International Nuclear Information System (INIS)

    Bean, J.E.; Sanchez, M.; Arguello, J.G.

    2012-01-01

    Document available in extended abstract form only. Because, until recently, U.S. efforts had been focused on the volcanic tuff site at Yucca Mountain, radioactive waste disposal in U.S. clay/shale formations has not been considered for many years. However, advances in multi-physics computational modeling and research into clay mineralogy continue to improve the scientific basis for assessing nuclear waste repository performance in such formations. Disposal of high-level radioactive waste (HLW) in suitable clay/shale formations is attractive because the material is essentially impermeable and self-sealing, conditions are chemically reducing, and sorption tends to prevent radionuclide transport. Vertically and laterally extensive shale and clay formations exist in multiple locations in the contiguous 48 states. This paper describes an emerging massively parallel (MP) high performance computing (HPC) capability - SIERRA Mechanics - that is applicable to the simulation of coupled-physics processes occurring within a potential clay/shale repository for disposal of HLW within the U.S. The SIERRA Mechanics code development project has been underway at Sandia National Laboratories for approximately the past decade under the auspices of the U.S. Department of Energy's Advanced Scientific Computing (ASC) program. SIERRA Mechanics was designed and developed from its inception to run on the latest and most sophisticated massively parallel computing hardware, with the capability to span the hardware range from single workstations to systems with thousands of processors. The foundation of SIERRA Mechanics is the SIERRA tool-kit, which provides finite element application-code services such as: (1) mesh and field data management, both parallel and distributed; (2) transfer operators for mapping field variables from one mechanics application to another; (3) a solution controller for code coupling; and (4) included third party libraries (e.g., solver libraries, communications

  2. Influence of substitutional atoms on the solubility limit of carbon in bcc iron

    International Nuclear Information System (INIS)

    Saitoh, Hajime; Ushioda, Kohsaku; Yoshinaga, Naoki; Yamada, Wataru

    2011-01-01

    The influence of substitutional atoms (Mn, Cr, Si, P, and Al) on the solubility limit of C in body-centered cubic iron in equilibrium with cementite was investigated in low-carbon steels at a temperature of 700 o C. The C solubility limit was determined from internal friction measurements combined with infrared analysis of C using a high-frequency combustion technique. Experiments clarified that Mn, Cr and Al hardly change the C solubility limit, whereas P and Si increase it.

  3. Hanford Supplemental Treatment: Literature and Modeling Review of SRS HLW Salt Dissolution and Fractional Crystallization

    Energy Technology Data Exchange (ETDEWEB)

    Choi, A. S.; Flach, G. P.; Martino, C. J.; Zamecnik, J. R.; Harris, M. K.; Wilmarth, W. R.; Calloway, T. B.

    2005-03-23

    In order to accelerate waste treatment and disposal of Hanford tank waste by 2028, the Department of Energy (DOE) and CH2M Hill Hanford Group (CHG), Inc. are evaluating alternative technologies which will be used in conjunction with the Waste Treatment Plant (WTP) to safely pretreat and immobilize the tank waste. Several technologies (Bulk Vitrification and Steam Reforming) are currently being evaluated for immobilizing the pretreated waste. Since the WTP does not have sufficient capacity to pretreat all the waste going to supplemental treatment by the 2028 milestone, two technologies (Selective Dissolution and Fractional Crystallization) are being considered for pretreatment of salt waste. The scope of this task was to: (1) evaluate the recent Savannah River Site (SRS) Tank 41 dissolution campaign and other literature to provide a more complete understanding of selective dissolution, (2) provide an update on the progress of salt dissolution and modeling activities at SRS, (3) investigate SRS experience and outside literature sources on industrial equipment and experimental results of previous fractional crystallization processes, and (4) evaluate recent Hanford AP104 boildown experiments and modeling results and recommend enhancements to the Environmental Simulation Program (ESP) to improve its predictive capabilities. This report provides a summary of this work and suggested recommendations.

  4. Solubility of simulated PWR primary circuit corrosion products

    International Nuclear Information System (INIS)

    Kunig, R.H.; Sandler, Y.L.

    1986-08-01

    The solubility behavior of non-stoichiometric nickel ferrites, nickel-cobalt ferrites, and magnetite, as model substances for the corrosion products (''crud'') formed in nuclear pressurized water reactors, was studied in a flow system in aqueous solutions of lithium hydroxide, boric acid, and hydrogen with pH, temperature, and hydrogen concentrations as parameters. Below the temperature region of 300 to 330 0 C, at hydrogen concentrations of 25 to 40 cm 3 /kg H 2 O as used during reactor operation, the solubility of nickel-cobalt ferrite is the same as that of Ni and Co/sub x/Fe/sub 3-x/O 4 (x 3 /kg of hydrogen, the equilibrium iron and nickel solubilities increase congruently down to about 100 0 C, in a manner consistent with the solubility of Fe 3 O 4 , but sharply decline at lower temperatures, apparently due to formation of a borated layer. A cooldown experiment on a time scale of a typical Westinghouse reactor shutdown, as well as static experiments carried out on various ferrite samples at 60 0 C show that after addition of oxygen or peroxide evolution of nickel (and possibly cobalt) above the equilibrium solubility in hydrogen depends on the presence of dissociation products prior to oxidation. Thermodynamic calculations of various reduction and oxidative decomposition reactions for stoichiometric and non-stoichiometric nickel ferrite and cobalt ferrite are presented. Their significance to evolutions of nickel and cobalt on reactor shutdown is discussed. 30 refs., 38 figs., 34 tabs

  5. 10 years of transport of vitrified High Level Waste (HLW) from COGEMA La Hague

    International Nuclear Information System (INIS)

    Lancelot, J.; Martinotti, B.; Tourneux, F.

    2004-01-01

    COGEMA has been using, for decades, its large experience of Reprocessing in both Gas Cooled reactors (GCR) and LWR fuels with the following facilities: Marcoule UP1 plant started up in late 50's: La Hague UP2 plant started up in 1966 first with GCR fuels and from 1976 with LWR Fuel (including foreign fuels): La Hague UP3 plant started up in 1990 Foreign Utilities signed Reprocessing Contracts with COGEMA from 1970's, providing returns of residues to the country of origin where they will be managed in a safe storage facility. Therefore, for nearly 30 years Spent Fuel coming from Japan, Germany, Belgium, Switzerland and the Netherlands are processed on La Hague site

  6. JSS project phase 4: Experimental and modelling studies of HLW glass dissolution in repository environments

    International Nuclear Information System (INIS)

    1987-10-01

    A goal of the JSS project was to develop a scientific basis for understanding the effects of waste package components, groundwater chemistry, and other repository conditions on glass dissolution behaviour, and to develop and refine a model for the processes governing glass dissolution. The fourth phase of the project, which was performed by the Hahn-Meitner-Institut, Berlin, FRG, dealt specifically with model development and application. Phase 4 also adressed whether basaltic glasses could serve as natural analogues for nuclear waste glasses, thus providing a means to test the capability of the model for long-term predictions. Additional experiments were performed in order to complete the data base necessary to model interactions between the glass and bentonite and between glass and steel corrosion products. More data on temperature, S/V, and pH dependence of the glass/water reaction were also collected. In this report, the data acquired during phase 4 are presented and discussed. (orig./DG)

  7. Review of the effective approaches for providing the R and D information on the geological disposal of HLW

    International Nuclear Information System (INIS)

    Mitsuhashi, Hiroshi; Okuhara, Hidehiko; Nanjo, Yuki

    2001-03-01

    Japan Nuclear Cycle Development Institute (JNC) had already carried out Research and development (R and D) activities for the Geological Disposal of High-level Radioactive Waste (HLW) in Japan, the information activities in order to gain a public understanding in Japan. At present, however, the information on the geological disposal project including R and D is still unpopular among the public and does not draw so much attention compared to the other current topics. To make a national consensus on the project, the effective public relational activities with the suitable approaches for the various groups/classes among the public should be done. From the viewpoint of gaining the social recognition, having the valuable interviews with the authorities, opinion leaders and other specialists, we reviewed the approaches of the effective information activities to gain the public attention and let them have proper understanding. We also had some group interviews subject to the university students and housewives, who are expected to have no concern with the geological disposal. During these interviews, we had monitored the degree of understanding on the geological disposal and JNC's R and D activities utilizing the conventional materials that JNC had already prepared, such as brochures and video tape recording, and found if the materials were helpful or not, for proper understanding. A questionnaire survey on the internet was done, as one of yardsticks for the effect of the JNC's activities. We studied the degree of understanding of the respondents, and analyzed the effect of the JNC's public relational activities. Based on the another questionnaire survey results at 'Forum on geological disposal', which was held by JNC, we also analyzed the effect of the forum as one of two-way communications tools. Following the above analysis, the effective approaches of the future public relational activities of the Geological disposal was reviewed. (author)

  8. A comparative Study between GoldSim and AMBER Based Biosphere Assessment Models for an HLW Repository

    International Nuclear Information System (INIS)

    Lee, Youn-Myoung; Hwang, Yong-Soo

    2007-01-01

    To demonstrate the performance of a repository, the dose exposure rate to human being due to long-term nuclide releases from a high-level waste repository (HLW) should be evaluated and the results compared to the dose limit presented by the regulatory bodies. To evaluate such a dose rate to an individual, biosphere assessment models have been developed and implemented for a practical calculation with the aid of such commercial tools as AMBER and GoldSim, both of which are capable of probabilistic and deterministic calculation. AMBER is a general purpose compartment modeling tool and GoldSim is another multipurpose simulation tool for dynamically modeling complex systems, supporting a higher graphical user interface than AMBER and a postprocessing feature. And also unlike AMBER, any kind of compartment scheme can be rather simply constructed with an appropriate transition rate between compartments, GoldSim is designed to facilitate the object-oriented modules to address any specialized programs, similar to solving jig saw puzzles. During the last couple of years a compartment modeling approach for a biosphere has been mainly carried out with AMBER in KAERI in order to conservatively or rather roughly provide dose conversion factors to get the final exposure rate due to a nuclide flux into biosphere over various geosphere-biosphere interfaces (GBIs) calculated through nuclide transport modules. This caused a necessity for a newly devised biosphere model that could be coupled to a nuclide transport model with less conservatism in the frame of the development of a total system performance assessment modeling tool, which could be successfully done with the aid of GoldSim. Therefore, through the current study, some comparison results of the AMBER and the GoldSim approaches for the same case of a biosphere modeling without any consideration of geosphere transport are introduced by extending a previous study

  9. Solubility of sparingly soluble drug derivatives of anthranilic acid.

    Science.gov (United States)

    Domańska, Urszula; Pobudkowska, Aneta; Pelczarska, Aleksandra

    2011-03-24

    This work is a continuation of our systematic study of the solubility of pharmaceuticals (Pharms). All substances here are derivatives of anthranilic acid, and have an anti-inflammatory direction of action (niflumic acid, flufenamic acid, and diclofenac sodium). The basic thermal properties of pure Pharms, i.e., melting and glass-transition temperatures as well as the enthalpy of melting, have been measured with the differential scanning microcalorimetry technique (DSC). Molar volumes have been calculated with the Barton group contribution method. The equilibrium mole fraction solubilities of three pharmaceuticals were measured in a range of temperatures from 285 to 355 K in three important solvents for Pharm investigations: water, ethanol, and 1-octanol using a dynamic method and spectroscopic UV-vis method. The experimental solubility data have been correlated by means of the commonly known G(E) equation: the NRTL, with the assumption that the systems studied here have revealed simple eutectic mixtures. pK(a) precise measurement values have been investigated with the Bates-Schwarzenbach spectrophotometric method. © 2011 American Chemical Society

  10. Thermo-hydro-mechanical processes in the nearfield around a HLW repository in argillaceous formations. Vol. II. In-situ-investigations and interpretative modelling. May 2007 to May 2013

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Chun-Liang; Czaikowski, Oliver; Komischke, Michael; Wieczorek, Klaus

    2014-06-15

    Deep disposal of heat-emitting high-level radioactive waste (HLW) in clay formations will inevitably induce thermo-hydro-mechanical-chemical disturbances to the host rock and engineered barriers over very long periods of time. The responses and resulting property changes of the natural and engineered barriers are to be well understood, characterized, and predicted for assessing the long-term performance and safety of the repositories. In accordance with the R and D programme defined by the German Federal Ministry of Economics and Technology (BMWi), GRS has intensively performed site-independent research work on argillaceous rocks during the last decade. Most of the investigations have been carried out on the Callovo-Oxfordian argillite and the Opalinus clay by par-ticipation in international research projects conducted at the underground research laboratories at Bure in France (MHM-URL) and Mont-Terri in Switzerland (MT-URL). The THM-TON project, which was funded by BMWi under contract number 02E10377, in-vestigated the THM behaviours of the clay host rock and clay-based backfill/sealing materials with laboratory tests, in situ experiments and numerical modelling.

  11. New insight for social risk communication of nuclear power towards social consensus for HLW disposal

    International Nuclear Information System (INIS)

    Kugo, Akihide; Yoshikawa, Hidekazu; Shimoda, Hiroshi; Uda, Akinobu; Wakabayashi, Yasunaga

    2004-01-01

    For the construction of effective knowledge base on safety and non-anxiety for nuclear power, a study on new communication system about social risk information has been initiated by noticing the rapid expansion of Internet in the society. By constructing Internet Website communication system on the geological disposal of high-level radioactive wastes, we conducted the experiment of communication for verifying the principles such as that the basic technical knowledge and trust, and social ethics are indispensable in this process to close the perception gap between nuclear specialists and the general public. The cognition structural equation model by means of the variables reduction method of multiple regression analysis and by compiling the significant paths by covariance structure analysis was built based on this experimental data. Moreover, by investigating more detailed public subconscious on the high-level radioactive wastes by 'text mining method' with the special reference to the Public Comment in July 2000 and the literature survey, it was found that the freely discussing ideas based on the environmental ethics such as 'fairness in results' and 'fairness in opportunity' from scratch would gain a potential of enhancing the social receptivity. (author)

  12. Geochemical site-selection criteria for HLW repositories in Europe and North America

    International Nuclear Information System (INIS)

    Savage, David; Arthur, Randolph C.; Sasamoto, Hiroshi; Shibata, Masahiro; Yui, Mikazu

    2000-01-01

    Geochemical as well as socio-economic issues associated with the selection of potential sites to host a high-level nuclear waste repository have received considerable attention in repository programs in Europe (Belgium, Finland, France, Germany, Spain, Sweden, Switzerland and the U.K.) and North America (Canada and the United States). The objective of the present study is to summarize this international experience with particular emphasis on geochemical properties that factor into the adopted site-selection strategies. Results indicate that the geochemical properties of a site play a subordinate role, at best, to other geotechnical properties in the international site-selection approaches. In countries where geochemical properties are acknowledged in the site-selection approach, requirements are stated qualitatively and tend to focus on associated impacts on the stability of the engineered barrier system and on radionuclide transport. Site geochemical properties that are likely to control the long-term stability of geochemical conditions and radionuclide migration behavior are unspecified, however. This non-prescriptive approach may be reasonable for purposes of screening among potential sites, but a better understanding of site properties that are most important in controlling the long-term geochemical evolution of the site over a range of possible scenarios would enable the potential sites to be ranked in terms of their suitability to host a repository. (author)

  13. Study on risk communication by using web system for the social consensus toward HLW final disposal

    International Nuclear Information System (INIS)

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

    2008-01-01

    The web site that has illustrated characters to navigate information pertaining to unfamiliar issue such as high-level radioactive waste geological disposal is an effective method. However, since the information was provided mainly from a pro-nuclear power generation group, it resulted in frustration for the web site user because viewpoints outside the group were not considered nor the explanations were based on only rational aspects, the persuasive explanation based on technical viewpoints in other words. To close this communication gap, this research aims to enhance a better sense of involvement and social collaboration by creating an interactive communication model promoting emotional acceptance and independent thinking with Web system. This purpose was accomplished by the dialog-mode explanation and the scenarios with norm activation theory supported by facial expressions of the illustrated navigators to stimulate the emotional involvement of viewers and the specialists' reliable response on the electrical bulletin board system, then we conducted preparatory experiments concerning its effects and assessed its affectiveness by making this model available over the Internet. (author)

  14. Near-field solubility studies

    International Nuclear Information System (INIS)

    Thomason, H.P.; Williams, S.J.

    1992-02-01

    Experimental determinations of the solubilities of americium, plutonium, neptunium, protactinium, thorium, radium, lead, tin, palladium and zirconium are reported. These elements have radioactive isotopes of concern in assessments of radioactive waste disposal. All measurements were made under the highly alkaline conditions typical of the near field of a radioactive waste repository which uses cementitious materials for many of the immobilisation matrices, the backfill and the engineered structures. Low redox potentials, typical of those resulting from the corrosion of iron and steel, were simulated for those elements having more than one accessible oxidation state. The dissolved concentrations of the elements were defined using ultrafiltration. In addition, the corrosion of iron and stainless steel was shown to generate low redox potentials in solution and the solubility of iron(II) at high pH was measured and found to be sufficient for it to act as a redox buffer with respect to neptunium and plutonium. (author)

  15. pH-metric solubility. 3. Dissolution titration template method for solubility determination.

    Science.gov (United States)

    Avdeef, A; Berger, C M

    2001-12-01

    The main objective of this study was to develop an effective potentiometric saturation titration protocol for determining the aqueous intrinsic solubility and the solubility-pH profile of ionizable molecules, with the specific aim of overcoming incomplete dissolution conditions, while attempting to shorten the data collection time. A modern theory of dissolution kinetics (an extension of the Noyes-Whitney approach) was applied to acid-base titration experiments. A thermodynamic method was developed, based on a three-component model, to calculate interfacial, diffusion-layer, and bulk-water reactant concentrations in saturated solutions of ionizable compounds perturbed by additions of acid/base titrant, leading to partial dissolution of the solid material. Ten commercial drugs (cimetidine, diltiazem hydrochloride, enalapril maleate, metoprolol tartrate, nadolol, propoxyphene hydrochloride, quinine hydrochloride, terfenadine, trovafloxacin mesylate, and benzoic acid) were chosen to illustrate the new titration methodology. It was shown that the new method is about 10 times faster in determining equilibrium solubility constants, compared to the traditional saturation shake-flask methods.

  16. DATA SUMMARY REPORT SMALL SCALE MELTER TESTING OF HLW ALGORITHM GLASSES MATRIX1 TESTS VSL-07S1220-1 REV 0 7/25/07

    Energy Technology Data Exchange (ETDEWEB)

    KRUGER AA; MATLACK KS; PEGG IL

    2011-12-29

    Eight tests using different HLW feeds were conducted on the DM100-BL to determine the effect of variations in glass properties and feed composition on processing rates and melter conditions (off-gas characteristics, glass processing, foaming, cold cap, etc.) at constant bubbling rate. In over seven hundred hours of testing, the property extremes of glass viscosity, electrical conductivity, and T{sub 1%}, as well as minimum and maximum concentrations of several major and minor glass components were evaluated using glass compositions that have been tested previously at the crucible scale. Other parameters evaluated with respect to glass processing properties were +/-15% batching errors in the addition of glass forming chemicals (GFCs) to the feed, and variation in the sources of boron and sodium used in the GFCs. Tests evaluating batching errors and GFC source employed variations on the HLW98-86 formulation (a glass composition formulated for HLW C-106/AY-102 waste and processed in several previous melter tests) in order to best isolate the effect of each test variable. These tests are outlined in a Test Plan that was prepared in response to the Test Specification for this work. The present report provides summary level data for all of the tests in the first test matrix (Matrix 1) in the Test Plan. Summary results from the remaining tests, investigating minimum and maximum concentrations of major and minor glass components employing variations on the HLW98-86 formulation and glasses generated by the HLW glass formulation algorithm, will be reported separately after those tests are completed. The test data summarized herein include glass production rates, the type and amount of feed used, a variety of measured melter parameters including temperatures and electrode power, feed sample analysis, measured glass properties, and gaseous emissions rates. More detailed information and analysis from the melter tests with complete emission chemistry, glass durability, and

  17. Solubility of xenon in liquid sodium

    International Nuclear Information System (INIS)

    Veleckis, E.; Cafasso, F.A.; Feder, H.M.

    1976-01-01

    The solubility of xenon in liquid sodium was measured as a function of pressure (2-8 atm) and temperature (350-600 0 C). Henry's law was obeyed with the value of the Henry's law constant, K/sub H/ = N/sub Xe//P, ranging from 1.38 x 10 -10 atm -1 at 350C, to 1.59 x 10 -8 atm -1 at 600 0 C where N/sub Xe/ and P are the atom fraction and the partial pressure of xenon, respectively. The temperature dependence of solubility may be represented by log 10 lambda = (0.663 +- 0.01) - (4500 +- 73) T -1 , where lambda is the Ostwald coefficient (the volume of xenon dissolved per unit volume of sodium at the temperature of the experiment). The heat of solution of xenon in sodium was 20.6 +- 0.7 kcal/mole, where the standard state of xenon is defined as that of 1 mole of an ideal gas, confined to a volume equal to the molar volume of sodium

  18. The Solubility Parameters of Ionic Liquids

    Science.gov (United States)

    Marciniak, Andrzej

    2010-01-01

    The Hildebrand’s solubility parameters have been calculated for 18 ionic liquids from the inverse gas chromatography measurements of the activity coefficients at infinite dilution. Retention data were used for the calculation. The solubility parameters are helpful for the prediction of the solubility in the binary solvent mixtures. From the solubility parameters, the standard enthalpies of vaporization of ionic liquids were estimated. PMID:20559495

  19. The Solubility Parameters of Ionic Liquids

    Directory of Open Access Journals (Sweden)

    Andrzej Marciniak

    2010-04-01

    Full Text Available The Hildebrand’s solubility parameters have been calculated for 18 ionic liquids from the inverse gas chromatography measurements of the activity coefficients at infinite dilution. Retention data were used for the calculation. The solubility parameters are helpful for the prediction of the solubility in the binary solvent mixtures. From the solubility parameters, the standard enthalpies of vaporization of ionic liquids were estimated.

  20. Solubility of Carbon in Nanocrystalline -Iron

    OpenAIRE

    Alexander Kirchner; Bernd Kieback

    2012-01-01

    A thermodynamic model for nanocrystalline interstitial alloys is presented. The equilibrium solid solubility of carbon in -iron is calculated for given grain size. Inside the strained nanograins local variation of the carbon content is predicted. Due to the nonlinear relation between strain and solubility, the averaged solubility in the grain interior increases with decreasing grain size. The majority of the global solubility enhancement is due to grain boundary enrichment however. Therefor...

  1. Solubility of hydrogen isotopes in liquid LiPb

    International Nuclear Information System (INIS)

    Konishi, S.; Yamamoto, Y.; Noborio, K.; Calderoni, P.; Merrill, B.

    2014-01-01

    This research was performed mainly in the first half of the task 1-2 of TITAN project to investigate the interaction between hydrogen isotopes and liquid LiPb. Solubility of hydrogen in liquid LiPb was measured under a static condition. Kyoto University provided the first experimental apparatus shipped to Idaho, and Kyushu University succeeded the experiment and further improved. Obtained solubility generally agreed with some previous reports, but varied orders of magnitudes suggesting influence of impurity or other chemical processes. (author)

  2. Discrete and continuum approaches for the analysis of coupled thermal-mechanical processes in the near field of a HLW repository

    International Nuclear Information System (INIS)

    Shimizu, Hiroyuki; Fujita, Tomoo; Nakama, Shigeo; Koyama, Tomofumi; Chijimatsu, Masakazu

    2011-01-01

    This paper reports on the results of the numerical simulations for the analysis of coupled thermal-mechanical processes in the near field of a HLW repository using Finite Element Method (FEM) and Distinct Element Method (DEM). The FEM approach provides quantitative information of the change of stress during excavation and heating process. On the other hand, the DEM approach shows the crack propagation process at the borehole surface, and this result agrees qualitatively well with the experimental observation. By comparing these results obtained from both approaches, quantitative and qualitative insights into various aspects of the processes occurred in the near field can be obtained. (author)

  3. Development of the Internet Library for the Second Progress Report on R and D for the geological disposal of HLW in Japan

    International Nuclear Information System (INIS)

    Shiotsuki, Masao; Ishikawa, Hirohisa

    2000-01-01

    This paper describes an Internet Library, the goal of which is to improve the quality assurance of the technical content of the Second Progress Report on R and D into the geological disposal of HLW in Japan. The Internet Library is used to centralize information management for the Second Progress Report. It uses a database system which stores a large quantity of technical memoranda and numeric data which provide the technical basis for the report. Members of the public and specialists are allowed access the data held on the system and may communicate their opinions and expert reviews, through the Internet. (author)

  4. Geological boundary conditions for a safety demonstration and verification concept for a HLW repository in claystone in Germany. AnSichT

    Energy Technology Data Exchange (ETDEWEB)

    Stark, Lena; Bebiolka, Anke; Gerardi, Johannes [Federal Institute for Geosciences and Natural Resources (BGR), Hannover (Germany). Dept. of Underground Space for Storage and Economic Use; and others

    2015-07-01

    Within the framework of the R and D project ''AnSichT'', DBE TECHNOLOGY, BGR and GRS are developing a method to demonstrate the safety of a HLW repository in claystone in Germany. The methodological approach basing on a holistic concept, links the legal and geologic boundary conditions, the disposal and closure concept, the demonstration of barrier integrity, and the long-term analysis of the repository evolution as well. The geologic boundary conditions are specified by the description of the geological situation and generic models, the selection of representative parameters and geoscientific long-term predictions. They form a fundament for the system analysis.

  5. Solubility studies of Np(V) in simulated underground water

    International Nuclear Information System (INIS)

    Zhang Yingjie; Ren Lilong; Jiao Haiyang; Yao Jun; Su Xiguang; Fan Xianhua

    2004-01-01

    The solubility of Np(V) in simulated underground water has been measured with the variation of pH, storage time (0-100 days). All experiments were performed in an Ar glove box which contained high purity Ar, with an oxygen content of less than 5ppm. Experimental results show that the solubility of Np(V) in simulated underground water decreased with increasing pH value of solution; the solubility of Np(V) in simulated underground water determined at different pH is : pH=6.96, [Np(V)]=(3.52±0.37) x 10 -4 mol/L; pH=8.04, [Np(V)]=(8.24±0.32) x 10 -5 mol/L; pH=9.01, [Np(V)]=(3.04±0.48) x 10'- 5 mol/L, respectively. (author)

  6. Solubility of Plutonium (IV) Oxalate During Americium/Curium Pretreatment

    International Nuclear Information System (INIS)

    Rudisill, T.S.

    1999-01-01

    Approximately 15,000 L of solution containing isotopes of americium and curium (Am/Cm) will undergo stabilization by vitrification at the Savannah River Site (SRS). Prior to vitrification, an in-tank pretreatment will be used to remove metal impurities from the solution using an oxalate precipitation process. Material balance calculations for this process, based on solubility data in pure nitric acid, predict approximately 80 percent of the plutonium in the solution will be lost to waste. Due to the uncertainty associated with the plutonium losses during processing, solubility experiments were performed to measure the recovery of plutonium during pretreatment and a subsequent precipitation process to prepare a slurry feed for a batch melter. A good estimate of the plutonium content of the glass is required for planning the shipment of the vitrified Am/Cm product to Oak Ridge National Laboratory (ORNL).The plutonium solubility in the oxalate precipitation supernate during pretreatment was 10 mg/mL at 35 degrees C. In two subsequent washes with a 0.25M oxalic acid/0.5M nitric acid solution, the solubility dropped to less than 5 mg/mL. During the precipitation and washing steps, lanthanide fission products in the solution were mostly insoluble. Uranium, and alkali, alkaline earth, and transition metal impurities were soluble as expected. An elemental material balance for plutonium showed that greater than 94 percent of the plutonium was recovered in the dissolved precipitate. The recovery of the lanthanide elements was generally 94 percent or higher except for the more soluble lanthanum. The recovery of soluble metal impurities from the precipitate slurry ranged from 15 to 22 percent. Theoretically, 16 percent of the soluble oxalates should have been present in the dissolved slurry based on the dilution effects and volumes of supernate and wash solutions removed. A trace level material balance showed greater than 97 percent recovery of americium-241 (from the beta dec

  7. On the solubility of plutonium in water

    International Nuclear Information System (INIS)

    Naegele, G.

    1977-12-01

    In a theoretical study, the chemical equilibrium state of saturated Pu solutions in water was determined and the effect of the addition of EDTA on the solubility of Pu estimated. Concentrations of Plutonium in true solution in the range of grams/litre seem to be achievable, at least in principle. The amount of EDTA necessary is not larger than the total amount of Pu. It is however questionable, specially after taking into account all possible effects of reaction kinetics, whether such high concentrations can be achieved at all under normal environmental conditions. Only experiments under real world conditions can give an answer to this question. (orig./HK) 891 HK 892 AP [de

  8. Investigation of two-phase flow phenomena associated with corrosion in an SF/HLW repository in Opalinus Clay, Switzerland

    International Nuclear Information System (INIS)

    Senger, R.; Marschall, P.; Finsterle, S.

    2008-01-01

    Gas generation from corrosion of the waste canisters and gas accumulation in the backfilled emplacement tunnels is a key issue in the assessment of long-term radiological safety of the proposed repository for spent fuel and high-level waste (SF/HLW) sited in the Opalinus Clay formation of Northern Switzerland. Previous modeling studies indicated a significant pressure buildup in the backfilled emplacement tunnels for those sensitivity runs, where corrosion rates were high and the permeability of the Opalinus Clay was very low. As an extension to those studies, a refined process model of the canister corrosion phenomena has been developed, which accounts not only for the gas generation but also for the water consumption associated with the chemical reaction of corrosion of steel under anaerobic conditions. The simulations with the new process model indicate, that with increasing corrosion rates and decreasing host-rock permeability, pressure buildup increased, as expected. However, the simulations taking into account water consumption show that the pressure buildup is reduced compared to the simulation considering only gas generation. The pressure reduction is enhanced for lower permeability of the Opalinus Clay and for higher corrosion rates, which correspond to higher gas generations rates and higher water consumption rates. Moreover, the simulated two-phase flow patterns in the engineered barrier system (EBS) and surrounding Opalinus Clay show important differences at late time of the gas production phase as the generated gas continues to migrate outward into the surrounding host rock. For the case without water consumption, the water flow indicates overall downward flow due to a change in the overall density of the gas-fluid mixture from that based on the initially prescribed hydrostatic pressure gradient. For the case with water consumption, water flow converges toward the waste canister at a rate corresponding to the water consumption rate associated with the

  9. Development of methodology to construct a generic conceptual model of river-valley evolution for performance assessment of HLW geological disposal

    International Nuclear Information System (INIS)

    Kawamura, Makoto; Tanikawa, Shin-ichi; Yasue, Ken-ichi; Niizato, Tadafumi

    2011-01-01

    In order to assess the long-term safety of a geological disposal system for high-level radioactive waste (HLW), it is important to consider the impact of uplift and erosion, which cannot be precluded on a timescale in the order of several hundred thousand years for many locations in Japan. Geomorphic evolution, caused by uplift and erosion and coupled to climatic and sea-level changes, will impact the geological disposal system due to resulting spatial and temporal changes in the disposal environment. Degradation of HLW barrier performance will be particularly significant when the remnant repository structures near, and are eventually exposed at, the ground surface. In previous studies, fluvial erosion was densified as the key concern in most settings in Japan. Interpretation of the impact of the phenomena at relevant locations in Japan has led to development of a generic conceptual model which contains the features typical at middle reach of rivers. Here, therefore, we present a methodology for development of a generic conceptual model based on best current understanding of fluvial erosion in Japan, which identifies the simplifications and uncertainties involved and assesses their consequences in the context of repository performance. (author)

  10. HLW disposal dilemma

    International Nuclear Information System (INIS)

    Andrei, V.; Glodeanu, F.

    2003-01-01

    The radioactive waste is an inevitable residue from the use of radioactive materials in industry, research and medicine, and from the operation of generating electricity nuclear power stations. The management and disposal of such waste is therefore an issue relevant to almost all countries. Undoubtedly the biggest issue concerning radioactive waste management is that of high level waste. The long-lived nature of some types of radioactive wastes and the associated safety implications of disposal plans have raised concern amongst those who may be affected by such facilities. For these reasons the subject of radioactive waste management has taken on a high profile in many countries. Not one Member State in the European Union can say that their high level waste will be disposed of at a specific site. Nobody can say 'that is where it is going to go'. Now, there is a very broad consensus on the concept of geological disposal. The experts have little, if any doubt that we could safely dispose of the high level wastes. Large sectors of the public continue to oppose to most proposals concerning the siting of repositories. Given this, it is increasingly difficult to get political support, or even political decisions, on such sites. The failure to advance to the next step in the waste management process reinforces the public's initial suspicion and resistance. In turn, this makes the political decisions even harder. In turn, this makes the political decisions even harder. The management of spent fuel from nuclear power plant became a crucial issue, as the cooling pond of the Romanian NPP is reaching saturation. During the autumn of 2000, the plant owner proceeded with an international tendering process for the supply of a dry storage system to be implemented at the Cernavoda station to store the spent fuel from Unit 1 and eventually from Unit 2 for a minimum period of 50 years. The facility is now in operation. As concern the disposal of the spent fuel, the 'wait and see' strategy is now considered. There is a broad agreement that national organizations are responsible for finding their own solutions for disposal of their wastes. However, this does not mean that they have to find solutions within their own countries. This is the concept of international or multinational sheared repositories, well sited and safe facilities operated for the benefit of a number of users, with effective use of shared resources. This may be the only realistic option for some national programmes. On 22nd February 2002 a small group of organisations from 5 countries inaugurated a new association to support the concept of sharing facilities for storage and disposal of all types of long-lived radioactive wastes. The founding members are from Belgium (ONDRAF Waste Agency), Bulgaria (Kozloduy Power Plant), Hungary (PURAM Waste Agency), Japan (Obayashi Corporation) and Switzerland (Colenco Power Engineering, backed by two of the Swiss nuclear power utilities). The Association is open to all organisations sharing its goals; discussions with a range of further potential members are already underway. Romania might consider the regional disposal option. (authors)

  11. Analysis Science Process Skills Content in Chemistry Textbooks Grade XI at Solubility and Solubility Product Concept

    Directory of Open Access Journals (Sweden)

    Bayu Antrakusuma

    2017-12-01

    Full Text Available The aim of this research was to determine the analysis of science process skills in textbooks of chemistry grade XI in SMA N 1 Teras, Boyolali. This research used the descriptive method. The instruments were developed based on 10 indicators of science process skills (observing, classifying, finding a conclusion, predicting, raising the question, hypothesizing, planning an experiment, manipulating materials, and equipment, Applying, and communicating. We analyzed 3 different chemistry textbooks that often used by teachers in teaching. The material analyzed in the book was solubility and solubility product concept in terms of concept explanation and student activity. The results of this research showed different science process skill criteria in 3 different chemistry textbooks. Book A appeared 50% of all aspects of science process skills, in Book B appeared 80% of all aspects of science process skills, and in Book C there was 40% of all aspects of the science process skills. The most common indicator in all books was observing (33.3%, followed by prediction (19.05%, classifying (11.90%, Applying (11.90% , planning experiments (9.52%, manipulating materials and equipment (7.14%, finding conclusion (4.76%, communicating (2.38%. Asking the question and hypothesizing did not appear in textbooks.

  12. Metformin as a prevention and treatment for preeclampsia: effects on soluble fms-like tyrosine kinase 1 and soluble endoglin secretion and endothelial dysfunction.

    Science.gov (United States)

    Brownfoot, Fiona C; Hastie, Roxanne; Hannan, Natalie J; Cannon, Ping; Tuohey, Laura; Parry, Laura J; Senadheera, Sevvandi; Illanes, Sebastian E; Kaitu'u-Lino, Tu'uhevaha J; Tong, Stephen

    2016-03-01

    Preeclampsia is associated with placental ischemia/hypoxia and secretion of soluble fms-like tyrosine kinase 1 and soluble endoglin into the maternal circulation. This causes widespread endothelial dysfunction that manifests clinically as hypertension and multisystem organ injury. Recently, small molecule inhibitors of hypoxic inducible factor 1α have been found to reduce soluble fms-like tyrosine kinase 1 and soluble endoglin secretion. However, their safety profile in pregnancy is unknown. Metformin is safe in pregnancy and is also reported to inhibit hypoxic inducible factor 1α by reducing mitochondrial electron transport chain activity. The purposes of this study were to determine (1) the effects of metformin on placental soluble fms-like tyrosine kinase 1 and soluble endoglin secretion, (2) to investigate whether the effects of metformin on soluble fms-like tyrosine kinase 1 and soluble endoglin secretion are regulated through the mitochondrial electron transport chain, and (3) to examine its effects on endothelial dysfunction, maternal blood vessel vasodilation, and angiogenesis. We performed functional (in vitro and ex vivo) experiments using primary human tissues to examine the effects of metformin on soluble fms-like tyrosine kinase 1 and soluble endoglin secretion from placenta, endothelial cells, and placental villous explants. We used succinate, mitochondrial complex II substrate, to examine whether the effects of metformin on soluble fms-like tyrosine kinase 1 and soluble endoglin secretion were mediated through the mitochondria. We also isolated mitochondria from preterm preeclamptic placentas and gestationally matched control subjects and measured mitochondrial electron transport chain activity using kinetic spectrophotometric assays. Endothelial cells or whole maternal vessels were incubated with metformin to determine whether it rescued endothelial dysfunction induced by either tumor necrosis factor-α (to endothelial cells) or placenta villous

  13. Radionuclide solubility control by solid solutions

    Energy Technology Data Exchange (ETDEWEB)

    Brandt, F.; Klinkenberg, M.; Rozov, K.; Bosbach, D. [Forschungszentrum Juelich GmbH (Germany). Inst. of Energy and Climate Research - Nuclear Waste Management and Reactor Safety (IEK-6); Vinograd, V. [Frankfurt Univ. (Germany). Inst. of Geosciences

    2015-07-01

    The migration of radionuclides in the geosphere is to a large extend controlled by sorption processes onto minerals and colloids. On a molecular level, sorption phenomena involve surface complexation, ion exchange as well as solid solution formation. The formation of solid solutions leads to the structural incorporation of radionuclides in a host structure. Such solid solutions are ubiquitous in natural systems - most minerals in nature are atomistic mixtures of elements rather than pure compounds because their formation leads to a thermodynamically more stable situation compared to the formation of pure compounds. However, due to a lack of reliable data for the expected scenario at close-to equilibrium conditions, solid solution systems have so far not been considered in long-term safety assessments for nuclear waste repositories. In recent years, various solid-solution aqueous solution systems have been studied. Here we present state-of-the art results regarding the formation of (Ra,Ba)SO{sub 4} solid solutions. In some scenarios describing a waste repository system for spent nuclear fuel in crystalline rocks {sup 226}Ra dominates the radiological impact to the environment associated with the potential release of radionuclides from the repository in the future. The solubility of Ra in equilibrium with (Ra,Ba)SO{sub 4} is much lower than the one calculated with RaSO{sub 4} as solubility limiting phase. Especially, the available literature data for the interaction parameter W{sub BaRa}, which describes the non-ideality of the solid solution, vary by about one order of magnitude (Zhu, 2004; Curti et al., 2010). The final {sup 226}Ra concentration in this system is extremely sensitive to the amount of barite, the difference in the solubility products of the end-member phases, and the degree of non-ideality of the solid solution phase. Here, we have enhanced the fundamental understanding regarding (1) the thermodynamics of (Ra,Ba)SO{sub 4} solid solutions and (2) the

  14. Experimental Study of CO2 Solubility in Ionic Liquids and Polyethylene Glycols

    OpenAIRE

    Huang, Huang

    2015-01-01

    The parameter of density, viscosity are tested and fitted with the result of solubility measurement. With series of experiments, this chemical blend is considered with a good effect. The mixture of 50% tetrabutylphosphonium glycine with 50% polyethylene glycol (molecular weight: 400) is the suggested blend, and the most suitable temperature is absorption in 120C and desorption in 60C. But the solubility reduced rapidly from the second cycle of experiment, thus recycled use is not recommended.

  15. Solubility of chromate in a hydrated OPC

    International Nuclear Information System (INIS)

    Leisinger, Sabine M.; Bhatnagar, Amit; Lothenbach, Barbara; Johnson, C. Annette

    2014-01-01

    Highlights: • Solid solutions exist between gypsum and calcium chromate. • The cementitious matrix can bind chromate concentrations up to 0.1 mol/kg. • The chromate binding phase in the cementitious matrix is CrO 4 -ettringite. - Abstract: The knowledge of the chromate binding mechanisms is essential for the prediction of the long-term leachability of cement-based solidified waste containing increased chromate concentrations because of its toxicity and high mobility. In this paper pore water concentrations from OPC doped with varying CaCrO 4 concentrations (0.01–0.8 mol/kg), equilibrated for 28 days were reported. It could be shown that the cementitious matrix can bind chromate concentrations up to 0.1 mol/kg and that the chromate solubility limiting phase was CrO 4 -ettringite, while chromate containing AFm (monochromate) was unstable. Comparison with thermodynamic modelling indicated that at lower chromate dosages chromate was mainly bound by CrO 4 -ettringite while at very high dosages also a mixed CaCrO 4 –CaSO 4 ·2H 2 O phase precipitated. Additional experiments indicated a solubility product of 10 −3.66 for CaCrO 4 and verified the solid solution formation with CaSO 4 ·2H 2 O. Leaching tests indicated a strong chromate binding mainly in the pH range 10.5–13.5, while at pH < 10 very little chromate was bound as ettringite, monocarbonate and C–S–H phases were destabilized. Generally the thermodynamic modeling underestimated chromate uptake indicating that an additional chromate binding possibly on C–S–H or on mixed chromate–carbonate–hydroxide AFm phases

  16. Nanosuspension Technology for Solubilizing Poorly Soluble Drugs

    OpenAIRE

    Deoli Mukesh

    2012-01-01

    Poor water solubility for many drugs and drug candidates remains a major obstacle to their development and clinical application. It is estimated that around 40% of drugs in the pipeline cannot be delivered through the preferred route or in some cases, at all owing to poor water solubility. Conventional formulations to improve solubility suffer from low bioavailability and poor pharmacokinetics, with some carriers rendering systemic toxicities (e.g. Cremophor1 EL). To date, nanoscale systems f...

  17. Soluble theory with massive ghosts

    International Nuclear Information System (INIS)

    Pisarski, R.D.

    1983-01-01

    To investigate the unitarity of asymptotically free, higher-derivative theories, like certain models of quantum gravity, I study a prototype in two space-time dimensions. The prototype is a kind of higher-derivative nonlinear sigma model; it is asymptotically free, exhibits dimensional transmutation, and is soluble in a large-N expansion. The S-matrix elements, constructed from the analytic continuation of the Euclidean Green's functions, conserve probability to approx.O(N -1 ), but violate unitarity at approx.O(N -2 ). The model demonstrates that in higher-derivative theories unitarity, or the lack thereof, cannot be decided without explicit control over the infrared limit. Even so, the results suggest that there may exist some (rather special) asymptotically free, higher-derivative theories which are unitary

  18. Issues concerning the determination of solubility products of sparingly soluble crystalline solids. Solubility of HfO2(cr)

    International Nuclear Information System (INIS)

    Rai, Dhanpat; Kitamura, Akira; Rosso, Kevin M.; Sasaki, Takayuki; Kobayashi, Taishi

    2016-01-01

    Solubility studies were conducted with HfO 2 (cr) solid as a function HCl and ionic strength ranging from 2.0 to 0.004 mol kg -1 . These studies involved (1) using two different amounts of the solid phase, (2) acid washing the bulk solid phase, (3) preheating the solid phase to 1400 C, and (4) heating amorphous HfO 2 (am) suspensions to 90 C to ascertain whether the HfO 2 (am) converts to HfO 2 (cr) and to determine the solubility from the oversaturation direction. Based on the results of these treatments it is concluded that the HfO 2 (cr) contains a small fraction of less crystalline, but not amorphous, material [HfO 2 (lcr)] and this, rather than the HfO 2 (cr), is the solubility-controlling phase in the range of experimental variables investigated in this study. The solubility data are interpreted using both the Pitzer and SIT models and they provide log 10 K 0 values of -(59.75±0.35) and -(59.48±0.41), respectively, for the solubility product of HfO 2 (lcr)[HfO 2 (lcr) + 2H 2 O ↔ Hf 4+ + 4OH - ]. The log 10 of the solubility product of HfO 2 (cr) is estimated to be < -63. The observation of a small fraction of less crystalline higher solubility material is consistent with the general picture that mineral surfaces are often structurally and/or compositionally imperfect leading to a higher solubility than the bulk crystalline solid. This study stresses the urgent need, during interpretation of solubility data, of taking precautions to make certain that the observed solubility behavior for sparingly-soluble solids is assigned to the proper solid phase.

  19. Issues concerning the determination of solubility products of sparingly soluble crystalline solids. Solubility of HfO{sub 2}(cr)

    Energy Technology Data Exchange (ETDEWEB)

    Rai, Dhanpat [Rai Enviro-Chem, LLC, Yachats, OR (United States); Kitamura, Akira [Japan Atomic Energy Agency, Ibaraki (Japan); Rosso, Kevin M. [Pacific Northwest National Laboratory, Richland, WA (United States); Sasaki, Takayuki; Kobayashi, Taishi [Kyoto Univ. (Japan)

    2016-11-01

    Solubility studies were conducted with HfO{sub 2}(cr) solid as a function HCl and ionic strength ranging from 2.0 to 0.004 mol kg{sup -1}. These studies involved (1) using two different amounts of the solid phase, (2) acid washing the bulk solid phase, (3) preheating the solid phase to 1400 C, and (4) heating amorphous HfO{sub 2}(am) suspensions to 90 C to ascertain whether the HfO{sub 2}(am) converts to HfO{sub 2}(cr) and to determine the solubility from the oversaturation direction. Based on the results of these treatments it is concluded that the HfO{sub 2}(cr) contains a small fraction of less crystalline, but not amorphous, material [HfO{sub 2}(lcr)] and this, rather than the HfO{sub 2}(cr), is the solubility-controlling phase in the range of experimental variables investigated in this study. The solubility data are interpreted using both the Pitzer and SIT models and they provide log{sub 10} K{sup 0} values of -(59.75±0.35) and -(59.48±0.41), respectively, for the solubility product of HfO{sub 2}(lcr)[HfO{sub 2}(lcr) + 2H{sub 2}O ↔ Hf{sup 4+} + 4OH{sup -}]. The log{sub 10} of the solubility product of HfO{sub 2}(cr) is estimated to be < -63. The observation of a small fraction of less crystalline higher solubility material is consistent with the general picture that mineral surfaces are often structurally and/or compositionally imperfect leading to a higher solubility than the bulk crystalline solid. This study stresses the urgent need, during interpretation of solubility data, of taking precautions to make certain that the observed solubility behavior for sparingly-soluble solids is assigned to the proper solid phase.

  20. Retrograde curves of solidus and solubility

    International Nuclear Information System (INIS)

    Vasil'ev, M.V.

    1979-01-01

    The investigation was concerned with the constitutional diagrams of the eutectic type with ''retrograde solidus'' and ''retrograde solubility curve'' which must be considered as diagrams with degenerate monotectic transformation. The solidus and the solubility curves form a retrograde curve with a common retrograde point representing the solubility maximum. The two branches of the Aetrograde curve can be described with the aid of two similar equations. Presented are corresponding equations for the Cd-Zn system and shown is the possibility of predicting the run of the solubility curve

  1. Solubility limits of importance to leaching

    International Nuclear Information System (INIS)

    Ogard, A.; Bentley, G.; Bryant, E.; Duffy, C.; Grisham, J.; Norris, E.; Orth, C.; Thomas, K.

    1981-01-01

    The solubilities of some radionuclides, especially rare earths and actinides, may be an important and controlling factor in leaching of waste forms. These solubilities should be measured accurately as a function of pH and not as a part of a multicomponent system. Individual solubilities should be measured as a function of temperature to determine if a kinetic effect is being observed in the data. A negative temperature coefficient of solubility for actinides and rare earths in water would have important consequences for nuclear reactor safety and for the management of nuclear wastes

  2. Novel furosemide cocrystals and selection of high solubility drug forms.

    Science.gov (United States)

    Goud, N Rajesh; Gangavaram, Swarupa; Suresh, Kuthuru; Pal, Sharmistha; Manjunatha, Sulur G; Nambiar, Sudhir; Nangia, Ashwini

    2012-02-01

    Furosemide was screened in cocrystallization experiments with pharmaceutically acceptable coformer molecules to discover cocrystals of improved physicochemical properties, that is high solubility and good stability. Eight novel equimolar cocrystals of furosemide were obtained by liquid-assisted grinding with (i) caffeine, (ii) urea, (iii) p-aminobenzoic acid, (iv) acetamide, (v) nicotinamide, (vi) isonicotinamide, (vii) adenine, and (viii) cytosine. The product crystalline phases were characterized by powder x-ray diffraction, differential scanning calorimetry, infrared, Raman, near IR, and (13) C solid-state NMR spectroscopy. Furosemide-caffeine was characterized as a neutral cocrystal and furosemide-cytosine an ionic salt by single crystal x-ray diffraction. The stability of furosemide-caffeine, furosemide-adenine, and furosemide-cytosine was comparable to the reference drug in 10% ethanol-water slurry; there was no evidence of dissociation of the cocrystal to furosemide for up to 48 h. The other five cocrystals transformed to furosemide within 24 h. The solubility order for the stable forms is furosemide-cytosine > furosemide-adenine > furosemide-caffeine, and their solubilities are approximately 11-, 7-, and 6-fold higher than furosemide. The dissolution rates of furosemide cocrystals were about two times faster than the pure drug. Three novel furosemide compounds of higher solubility and good phase stability were identified in a solid form screen. Copyright © 2011 Wiley Periodicals, Inc.

  3. Implementation of a geological disposal facility (GDF) in the UK by the NDA Radioactive Waste Management Directorate (RWMD): the potential for interaction between the co-located ILW/LLW and HLW/SF components of a GDF - 16306

    International Nuclear Information System (INIS)

    Towler, George; Hicks, Tim; Watson, Sarah; Norris, Simon

    2009-01-01

    In June 2008 the UK government published a 'White Paper' as part of the 'Managing Radioactive Waste Safety' (MRWS) programme to provide a framework for managing higher activity radioactive wastes in the long-term through geological disposal. The White Paper identifies that there are benefits to disposing all of the UK's higher activity wastes (Low and Intermediate Level Waste (LLW and ILW), High Level Waste (HLW), Spent Fuel (SF), Uranium (U) and Plutonium (Pu)) at the same site, and this is currently the preferred option. It also notes that research will be required to support the detailed design and safety assessment in relation to any potentially detrimental interactions between the different modules. Different disposal system designs and associated Engineered Barrier Systems (EBS) will be required for these different waste types, i.e. ILW/LLW and HLW/SF. If declared as waste U would be disposed as ILW and Pu as HLW/SF. The Geological Disposal Facility (GDF) would therefore comprise two co-located modules (respectively for ILW/LLW and HLW/SF). This paper presents an overview of a study undertaken to assess the implications of co-location by identifying the key Thermo-Hydro-Mechanical-Chemical (THMC) interactions that might occur during both the operational and post-closure phases, and their consequences for GDF design, performance and safety. The MRWS programme is currently seeking expressions of interest from communities to host a GDF. Therefore, the study was required to consider a wide range of potential GDF host rocks and consistent, conceptual disposal system designs. Two example disposal concepts (i.e. combinations of host rock, GDF design including wasteform and layout, etc.) were carried forward for detailed assessment and a third for qualitative analysis. Dimensional and 1D analyses were used to identify the key interactions, and 3D models were used to investigate selected interactions in more detail. The results of this study show that it is possible

  4. Salt and cocrystals of sildenafil with dicarboxylic acids: solubility and pharmacokinetic advantage of the glutarate salt.

    Science.gov (United States)

    Sanphui, Palash; Tothadi, Srinu; Ganguly, Somnath; Desiraju, Gautam R

    2013-12-02

    Sildenafil is a drug used to treat erectile dysfunction and pulmonary arterial hypertension. Because of poor aqueous solubility of the drug, the citrate salt, with improved solubility and pharmacokinetics, has been marketed. However, the citrate salt requires an hour to reach its peak plasma concentration. Thus, to improve solubility and bioavailability characteristics, cocrystals and salts of the drug have been prepared by treating aliphatic dicarboxylic acids with sildenafil; the N-methylated piperazine of the drug molecule interacts with the carboxyl group of the acid to form a heterosynthon. Salts are formed with oxalic and fumaric acid; salt monoanions are formed with succinic and glutaric acid. Sildenafil forms cocrystals with longer chain dicarboxylic acids such as adipic, pimelic, suberic, and sebacic acids. Auxiliary stabilization via C-H···O interactions is also present in these cocrystals and salts. Solubility experiments of sildenafil cocrystal/salts were carried out in 0.1N HCl aqueous medium and compared with the solubility of the citrate salt. The glutarate salt and pimelic acid cocrystal dissolve faster than the citrate salt in a two hour dissolution experiment. The glutarate salt exhibits improved solubility (3.2-fold) compared to the citrate salt in water. Solubilities of the binary salts follow an inverse correlation with their melting points, while the solubilities of the cocrystals follow solubilities of the coformer. Pharmacokinetic studies on rats showed that the glutarate salt exhibits doubled plasma AUC values in a single dose within an hour compared to the citrate salt. The high solubility of glutaric acid, in part originating from the strained conformation of the molecule and its high permeability, may be the reason for higher plasma levels of the drug.

  5. Determination of soluble protein contents from RVNRL

    International Nuclear Information System (INIS)

    Wan Manshol Wan Zin; Nurulhuda Othman

    1996-01-01

    This project was carried out to determine the soluble protein contents on RVNRL film vulcanisates, with respect to the RVNRL storage time, gamma irradiation dose absorbed by the latex and the effect of different leaching time and leaching conditions. These three factors are important in the hope to determine the best possible mean of minimizing the soluble protein contents in products made from RVNRL. Within the nine months storage period employed in the study, the results show that, the longer the storage period the less the soluble protein extracted from the film samples. Gamma irradiation dose absorbed by the samples, between 5.3 kGy to 25.2 kGy seems to influence the soluble protein contents of the RVNRL films vulcanisates. The higher the dose the more was the soluble protein extracted from the film samples. At an absorbed dose of 5.3 kGy and 25.2 kGy, the soluble contents were 0. 198 mg/ml and 0.247 mg/ml respectively. At a fixed leaching temperature, the soluble proteins increases with leaching time and at a fixed leaching time, the soluble proteins increases with leaching temperature. ne highest extractable protein contents was determined at a leaching time of 10 minutes and leaching temperature of 90'C The protein analysis were done by using Modified Lowry Method

  6. Solubility Study of Curatives in Various Rubbers

    NARCIS (Netherlands)

    Guo, R.; Talma, Auke; Datta, Rabin; Dierkes, Wilma K.; Noordermeer, Jacobus W.M.

    2008-01-01

    The previous works on solubility of curatives in rubbers were mainly carried out in natural rubber. Not too much information available on dissimilar rubbers and this is important because most of the compounds today are blends of dissimilar rubbers. Although solubility can be expected to certain

  7. Solubility Products of M(II) - Carbonates

    International Nuclear Information System (INIS)

    Grauer, Rolf; Berner, Urs

    1999-01-01

    Many solubility data for M(II) carbonates commonly compiled in tables are contradictory and sometimes obviously wrong. The quality of such data has been evaluated based on the original publications and reliable solubility constants have been selected for the carbonates of Mn, Fe, Co, Ni, Cu, Zn, Cd and Pb with the help of cross-comparisons. (author)

  8. Hansen Solubility Parameters for Octahedral Oligomeric Silsesquioxanes

    Science.gov (United States)

    2012-08-28

    1997, 80, 386-&. 5. Hansen, C. M. The three-dimensional solubility parameter -- key to paint component affinities I. J. Paint Technol. 1967, 39, 104...Chai, J.; Zhang, Q. X.; Han, D. X.; Niu, L. Synthesis and Application of Widely Soluble Graphene Sheets. Langmuir 2010, 26, 12314-12320. 12. Hansen, C

  9. A Colorful Solubility Exercise for Organic Chemistry

    Science.gov (United States)

    Shugrue, Christopher R.; Mentzen, Hans H., II; Linton, Brian R.

    2015-01-01

    A discovery chemistry laboratory has been developed for the introductory organic chemistry student to investigate the concepts of polarity, miscibility, solubility, and density. The simple procedure takes advantage of the solubility of two colored dyes in a series of solvents or solvent mixtures, and the diffusion of colors can be easily…

  10. Relationship Between Urinary Concentrations of Nine Water-soluble Vitamins and their Vitamin Intakes in Japanese Adult Males

    OpenAIRE

    Shibata, Katsumi; Hirose, Junko; Fukuwatari, Tsutomu

    2014-01-01

    Excess water-soluble vitamins are thought to be eliminated in the urine. We have reported a strong relationship between water-soluble vitamin intake and urinary excretion in females. The relationship, however, is not well understood in males. In the present experiment, 10 Japanese male subjects were given a standard Japanese diet for the first week. The subjects remained on the same diet, and a synthesized water-soluble vitamin mixture containing one time the Dietary Reference Intakes (DRIs) ...

  11. Distributions of 14 elements on 60 selected absorbers from two simulant solutions (acid-dissolved sludge and alkaline supernate) for Hanford HLW Tank 102-SY

    International Nuclear Information System (INIS)

    Marsh, S.F.; Svitra, Z.V.; Bowen, S.M.

    1993-10-01

    Sixty commercially available or experimental absorber materials were evaluated for partitioning high-level radioactive waste. These absorbers included cation and anion exchange resins, inorganic exchangers, composite absorbers, and a series of liquid extractants sorbed on porous support-beads. The distributions of 14 elements onto each absorber were measured from simulated solutions that represent acid-dissolved sludge and alkaline supernate solutions from Hanford high-level waste (HLW) Tank 102-SY. The selected elements, which represent fission products (Ce, Cs, Sr, Tc, and Y); actinides (U, Pu, and Am); and matrix elements (Cr, Co, Fe, Mn, Zn, and Zr), were traced by radionuclides and assayed by gamma spectrometry. Distribution coefficients for each of the 1680 element/absorber/solution combinations were measured for dynamic contact periods of 30 min, 2 h, and 6 h to provide sorption kinetics information for the specified elements from these complex media. More than 5000 measured distribution coefficients are tabulated

  12. MIIT: International in-situ testing of simulated HLW forms--preliminary analyses of SRL 165/TDS waste glass and metal systems

    International Nuclear Information System (INIS)

    Wicks, G.G.; Lodding, A.R.; Macedo, P.B.; Molecke, M.A.

    1989-01-01

    The first in-situ tests involving burial of simulated high-level waste (HLW) forms conducted in the United States were started on July 22, 1986. This effort, called the Materials Interface Interactions Tests (MIIT), comprises the largest, most cooperative field testing venture in the international waste management community. Included in the study are over 900 waste form samples comprising 15 different systems supplied by seven countries. Also included are almost 300 potential canister or overpack metal samples of 11 different metals along with more than 500 geologic and backfill specimens. There are a total of 1926 relevant interactions that characterize this effort which is being conducted in the bedded salt site at the Waste Isolation Pilot Plant (WIPP), near Carlsbad, New Mexico

  13. Distributions of 14 elements on 63 absorbers from three simulant solutions (acid-dissolved sludge, acidified supernate, and alkaline supernate) for Hanford HLW Tank 102-SY

    International Nuclear Information System (INIS)

    Marsh, S.F.; Svitra, Z.V.; Bowen, S.M.

    1994-08-01

    As part of the Hanford Tank Waste Remediation System program at Los Alamos, we evaluated 63 commercially available or experimental absorber materials for their ability to remove hazardous components from high-level waste (HLW). These absorbers included cation and anion exchange resins, inorganic exchangers, composite absorbers, and a series of liquid extractants sorbed on porous support-beads. We tested these absorbers with three solutions prepared to simulate acid-dissolved sludge (pH 0.6), acidified supernate (pH 3.5), and alkaline supernate (pH 13.9) from underground storage tank 102-SY at the Hanford Reservation near Richland, Washington. To these simulants we added the appropriate radionuclides and used gamma spectrometry to measure fission products (Ce, Cs, Sr, Tc, and Y), actinides (U, Pu, and Am), and matrix elements (Cr, Co, Fe, Mn, Zn, and Zr). For each of more than 2500 element/absorber/solution combinations, we measured distribution coefficients for dynamic contact periods of 30 min, 2 h, and 6 h to obtain information about sorption kinetics. Because we measured the sorption of many different elements, the tabulated results indicate those elements most likely to interfere with the sorption of elements of greater interest. On the basis of nearly 7500 measured distribution coefficients, we determined that many of these absorbers appear suitable for processing HLW. This study supersedes the previous version of LA-12654, in which results attributed to a solution identified as an alkaline supernate simulant were misleading because that solution contained insufficient hydroxide

  14. Heat-induced alterations in cashew allergen solubility and IgE binding

    Directory of Open Access Journals (Sweden)

    Christopher P. Mattison

    Full Text Available Cashew nuts are an increasingly common cause of food allergy. We compare the soluble protein profile of cashew nuts following heating. SDS-PAGE indicate that heating can alter the solubility of cashew nut proteins. The 11S legumin, Ana o 2, dominates the soluble protein content in ready to eat and mildly heated cashew nuts. However, we found that in dark-roasted cashew nuts, the soluble protein profile shifts and the 2S albumin Ana o 3 composes up to 40% of the soluble protein. Analysis of trypsin-treated extracts by LC/MS/MS indicate changes in the relative number and intensity of peptides. The relative cumulative intensity of the 5 most commonly observed Ana o 1 and 2 peptides are altered by heating, while those of the 5 most commonly observed Ana o 3 peptides remaine relatively constant. ELISA experiments indicate that there is a decrease in rabbit IgG and human serum IgE binding to soluble cashew proteins following heating. Our findings indicate that heating can alter the solubility of cashew allergens, resulting in altered IgE binding. Our results support the use of both Ana o 2 and Ana o 3 as potential cashew allergen diagnostic targets. Keywords: Cashew nut, Food allergy, Immunoglobulin E, Mass-spectrometry, Peptide, Solubility

  15. Final Report Determination Of The Processing Rate Of RPP-WTP HLW Simulants Using A Duramelter J 1000 Vitrification System VSL-00R2590-2, Rev. 0, 8/21/00

    International Nuclear Information System (INIS)

    Kruger, A.A.; Matlack, K.S.; Kot, W.K.; Perez-Cardenas, F.; Pegg, I.L.

    2011-01-01

    This report provides data, analysis, and conclusions from a series of tests that were conducted at the Vitreous State Laboratory of The Catholic University of America (VSL) to determine the melter processing rates that are achievable with RPP-WTP HLW simulants. The principal findings were presented earlier in a summary report (VSL-00R2S90-l) but the present report provides additional details. One of the most critical pieces of information in determining the required size of the RPP-WTP HLW melter is the specific glass production rate in terms of the mass of glass that can be produced per unit area of melt surface per unit time. The specific glass production rate together with the waste loading (essentially, the ratio of waste-in to glass-out, which is determined from glass formulation activities) determines the melt area that is needed to achieve a given waste processing rate with due allowance for system availability. As a consequence of the limited amount of relevant information, there exists, for good reasons, a significant disparity between design-base specific glass production rates for the RPP-WTP LAW and HLW conceptual designs (1.0 MT/m 2 /d and 0.4 MT/m 2 /d, respectively); furthermore, small-scale melter tests with HLW simulants that were conducted during Part A indicated typical processing rates with bubbling of around 2.0 MT/m 2 /d. This range translates into more than a factor of five variation in the resultant surface area of the HLW melter, which is clearly not without significant consequence. It is clear that an undersized melter is undesirable in that it will not be able to support the required waste processing rates. It is less obvious that there are potential disadvantages associated with an oversized melter, over and above the increased capital costs. A melt surface that is consistently underutilized will have poor cold cap coverage, which will result in increased volatilization from the melt (which is generally undesirable) and increased plenum

  16. FINAL REPORT DETERMINATION OF THE PROCESSING RATE OF RPP WTP HLW SIMULANTS USING A DURAMELTER J 1000 VITRIFICATION SYSTEM VSL-00R2590-2 REV 0 8/21/00

    Energy Technology Data Exchange (ETDEWEB)

    KRUGER AA; MATLACK KS; KOT WK; PEREZ-CARDENAS F; PEGG IL

    2011-12-29

    This report provides data, analysis, and conclusions from a series of tests that were conducted at the Vitreous State Laboratory of The Catholic University of America (VSL) to determine the melter processing rates that are achievable with RPP-WTP HLW simulants. The principal findings were presented earlier in a summary report (VSL-00R2S90-l) but the present report provides additional details. One of the most critical pieces of information in determining the required size of the RPP-WTP HLW melter is the specific glass production rate in terms of the mass of glass that can be produced per unit area of melt surface per unit time. The specific glass production rate together with the waste loading (essentially, the ratio of waste-in to glass-out, which is determined from glass formulation activities) determines the melt area that is needed to achieve a given waste processing rate with due allowance for system availability. As a consequence of the limited amount of relevant information, there exists, for good reasons, a significant disparity between design-base specific glass production rates for the RPP-WTP LAW and HLW conceptual designs (1.0 MT/m{sup 2}/d and 0.4 MT/m{sup 2}/d, respectively); furthermore, small-scale melter tests with HLW simulants that were conducted during Part A indicated typical processing rates with bubbling of around 2.0 MT/m{sup 2}/d. This range translates into more than a factor of five variation in the resultant surface area of the HLW melter, which is clearly not without significant consequence. It is clear that an undersized melter is undesirable in that it will not be able to support the required waste processing rates. It is less obvious that there are potential disadvantages associated with an oversized melter, over and above the increased capital costs. A melt surface that is consistently underutilized will have poor cold cap coverage, which will result in increased volatilization from the melt (which is generally undesirable) and

  17. Proposal for geological site selection for L/ILW and HLW repositories. Statement of requirements, procedure and results. Technical report 08-03

    International Nuclear Information System (INIS)

    2008-10-01

    Important steps in the process of managing radioactive wastes have already been implemented in Switzerland. These include the handing and packaging of the waste, waste characterisation and documentation of waste inventories and interim storage along with associated transport. In terms of preparing for deep geological disposal, the necessary scientific and technical work is well advanced and the feasibility of constructing geological repositories that provide the required long-term safety has been successfully demonstrated for all waste types arising in Switzerland. Sufficient knowledge is available to allow the next steps in the selection of repository sites to be defined. The legal framework is also in place and organisational measures have been provided that will allow the tasks to be performed in the coming years to be implemented efficiently. The selection of geological siting regions and sites for repositories in Switzerland will be conducted in three stages. Stage 1 ends with the definition of geological siting regions within which the repository projects will be elaborated in more detail in stages 2 and 3. This report documents and justifies the siting proposals prepared by Nagra for the repositories for low- and intermediate-level waste (L/ILW) and high-level waste (HLW). Formulation of these proposals is conducted in five steps: 1) The waste inventory, which includes reserves for future developments, is allocated to the L/ILW and HLW repositories; 2) Based on this waste allocation, the second step involves defining the barrier and safety concepts for the two repositories. With a view to evaluating the geological siting possibilities, quantitative and qualitative guidelines and requirements on the geology are derived on the basis of these concepts. These relate to the time period to be considered, the space requirements for the repository, the properties of the host rock (depth, thickness, lateral extent, hydraulic conductivity), long-term stability

  18. Evaluation of methods and tools to develop safety concepts and to demonstrate safety for an HLW repository in salt. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Bollingerfehr, W.; Buhmann, D.; Doerr, S.; and others

    2017-03-15

    Salt formations have been the preferred option as host rocks for the disposal of high level radioactive waste in Germany for more than 40 years. During this period comprehensive geological investigations have been carried out together with a broad spectrum of concept and safety related R and D work. The behaviour of an HLW repository in salt formations, particularly in salt domes, has been analysed in terms of assessment of the total system performance. This was first carried out for concepts of generic waste repositories in salt and, since 1998, for a repository concept with specific boundary conditions, taking the geology of the Gorleben salt dome as an example. Suitable repository concepts and designs were developed, the technical feasibility has been proven and operational and long-term safety evaluated. Numerical modelling is an important input into the development of a comprehensive safety case for a waste repository. Significant progress in the development of numerical tools and their application for long-term safe ty assessment has been made in the last two decades. An integrated approach has been used in which the repository concept and relevant scientific and engineering data are combined with the results from iterative safety assessments to increase the clarity and the traceability of the evaluation. A safety concept that takes full credit of the favourable properties of salt formations was developed in the course of the R and D project ISIBEL, which started in 2005. This concept is based on the safe containment of radioactive waste in a specific part of the host rock formation, termed the containment providing rock zone, which comprises the geological barrier, the geotechnical barriers and the compacted backfill. The future evolution of the repository system will be analysed using a catalogue of Features, Events and Processes (FEP), scenario development and numerical analysis, all of which are adapted to suit the safety concept. Key elements of the

  19. Evaluation of methods and tools to develop safety concepts and to demonstrate safety for an HLW repository in salt. Final report

    International Nuclear Information System (INIS)

    Bollingerfehr, W.; Buhmann, D.; Doerr, S.

    2017-03-01

    Salt formations have been the preferred option as host rocks for the disposal of high level radioactive waste in Germany for more than 40 years. During this period comprehensive geological investigations have been carried out together with a broad spectrum of concept and safety related R and D work. The behaviour of an HLW repository in salt formations, particularly in salt domes, has been analysed in terms of assessment of the total system performance. This was first carried out for concepts of generic waste repositories in salt and, since 1998, for a repository concept with specific boundary conditions, taking the geology of the Gorleben salt dome as an example. Suitable repository concepts and designs were developed, the technical feasibility has been proven and operational and long-term safety evaluated. Numerical modelling is an important input into the development of a comprehensive safety case for a waste repository. Significant progress in the development of numerical tools and their application for long-term safe ty assessment has been made in the last two decades. An integrated approach has been used in which the repository concept and relevant scientific and engineering data are combined with the results from iterative safety assessments to increase the clarity and the traceability of the evaluation. A safety concept that takes full credit of the favourable properties of salt formations was developed in the course of the R and D project ISIBEL, which started in 2005. This concept is based on the safe containment of radioactive waste in a specific part of the host rock formation, termed the containment providing rock zone, which comprises the geological barrier, the geotechnical barriers and the compacted backfill. The future evolution of the repository system will be analysed using a catalogue of Features, Events and Processes (FEP), scenario development and numerical analysis, all of which are adapted to suit the safety concept. Key elements of the

  20. In vitro solubility of uranium tetrafluoride with oxidizing medium compared with in vivo solubility in rats

    International Nuclear Information System (INIS)

    Ansoborlo, E.; Chalabreysse, J.; Escallon, S.; Henge-Napoli, M.H.

    1990-01-01

    A simple in vitro solubility test for UF 4 was developed to investigate effects of addition of enzymes, proteins or gases (eg O 2 ) to synthetic biological fluid or Gamble solvent. Tests were made concomitantly with an in vivo inhalation study using male rats. With Gamble solvent alone, UF 4 showed class Y behaviour with dissolution half-time 300-500 days. When O 2 or carbonates were added to Gamble solvent, UF 4 showed class W behaviour (half-time 25-50 days). In the presence of oxygen and pyrogallol, the superoxide ion was formed and UF 4 behaved as class D (half-time 2-3 days). Results correlated with those of the inhalation experiment in which dissolution half-time was 2.5 and 5.2 days. Data also agree with urine monitoring data for workers exposed to UF 4 over 20 years. (author)

  1. Indomethacin solubility estimation in 1,4-dioxane + water mixtures by the extended hildebrand solubility approach

    Directory of Open Access Journals (Sweden)

    Miller A Ruidiaz

    2011-09-01

    Full Text Available Extended Hildebrand Solubility Approach (EHSA was successfully applied to evaluate the solubility of Indomethacin in 1,4-dioxane + water mixtures at 298.15 K. An acceptable correlation-performance of EHSA was found by using a regular polynomial model in order four of the W interaction parameter vs. solubility parameter of the mixtures (overall deviation was 8.9%. Although the mean deviation obtained was similar to that obtained directly by means of an empiric regression of the experimental solubility vs. mixtures solubility parameters, the advantages of EHSA are evident because it requires physicochemical properties easily available for drugs.

  2. Solubilities of Actinide Oxides in the KURT Groundwater

    International Nuclear Information System (INIS)

    Kim, Seung Soo; Baik, Min Hoon; Choi, Jong Won

    2009-12-01

    For the estimation of solubilities of actinides in a deep underground condition, The solubilities of UO 2 , ThO 2 , NpO 2 and Am(OH) 3 in the KURT ground water have been measured under various redox conditions, and their solubilities and aqueous species in the same conditions as the experimental solutions were also calculated by using a geochemical code. Then these results were compared with each other as well as with literature results. For the calculation of solubility of a radionuclide, the thermodynamic data of the radionuclide complex from OECD/NEA, Nagra/PSI, KAERI, JAEA, SKB and recent literatures were collected and compared. Additionally, the methods for the correction of ionic strength and temperature of the solution were described in this report. The analysis techniques and recent research for measurement of species of actinides were also introduced. The concentrations of U, Th and Np dissolved were less than 10 -7 mol/L under Eh≤-0.2 of reducing condition from experiment and calculation, and the solubility of PuO 2 (cr) was estimated as lower than that of UO 2 (cr) by 1 ∼ 2 orders. However if amount of carbonate ion in the ground water increased, the concentration of tetra-valance actinides at pH 8 ∼ 11 would be greatly increased. The 1x10 -6 mol/L of americium might be a little conservative value in KURT groundwater. While carbonate or hydroxo-carbonatec complexes were presumed to be the dominant aqueous species in -0.2 ∼ -0.3 V of Eh and weakly alkaline solution, hydroxo complexes are dominant in strong reducing and high pH solution

  3. Solid dispersions enhance solubility, dissolution, and permeability of thalidomide.

    Science.gov (United States)

    Barea, Silvana A; Mattos, Cristiane B; Cruz, Ariadne C C; Chaves, Vitor C; Pereira, Rafael N; Simões, Claudia M O; Kratz, Jadel M; Koester, Letícia S

    2017-03-01

    Thalidomide (THD) is a BCS class II drug with renewed and growing therapeutic applicability. Along with the low aqueous solubility, additional poor biopharmaceutical properties of the drug, i.e. chemical instability, high crystallinity, and polymorphism, lead to a slow and variable oral absorption. In this view, we developed solid dispersions (SDs) containing THD dispersed in different self-emulsifying carriers aiming at an enhanced absorption profile for the drug. THD was dispersed in lauroyl macrogol-32 glycerides (Gelucire ® 44/14) and α-tocopherol polyethylene glycol succinate (Kolliphor ® TPGS), in the presence or absence of the precipitation inhibitor polyvinylpyrrolidone K30 (PVP K30), by means of the solvent method. Physicochemical analysis revealed the formation of semicrystalline SDs. X-ray diffraction and infrared spectroscopy analyses suggest that the remaining crystalline fraction of the drug in the SDs did not undergo polymorphic transition. The impact of the solubility-enhancing formulations on the THD biopharmaceutical properties was evaluated by several in vitro techniques. The developed SDs were able to increase the apparent solubility of the drug (up to 2-3x the equilibrium solubility) for a least 4 h. Dissolution experiments (paddle method, 75 rpm) in different pHs showed that around 80% of drug dissolved after 120 min (versus 40% of pure crystalline drug). Additionally, we demonstrated the enhanced solubility obtained via SDs could be translated into increased flux in a parallel artificial membrane permeability assay (PAMPA). In summary, the results demonstrate that SDs could be considered an interesting and unexplored strategy to improve the biopharmaceutical properties of THD, since SDs of this important drug have yet to be reported.

  4. Removal of soluble toxic metals from water

    International Nuclear Information System (INIS)

    Buckley, L.P.; Vijayan, S.; McConeghy, G.J.; Maves, S.R.; Martin, J.F.

    1990-05-01

    The removal of selected, soluble toxic metals from aqueous solutions has been accomplished using a combination of chemical treatment and ultrafiltration. The process has been evaluated at the bench-scale and is undergoing pilot-scale testing. Removal efficiencies in excess of 95-99% have been realized. The test program at the bench-scale investigated the limitations and established the optimum range of operating parameters for the process, while the tests conducted with the pilot-scale process equipment are providing information on longer-term process efficiencies, effective processing rates, and fouling potential of the membranes. With the typically found average concentrations of the toxic metals in groundwaters at Superfund sites used as the feed solution, the process has decreased levels up to 100-fold or more. Experiments were also conducted with concentrated solutions to determine their release from silica-based matrices. The solidified wastes were subjected to EP Toxicity test procedures and met the criteria successfully. The final phase of the program involving a field demonstration at a uranium tailings site will be outlined

  5. Factors affecting actinide solubility in a repository for spent fuel, 1

    International Nuclear Information System (INIS)

    Snellman, Margit

    1986-07-01

    The main tasks in the study were to get information on the chemical conditions in a repository for spent fuel and information on factors affecting releases of actinides from spent fuel and solubility of actinides in a repository for spent fuel. The work in this field started at the Reactor Laboratory of the Technical Research Centre of Finland (VTT) in 1982. This is a report on the effects on the main parameters, Eh, pH, carbonate, organic compounds, colloids, microbes and radiation on the actinide solubility in the nearfield of the repository. Another task has been to identify available models and reported experience from actinide solubility calculations with different codes. 167 refs

  6. Tainting by short-term exposure of Atlantic salmon to water soluble petroleum hydrocarbons

    International Nuclear Information System (INIS)

    Ackman, R.G.; Heras, H.

    1992-01-01

    Experiments were conducted to examine the extent of tainting of salmon by exposure to the soluble fraction of petroleum hydrocarbons. The experiments were conducted on Atlantic salmon in tanks containing seawater artificially contaminated at three different concentrations with the soluble fraction of a North Sea crude. The salmon flesh was analyzed by gas chromatography and taste tests were conducted on cooked salmon samples to determine the extent of tainting. Salmon in control tanks with uncontaminated seawater had muscle accumulations of total hydrocarbons of ca 1 ppM. The muscle accumulations of total hydrocarbons in the salmon were 13.5 ppM, 25.6 ppM, and 31.3 ppM for water soluble fraction concentrations of 0.45, 0.87, and 1.54 ppM respectively. The threshold for taint was clearly inferred to be less than 0.45 ppM of water soluble fraction. 18 refs., 2 figs

  7. Soluble Non-ammonia Nitrogen in Ruminal and Omasal Digesta of Korean Native Steers Supplemented with Soluble Proteins

    Directory of Open Access Journals (Sweden)

    C. W. Choi

    2012-09-01

    Full Text Available An experiment was conducted to study the effect of soluble protein supplements on concentration of soluble non-ammonia nitrogen (SNAN in the liquid phase of ruminal (RD and omasal digesta (OD of Korean native steers, and to investigate diurnal pattern in SNAN concentration in RD and OD. Three ruminally cannulated Korean native steers in a 3×3 Latin square design consumed a basal diet of rice straw and corn-based concentrate (control, and that supplemented (kg/d DM basis with intact casein (0.24; IC or acid hydrolyzed casein (0.46; AHC. Ruminal digesta was sampled using a vacuum pump, whereas OD was collected using an omasal sampling system at 2.0 h intervals after a morning feeding. The SNAN fractions (free amino acid (AA, peptide and soluble protein in RD and OD were assessed using the ninhydrin assay. Concentrations of free AA and total SNAN in RD were significantly (p<0.05 lower than those in OD. Although free AA concentration was relatively high, mean peptide was quantitatively the most important fraction of total SNAN in both RD and OD, indicating that degradation of peptide to AA rather than hydrolysis of soluble protein to peptide or deamination may be the most limiting step in rumen proteolysis of Korean native steers. Diurnal variation in peptide concentration in OD for the soluble protein supplemented diets during the feeding cycle peaked 2 h post-feeding and decreased thereafter whereas that for the control was relatively constant during the entire feeding cycle. Diurnal variation in peptide concentration was rather similar between RD and OD.

  8. Serum Soluble Corin is Decreased in Stroke.

    Science.gov (United States)

    Peng, Hao; Zhu, Fangfang; Shi, Jijun; Han, Xiujie; Zhou, Dan; Liu, Yan; Zhi, Zhongwen; Zhang, Fuding; Shen, Yun; Ma, Juanjuan; Song, Yulin; Hu, Weidong

    2015-07-01

    Soluble corin was decreased in coronary heart disease. Given the connections between cardiac dysfunction and stroke, circulating corin might be a candidate marker of stroke risk. However, the association between circulating corin and stroke has not yet been studied in humans. Here, we aimed to examine the association in patients wtith stroke and community-based healthy controls. Four hundred eighty-one patients with ischemic stroke, 116 patients with hemorrhagic stroke, and 2498 healthy controls were studied. Serum soluble corin and some conventional risk factors of stroke were examined. Because circulating corin was reported to be varied between men and women, the association between serum soluble corin and stroke was evaluated in men and women, respectively. Patients with ischemic and hemorrhagic stroke had a significantly lower level of serum soluble corin than healthy controls in men and women (all P values, stroke than men in the highest quartile. Women in the lowest quartile of serum soluble corin were also more likely to have ischemic (OR, 3.10; 95% confidence interval, 1.76-5.44) and hemorrhagic (OR, 8.54; 95% confidence interval, 2.35-31.02) stroke than women in the highest quartile. ORs of ischemic and hemorrhagic stroke were significantly increased with the decreasing levels of serum soluble corin in men and women (all P values for trend, stroke compared with healthy controls. Our findings raise the possibility that serum soluble corin may have a pathogenic role in stroke. © 2015 American Heart Association, Inc.

  9. Solubility limited radionuclide transport through geologic media

    International Nuclear Information System (INIS)

    Muraoka, Susumu; Iwamoto, Fumio; Pigford, T.H.

    1980-11-01

    Prior analyses for the migration of radionuclides neglect solubility limits of resolved radionuclide in geologic media. But actually some of the actinides may appear in chemical forms of very low solubility. In the present report we have proposed the migration model with no decay parents in which concentration of radionuclide is limited in concentration of solubility in ground water. In addition, the analytical solutions of the space-time-dependent concentration are presented in the case of step release, band release and exponential release. (author)

  10. Residual nilpotence and residual solubility of groups

    International Nuclear Information System (INIS)

    Mikhailov, R V

    2005-01-01

    The properties of the residual nilpotence and the residual solubility of groups are studied. The main objects under investigation are the class of residually nilpotent groups such that each central extension of these groups is also residually nilpotent and the class of residually soluble groups such that each Abelian extension of these groups is residually soluble. Various examples of groups not belonging to these classes are constructed by homological methods and methods of the theory of modules over group rings. Several applications of the theory under consideration are presented and problems concerning the residual nilpotence of one-relator groups are considered.

  11. Water Soluble Polymers for Pharmaceutical Applications

    Directory of Open Access Journals (Sweden)

    Veeran Gowda Kadajji

    2011-11-01

    Full Text Available Advances in polymer science have led to the development of novel drug delivery systems. Some polymers are obtained from natural resources and then chemically modified for various applications, while others are chemically synthesized and used. A large number of natural and synthetic polymers are available. In the present paper, only water soluble polymers are described. They have been explained in two categories (1 synthetic and (2 natural. Drug polymer conjugates, block copolymers, hydrogels and other water soluble drug polymer complexes have also been explained. The general properties and applications of different water soluble polymers in the formulation of different dosage forms, novel delivery systems and biomedical applications will be discussed.

  12. Molecular Thermodynamic Modeling of Mixed Solvent Solubility

    DEFF Research Database (Denmark)

    Ellegaard, Martin Dela; Abildskov, Jens; O’Connell, John P.

    2010-01-01

    A method based on statistical mechanical fluctuation solution theory for composition derivatives of activity coefficients is employed for estimating dilute solubilities of 11 solid pharmaceutical solutes in nearly 70 mixed aqueous and nonaqueous solvent systems. The solvent mixtures range from...... nearly ideal to strongly nonideal. The database covers a temperature range from 293 to 323 K. Comparisons with available data and other existing solubility methods show that the method successfully describes a variety of observed mixed solvent solubility behaviors using solute−solvent parameters from...

  13. The surface mock-up KENTEX: on the thermal-hydro-mechanical behaviors in the buffer of a Korean HLW repository

    International Nuclear Information System (INIS)

    Lee, Jae Owan; Cho, Won Jin; Choi, Jong Won

    2008-01-01

    The concept for a disposal of high-level wastes (HLW) in Korea is based upon a multi barrier system composed of engineered barriers and its surrounding plutonic rock (Kang et. al., 2002). A repository is constructed in a bedrock of several hundred meters in depth below the ground surface. The engineered barrier system (EBS), which is similar to the configuration considered by many other countries, consists of the HLW-encapsulating disposal container, the buffer between the container and the wall of a borehole, and the backfill in the inside space of the emplacement room, to isolate the HLW from the surrounding rock masses. The engineering performance of a HLW repository is dependent, to a large extent, upon the thermal-hydro-mechanical (THM) behaviors in the buffer which are complicated by the processes such as the decay heat generated from the HLW, the ground water flowing in from the surrounding host rock, and the swelling pressure exerted by compacted bentonite. For this reason, the Korea Atomic Energy Research Institute (KAERI), to investigate the THM behaviors in the buffer of the Korean reference disposal system (KRS), planned large-scale tests to be conducted in two stages: a surface mock-up and then a full-scale 'in situ' test. This paper deals with the surface mock-up called as 'KENTEX' and presents the THM behaviors in the buffer which have been investigated from the KENTEX test. The KENTEX is a third scale of the KRS. It consists of five major components: a heating system, a confining cylinder, a hydration tank, bentonite blocks, and sensors and instruments. The heating system measures 0.41 m in diameter and 0.68 m in length, which includes three heating elements in its inside, capable of supplying a thermal power of 1 kW each. The confining cylinder, which plays a role of the wall of a borehole excavated in the host rock, is a steel body with a length of 1.36 m and an inner diameter of 0.75 m, the inside wall of which is lined with layers of geotextile

  14. Solubility of carbohydrates in heavy water.

    Science.gov (United States)

    Cardoso, Marcus V C; Carvalho, Larissa V C; Sabadini, Edvaldo

    2012-05-15

    The solubility of several mono-(glucose and xylose), di-(sucrose and maltose), tri-(raffinose) and cyclic (α-cyclodextrin) saccharides in H(2)O and in D(2)O were measured over a range of temperatures. The solution enthalpies for the different carbohydrates in the two solvents were determined using the vant' Hoff equation and the values in D(2)O are presented here for the first time. Our findings indicate that the replacement of H(2)O by D(2)O remarkably decreases the solubilities of the less soluble carbohydrates, such as maltose, raffinose and α-cyclodextrin. On the other hand, the more soluble saccharides, glucose, xylose, and sucrose, are practically insensitive to the H/D replacement in water. Copyright © 2012 Elsevier Ltd. All rights reserved.

  15. Enhancement of Solubility and Bioavailability of Candesartan ...

    African Journals Online (AJOL)

    Purpose: To enhance the otherwise poor solubility and bioavailability of candesartan cilexetil (CDS). Methods: This ... PEG 6000-based solid dispersions showed 1st order drug release kinetics. ..... the liver due to quercetin's inhibitory effect on.

  16. An Introduction to the Understanding of Solubility.

    Science.gov (United States)

    Letcher, Trevor M.; Battino, Rubin

    2001-01-01

    Explores different solubility processes and related issues, including the second law of thermodynamics and ideal mixtures, real liquids, intermolecular forces, and solids in liquids or gases in liquids. (Contains 22 references.) (ASK)

  17. Progress in the research of neptunium solubility

    International Nuclear Information System (INIS)

    Jiang Tao; Liu Yongye; Yao Jun

    2012-01-01

    237 Np is considered a possible long-term potential threat for environment, because of its long half-life, high toxicity and its mobile nature under aerobic conditions due to the high chemical stability of its pentavalent state. Therefore 237 Np is considered as one of high-level radioactive waste and need to be disposed in deep geologic disposal repository. The dissolution behavior is an important aspect of migration research. The solubility is considered very important for high level waste geological disposal safety and environmental evaluation. The solubility determines the maximum concentration of the discharge, and then it is initial concentration of the radionuclides migration to the environment. The solubility impact directly on radionuclides migration in host rock, and can be used to predict the concentration and speciation of radionuclides in groundwater around disposal sites many years later. This paper focused on research results of the solubility, some proposals for Np dissolution chemistry research were also been suggested. (authors)

  18. Solubility Products of M(II) - Carbonates

    Energy Technology Data Exchange (ETDEWEB)

    Grauer, Rolf; Berner, Urs [ed.

    1999-01-01

    Many solubility data for M(II) carbonates commonly compiled in tables are contradictory and sometimes obviously wrong. The quality of such data has been evaluated based on the original publications and reliable solubility constants have been selected for the carbonates of Mn, Fe, Co, Ni, Cu, Zn, Cd and Pb with the help of cross-comparisons. (author) translated from a PSI internal report written in German in 1994 (TM-44-94-05). 5 figs., 1 tab., 68 refs.

  19. Hydrogen solubility in polycrystalline - and nonocrystalline niobium

    International Nuclear Information System (INIS)

    Ishikawa, T.T.; Silva, J.R.G. da

    1981-01-01

    Hydrogen solubility in polycrystalline and monocrystalline niobium was measured in the range 400 0 C to 1000 0 C at one atmosphere hydrogen partial pressure. The experimental technique consists of saturation of the solvent metal with hydrogen, followed by quenching and analysis of the solid solution. It is presented solubility curves versus reciprocal of the absolute doping temperature, associated with their thermodynamical equation. (Author) [pt

  20. Respiratory carcinogenicity assessment of soluble nickel compounds.

    OpenAIRE

    Oller, Adriana R

    2002-01-01

    The many chemical forms of nickel differ in physicochemical properties and biological effects. Health assessments for each main category of nickel species are needed. The carcinogenicity assessment of water-soluble nickel compounds has proven particularly difficult. Epidemiologic evidence indicates an association between inhalation exposures to nickel refinery dust containing soluble nickel compounds and increased risk of respiratory cancers. However, the nature of this association is unclear...

  1. Correlation of Helium Solubility in Liquid Nitrogen

    Science.gov (United States)

    VanDresar, Neil T.; Zimmerli, Gregory A.

    2012-01-01

    A correlation has been developed for the equilibrium mole fraction of soluble gaseous helium in liquid nitrogen as a function of temperature and pressure. Experimental solubility data was compiled and provided by National Institute of Standards and Technology (NIST). Data from six sources was used to develop a correlation within the range of 0.5 to 9.9 MPa and 72.0 to 119.6 K. The relative standard deviation of the correlation is 6.9 percent.

  2. Hydrothermal solubility of uraninite. Final technical report

    International Nuclear Information System (INIS)

    Parks, G.A.; Pohl, D.C.

    1985-01-01

    Experimental measurements of the solubility of UO 2 from 100 to 300 0 C under 500 bars H 2 , in NaCl solutions at pH from 1 to 8 do not agree with solubilities calculated using existing thermodynamic databases. For pH 2 (hyd) has precipitated and is controlling solubility. For pH > 8, solubilities at all temperatures are much lower than predicted, suggesting that the U(OH)/sub delta/ - complex is much weaker than predicted. Extrapolated to 25 0 C, high pH solubility agrees within experimental error with the upper limit suggested by Ryan and Rai (1983). In the pH range 2 to 6, solubilities are up to three orders of magnitude lower than predicted for temperatures exceeding 200 0 C and up to two orders higher than predicted at lower temperatures. pH dependence in this region is negligible suggesting that U(OH) 4 (aq) predominates, thus the stability of this species is higher than presently estimated at low temperatures, but the enthalpy of solution is smaller. A low maximum observed near pH approx. =3 is presently unexplained. 40 refs., 16 figs., 12 tabs

  3. Uranium solubility and speciation in ground water

    International Nuclear Information System (INIS)

    Ollila, K.

    1985-04-01

    The purpose of this study has been to assess the solubility and possible species of uranium in groundwater at the disposal conditions of spent fuel. The effects of radiolysis and bentonite are considered. The assessment is based on the theoretical calculations found in the literature. The Finnish experimental results are included. The conservative estimate for uranium solubility under the oxidizing conditions caused by alpha radiolysis is based on the oxidation of uranium to the U(VI) state and formation of carbonate complex. For the groundwater with the typical carbonate content of 275 mg/l and the high carbonate content of 485 mg/l due to bentonite, the solubility values of 360 mg u/l and 950 mg U/l, are obtained, respectively. The experimental results predict considerably lower values, 0.5-20 mg U/l. The solubility of uranium under the undisturbed reducing conditions may be calculated based on the hydrolysis, carbonate complexation and redox reactions. The results vary considerably depending on the thermodynamic data used. The wide ranges of the most important groundwater parameters are seen in the solubility values. The experimental results show the same trends. As a conservative value for the solubility in reducing groundwater 50-500 μg U/l is estimated. (author)

  4. Sibutramine characterization and solubility, a theoretical study

    Science.gov (United States)

    Aceves-Hernández, Juan M.; Nicolás Vázquez, Inés; Hinojosa-Torres, Jaime; Penieres Carrillo, Guillermo; Arroyo Razo, Gabriel; Miranda Ruvalcaba, René

    2013-04-01

    Solubility data from sibutramine (SBA) in a family of alcohols were obtained at different temperatures. Sibutramine was characterized by using thermal analysis and X-ray diffraction technique. Solubility data were obtained by the saturation method. The van't Hoff equation was used to obtain the theoretical solubility values and the ideal solvent activity coefficient. No polymorphic phenomena were found from the X-ray diffraction analysis, even though this compound is a racemic mixture of (+) and (-) enantiomers. Theoretical calculations showed that the polarisable continuum model was able to reproduce the solubility and stability of sibutramine molecule in gas phase, water and a family of alcohols at B3LYP/6-311++G (d,p) level of theory. Dielectric constant, dipolar moment and solubility in water values as physical parameters were used in those theoretical calculations for explaining that behavior. Experimental and theoretical results were compared and good agreement was obtained. Sibutramine solubility increased from methanol to 1-octanol in theoretical and experimental results.

  5. Solubility of Aragonite in Subduction Water-Rich Fluids

    Science.gov (United States)

    Daniel, I.; Facq, S.; Petitgirard, S.; Cardon, H.; Sverjensky, D. A.

    2017-12-01

    Carbonate dissolution in subduction zone fluids is critical to the carbon budget in subduction zones. Depending on the solubility of carbonate minerals in aqueous fluids, the subducting lithosphere may be either strongly depleted and the mantle metasomatized if the solubility is high, as recently suggested by natural samples or transport carbon deeper into the Earth's mantle if the solubility is low enough [1, 2]. Dissolution of carbonate minerals strongly depends on pressure and temperature as well as on the chemistry of the fluid, leading to a highly variable speciation of aqueous carbon. Thanks to recent advances in theoretical aqueous geochemistry [3, 4], combined experimental and theoretical efforts now allow the investigation of speciation and solubility of carbonate minerals in aqueous fluids at PT conditions higher than previously feasible [4, 5]. In this study, we present new in situ X-ray fluorescence measurements of aragonite dissolution up to 5 GPa and 500°C and the subsequent thermodynamic model of aragonite solubility in aqueous fluids thanks to the Deep Earth Water model. The amount of dissolved aragonite in the fluid was calculated from challenging and unprecedented measurements of the Ca fluorescence K-lines at low-energy. Experiments were performed at the ESRF, beamline ID27 using a dedicated design of an externally-heated diamond anvil cell and an incident high-flux and highly focused monochromatic X-Ray beam at 20 keV. The results show a spectacularly high solubility of aragonite at HP-HT in water, further enhanced in presence of NaCl and silica in the solution. [1] Frezzotti, M. L. et al. (2011) doi:10.1038/ngeo1246. [2] Ague, J. J. and Nicolescu, S. (2014) doi:10.1038/ngeo2143. [3] Pan, D. et al. (2013) doi: 10.1073/pnas.1221581110. [4] Sverjensky, D. A et al. (2014) doi: 10.1016/j.gca.2013.12.019. [5] Facq, S. et al. (2014) doi: 10.1016/j.gca.2014.01.030.

  6. Final Report Integrated DM1200 Melter Testing Of Bubbler Configurations Using HLW AZ-101 Simulants VSL-04R4800-4, Rev. 0, 10/5/04

    International Nuclear Information System (INIS)

    Kruger, A.A.; Matlack, K.S.; Gong, W.; Bardakci, T.; D'Angelo, N.A.; Lutze, W.; Callow, R.A.; Brandys, M.; Kot, W.K.; Pegg, I.L.

    2011-01-01

    This report documents melter and off-gas performance results obtained on the DM1200 HLW Pilot Melter during processing of AZ-101 HLW simulants. The tests reported herein are a subset of six tests from a larger series of tests described in the Test Plan for the work; results from the other tests have been reported separately. The solids contents of the melter feeds were based on the WTP baseline value for the solids content of the feeds from pretreatment which changed during these tests from 20% to 15% undissolved solids resulting in tests conducted at two feed solids contents. Based on the results of earlier tests with single outlet 'J' bubblers, initial tests were performed with a total bubbling rate of 651 pm. The first set of tests (Tests 1A-1E) addressed the effects of skewing this total air flow rate back and forth between the two installed bubblers in comparison to a fixed equal division of flow between them. The second set of tests (2A-2D) addressed the effects of bubbler depth. Subsequently, as the location, type and number of bubbling outlets were varied, the optimum bubbling rate for each was determined. A third (3A-3C) and fourth (8A-8C) set of tests evaluated the effects of alternative bubbler designs with two gas outlets per bubbler instead of one by placing four bubblers in positions simulating multiple-outlet bubblers. Data from the simulated multiple outlet bubblers were used to design bubblers with two outlets for an additional set of tests (9A-9C). Test 9 was also used to determine the effect of small sugar additions to the feed on ruthenium volatility. Another set of tests (10A-10D) evaluated the effects on production rate of spiking the feed with chloride and sulfate. Variables held constant to the extent possible included melt temperature, plenum temperature, cold cap coverage, the waste simulant composition, and the target glass composition. The feed rate was increased to the point that a constant, essentially complete, cold cap was achieved

  7. FINAL REPORT INTEGRATED DM1200 MELTER TESTING OF BUBBLER CONFIGURATIONS USING HLW AZ-101 SIMULANTS VSL-04R4800-4 REV 0 10/5/04

    Energy Technology Data Exchange (ETDEWEB)

    KRUGER AA; MATLACK KS; GONG W; BARDAKCI T; D' ANGELO NA; LUTZE W; CALLOW RA; BRANDYS M; KOT WK; PEGG IL

    2011-12-29

    This report documents melter and off-gas performance results obtained on the DM1200 HLW Pilot Melter during processing of AZ-101 HLW simulants. The tests reported herein are a subset of six tests from a larger series of tests described in the Test Plan for the work; results from the other tests have been reported separately. The solids contents of the melter feeds were based on the WTP baseline value for the solids content of the feeds from pretreatment which changed during these tests from 20% to 15% undissolved solids resulting in tests conducted at two feed solids contents. Based on the results of earlier tests with single outlet 'J' bubblers, initial tests were performed with a total bubbling rate of 651 pm. The first set of tests (Tests 1A-1E) addressed the effects of skewing this total air flow rate back and forth between the two installed bubblers in comparison to a fixed equal division of flow between them. The second set of tests (2A-2D) addressed the effects of bubbler depth. Subsequently, as the location, type and number of bubbling outlets were varied, the optimum bubbling rate for each was determined. A third (3A-3C) and fourth (8A-8C) set of tests evaluated the effects of alternative bubbler designs with two gas outlets per bubbler instead of one by placing four bubblers in positions simulating multiple-outlet bubblers. Data from the simulated multiple outlet bubblers were used to design bubblers with two outlets for an additional set of tests (9A-9C). Test 9 was also used to determine the effect of small sugar additions to the feed on ruthenium volatility. Another set of tests (10A-10D) evaluated the effects on production rate of spiking the feed with chloride and sulfate. Variables held constant to the extent possible included melt temperature, plenum temperature, cold cap coverage, the waste simulant composition, and the target glass composition. The feed rate was increased to the point that a constant, essentially complete, cold cap was

  8. Soluble FLT-1 rules placental destiny.

    Science.gov (United States)

    Yamashita, Michiko; Kumasawa, Keiichi; Nakamura, Hitomi; Kimura, Tadashi

    2018-02-19

    Placenta previa is an abnormality in which the placenta covers the internal uterine os, and it can cause serious morbidity and mortality in both mother and fetus due to catastrophic hemorrhage. Some pregnant women recover from placenta previa due to a phenomenon called "migration." However, the mechanism of "migration" of the placenta has not been elucidated. Human placentas were collected from patients with placenta previa and those with no abnormal placentation (control). A microarray analysis was performed to detect the genes up- or down-regulated only in the caudal part in the previa group. Specific mRNA expression was evaluated using real-time quantitative reverse transcription PCR (qRT-PCR). Unilateral uterine artery ablation of 8.5 dpc mice was performed to reproduce the reduction of placental blood supply, and weights of the placentas and fetuses were evaluated in 18.5 dpc. Specific mRNA expression was also evaluated in mice placentas. According to the result of the microarray analysis, we focused on soluble fms-like tyrosine kinase-1 (sFLT-1) and hypoxia-inducible factor-1 (HIF-1) alpha. The sFLT-1 expression level is locally high in the caudal part of the human placenta in patients with placenta previa. In mice experiments, the weights of the placentas and fetuses were significantly smaller in the ablation side than those in the control side, and the sFlt-1 expression level was significantly higher in the ablation side than in the control side. Our study suggests that "migration" of the placenta is derived from placental degeneration at the caudal part of the placenta, and sFlt-1 plays a role in this placental degeneration. Copyright © 2018 Elsevier Inc. All rights reserved.

  9. Preparation, characterization and solubility product constant of AmOHCO3

    International Nuclear Information System (INIS)

    Silva, R.J.

    1985-01-01

    An investigation into the nature and solubility of a stable solid phase formed by a trivalent actinide, 243 Am 3+ , in dilute aqueous carbonate solutions was conducted. The compound exhibited an x-ray powder diffraction pattern which was nearly identical to that reported for NdOHCO 3 - type A. The pattern could be indexed in the orthorhombic system with unit cell parameters a = 4.958, b = 8.487, and c = 7.215 A. The steady-state solubility of the compound was determined from the results of both dissolution and precipitation experiments. The average solubility product quotient for 0.1M ionic strength, 25 +- 1 0 C and 1 atmosphere pressure was found to be 583 +- 206. The solubility product constant for zero ionic strength was estimated to be 335 +- 120. 22 references, 3 tables

  10. Review: kinetics of water-soluble contrast media in the central nervous system

    International Nuclear Information System (INIS)

    Sage, M.R.

    1983-01-01

    In neuroradiology, intraarterial, intravenous, and intrathecal injections of water-soluble contrast media are made. With the growing importance of water-soluble myelography, interventional angiography, and enhanced computed tomography (CT), it is essential to have a clear understanding of the response of the nervous system to such procedures. The blood, cerebrospinal fluid (CSF), and extracellular fluid of the parenchyma form the fluid compartments of the brain with three interfaces between, namely, the blood-brain interface, the CSF-brain interface, and the blood-CSF interface. One of more of these interfaces are exposed to water-soluble contrast media after intraarterial, intravenous, or intrathecal administration. The behavior of water-soluble contrast media at these interfaces is discussed on the basis of local experience and a review of the literature

  11. Preparation, characterization and solubility product constant of AmOHCO/sub 3/

    Energy Technology Data Exchange (ETDEWEB)

    Silva, R.J.

    1985-01-12

    An investigation into the nature and solubility of a stable solid phase formed by a trivalent actinide, /sup 243/Am/sup 3 +/, in dilute aqueous carbonate solutions was conducted. The compound exhibited an x-ray powder diffraction pattern which was nearly identical to that reported for NdOHCO/sub 3/ - type A. The pattern could be indexed in the orthorhombic system with unit cell parameters a = 4.958, b = 8.487, and c = 7.215 A. The steady-state solubility of the compound was determined from the results of both dissolution and precipitation experiments. The average solubility product quotient for 0.1M ionic strength, 25 +- 1/sup 0/C and 1 atmosphere pressure was found to be 583 +- 206. The solubility product constant for zero ionic strength was estimated to be 335 +- 120. 22 references, 3 tables.

  12. Solubility of iron from combustion source particles in acidic media linked to iron speciation.

    Science.gov (United States)

    Fu, Hongbo; Lin, Jun; Shang, Guangfeng; Dong, Wenbo; Grassian, Vichi H; Carmichael, Gregory R; Li, Yan; Chen, Jianmin

    2012-10-16

    In this study, iron solubility from six combustion source particles was investigated in acidic media. For comparison, a Chinese loess (CL) dust was also included. The solubility experiments confirmed that iron solubility was highly variable and dependent on particle sources. Under dark and light conditions, the combustion source particles dissolved faster and to a greater extent relative to CL. Oil fly ash (FA) yielded the highest soluble iron as compared to the other samples. Total iron solubility fractions measured in the dark after 12 h ranged between 2.9 and 74.1% of the initial iron content for the combustion-derived particles (Oil FA > biomass burning particles (BP) > coal FA). Ferrous iron represented the dominant soluble form of Fe in the suspensions of straw BP and corn BP, while total dissolved Fe presented mainly as ferric iron in the cases of oil FA, coal FA, and CL. Mössbauer measurements and TEM analysis revealed that Fe in oil FA was commonly presented as nanosized Fe(3)O(4) aggregates and Fe/S-rich particles. Highly labile source of Fe in corn BP could be originated from amorphous Fe form mixed internally with K-rich particles. However, Fe in coal FA was dominated by the more insoluble forms of both Fe-bearing aluminosilicate glass and Fe oxides. The data presented herein showed that iron speciation varies by source and is an important factor controlling iron solubility from these anthropogenic emissions in acidic solutions, suggesting that the variability of iron solubility from combustion-derived particles is related to the inherent character and origin of the aerosols themselves. Such information can be useful in improving our understanding on iron solubility from combustion aerosols when they undergo acidic processing during atmospheric transport.

  13. Formulation of a Novel Nano emulsion System for Enhanced Solubility of a Sparingly Water Soluble Antibiotic, Clarithromycin

    International Nuclear Information System (INIS)

    Vatsraj, S.; Pathak, H.; Chauhan, K.

    2014-01-01

    The sparingly water soluble property of majority of medicinally significant drugs acts as a potential barrier towards its utilization for therapeutic purpose. The present study was thus aimed at development of a novel oil-in-water (o/w) nano emulsion (NE) system having ability to function as carrier for poorly soluble drugs with clarithromycin as a model antibiotic. The therapeutically effective concentration of clarithromycin, 5 mg/mL, was achieved using polysorbate 80 combined with olive oil as lipophilic counterion. A three-level three-factorial central composite experimental design was utilized to conduct the experiments. The effects of selected variables, polysorbate 80 and olive oil content and concentration of polyvinyl alcohol, were investigated. The particle size of clarithromycin for the optimized formulation was observed to be 30 nm. The morphology of the nano emulsion was explored using transmission electron microscopy (TEM). The emulsions prepared with the optimized formula demonstrated good physical stability during storage at room temperature. Antibacterial activity was conducted with the optimized nano emulsion NESH 01 and compared with free clarithromycin. Zone of inhibition was larger for NESH 01 as compared to that with free clarithromycin. This implies that the solubility and hence the bioavailability of clarithromycin has increased in the formulated nano emulsion system.

  14. Citizen Contributions to the Closure of High-Level Waste (HLW) Tanks 18 and 19 at the Department of Energy's (DOE) Savannah River Site (SRS) - 13448

    Energy Technology Data Exchange (ETDEWEB)

    Lawless, W.F. [Paine College, Departments of Math and Psychology, 1235 15th Street, Augusta, GA 30901 (United States)

    2013-07-01

    Citizen involvement in DOE's decision-making for the environmental cleanup from DOE's management of its nuclear wastes across the DOE complex has had a positive effect on the cleanup of its SRS site, characterized by an acceleration of cleanup not only for the Transuranic wastes at SRS, but also for DOE's first two closures of HLW tanks, both of which occurred at SRS. The Citizens around SRS had pushed successfully for the closures of Tanks 17 and 20 in 1997, becoming the first closures of HLW tanks under regulatory guidance in the USA. However, since then, HLW tank closures ceased due to a lawsuit, the application of new tank clean-up technology, interagency squabbling between DOE and NRC over tank closure criteria, and finally and almost fatally, from budget pressures. Despite an agreement with its regulators for the closure of Tanks 18 and 19 by the end of calendar year 2012, the outlook in Fall 2011 to close these two tanks had dimmed. It was at this point that the citizens around SRS became reengaged with tank closures, helping DOE to reach its agreed upon milestone. (authors)

  15. Experience gained with the Synroc demonstration plant at ANSTO and its relevance to plutonium immobilization

    Energy Technology Data Exchange (ETDEWEB)

    Jostsons, A.; Ridal, A.; Mercer, D.J.; Vance, E.R.L. [Australian Nuclear Science and Technology Organisation, Menai (Australia)

    1996-05-01

    The Synroc Demonstration Plant (SDP) was designed and constructed at Lucas Heights to demonstrate the feasibility of Synroc production on a commercial scale (10 kg/hr) with simulated Purex liquid HLW. Since commissioning of the SDP in 1987, over 6000 kg of Synroc has been fabricated with a range of feeds and waste loadings. The SDP utilises uniaxial hot-pressing to consolidate Synroc. Pressureless sintering and hot-isostatic pressing have also been studied at smaller scales. The results of this extensive process development have been incorporated in a conceptual design for a radioactive plant to condition HLW from a reprocessing plant with a capacity to treat 800 tpa of spent LWR fuel. Synroic containing TRU, including Pu, and fission products has been fabricated and characterised in a glove-box facility and hot cells, respectively. The extensive experience in processing of Synroc over the past 15 years is summarised and its relevance to immobilization of surplus plutonium is discussed.

  16. Synthesis of acid-soluble spore proteins by Bacillus subtilis.

    OpenAIRE

    Leventhal, J M; Chambliss, G H

    1982-01-01

    The major acid-soluble spore proteins (ASSPs) of Bacillus subtilis were detected by immunoprecipitation of radioactively labeled in vitro- and in vivo-synthesized proteins. ASSP synthesis in vivo began 2 h after the initiation of sporulation (t2) and reached its maximum rate at t7. This corresponded to the time of synthesis of mRNA that stimulated the maximum rate of ASSP synthesis in vitro. Under the set of conditions used in these experiments, protease synthesis began near t0, alkaline phos...

  17. Solubility study of Tc(IV) oxides

    International Nuclear Information System (INIS)

    Liu, D.J.; Fan, X.H.

    2005-01-01

    The deep geological disposal of the high level radioactive wastes is expected to be a safer disposal method in most countries. The long-lived fission product 99 Tc is present in large quantities in nuclear wastes and its chemical behavior in aqueous solution is of considerable interest. Under oxidizing conditions technetium exists as the anionic species TcO 4 - whereas under the reducing conditions, expected to exist in a deep geological repository, it is generally predicted that technetium will be present as TcO 2 ·nH 2 O. Hence, the mobility of Tc(IV) in reducing groundwater may be limited by the solubility of TcO 2 ·nH 2 O under these conditions. Due to this fact it is important to investigate the solubility of TcO 2 ·nH 2 O. The solubility determines the release of radionuclides from waste form and is used as a source term in radionuclide migration analysis in performance assessment of radioactive waste repository. Technetium oxide was prepared by reduction of a technetate solution with Sn 2 + . The solubility of Tc(IV) oxide has been determined in simulated groundwater and redistilled water under aerobic and anaerobic conditions. The effects of pH and CO 3 2- concentration of solution on solubility of Tc(IV) oxide were studied. The concentration of total technetium and Tc(IV) species in the solutions were periodically determined by separating the oxidized and reduced technetium species using a solvent extraction procedure and counting the beta activity of the 99 Tc with a liquid scintillation counter. The experimental results show that the rate of oxidation of Tc(IV) in simulated groundwater and redistilled water is about (1.49-1.86) x 10 -9 mol/(L·d) under aerobic conditions, but Tc(IV) in simulated groundwater and redistilled water is not oxidized under anaerobic conditions. Under aerobic or anaerobic conditions the solubility of Tc(IV) oxide in simulated groundwater and redistilled water is equal on the whole after centrifugation or ultrafiltration. The

  18. Influence of milling process on efavirenz solubility

    Directory of Open Access Journals (Sweden)

    Erizal Zaini

    2017-01-01

    Full Text Available Introduction: The aim of this study was to investigate the influence of the milling process on the solubility of efavirenz. Materials and Methods: Milling process was done using Nanomilling for 30, 60, and 180 min. Intact and milled efavirenz were characterized by powder X-ray diffraction, scanning electron microscopy (SEM, spectroscopy infrared (IR, differential scanning calorimetry (DSC, and solubility test. Results: The X-ray diffractogram showed a decline on peak intensity of milled efavirenz compared to intact efavirenz. The SEM graph depicted the change from crystalline to amorphous habit after milling process. The IR spectrum showed there was no difference between intact and milled efavirenz. Thermal analysis which performed by DSC showed a reduction on endothermic peak after milling process which related to decreasing of crystallinity. Solubility test of intact and milled efavirenz was conducted in distilled water free CO2with 0.25% sodium lauryl sulfate media and measured using high-performance liquid chromatography method with acetonitrile: distilled water (80:20 as mobile phases. The solubility was significantly increased (P < 0.05 after milling processes, which the intact efavirenz was 27.12 ± 2.05, while the milled efavirenz for 30, 60, and 180 min were 75.53 ± 1.59, 82.34 ± 1.23, and 104.75 ± 0.96 μg/mL, respectively. Conclusions: Based on the results, the solubility of efavirenz improved after milling process.

  19. Solubility of lithium deuteride in liquid lithium

    International Nuclear Information System (INIS)

    Veleckis, E.; Yonco, R.M.; Maroni, V.A.

    1977-01-01

    The solubility of LiD in liquid lithium between the eutectic and monotectic temperatures was measured using a direct sampling method. Solubilities were found to range from 0.0154 mol.% LiD at 199 0 C to 3.32 mol.% LiD at 498 0 C. The data were used in the derivation of an expression for the activity coefficient of LiD as a function of temperature and composition and an equation relating deuteride solubility and temperature, thus defining the liquidus curve. Similar equations were also derived for the Li-LiH system using the existing solubility data. Extrapolation of the liquidus curves yielded the eutectic concentrations (0.040 mol.% LiH and 0.035 mol.% LiD) and the freezing point depressions (0.23 0 C for Li-LiH and 0.20 0 C for Li-LiD) at the eutectic point. The results are compared with the literature data for hydrogen and deuterium. The implications of the relatively high solubility of hydrogen isotopes in lithium just above the melting point are discussed with respect to the cold trapping of tritium in fusion reactor blankets. (Auth.)

  20. Solubility of pllutonium in alkaline salt solutions

    International Nuclear Information System (INIS)

    Hobbs, D.T.; Edwards, T.B.

    1993-01-01

    Plutonium solubility data from several studies have been evaluated. For each data set, a predictive model has been developed where appropriate. In addition, a statistical model and corresponding prediction intervals for plutonium solubility as a quadratic function of the hydroxide concentration have been developed. Because of the wide range of solution compositions, the solubility of plutonium can vary by as much as three orders of magnitude for any given hydroxide concentration and still remain within the prediction interval. Any nuclear safety assessments that depend on the maximum amount of plutonium dissolved in alkaline salt solutions should use concentrations at least as great as the upper prediction limits developed in this study. To increase the confidence in the prediction model, it is recommended that additional solubility tests be conducted at low hydroxide concentrations and with all of the other solution components involved. To validate the model for application to actual waste solutions, it is recommended that the plutonium solubilities in actual waste solutions be determined and compared to the values predicted by the quadratic model

  1. 40 CFR Table 7 to Subpart Vvvvvv... - Partially Soluble HAP

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 14 2010-07-01 2010-07-01 false Partially Soluble HAP 7 Table 7 to... Pt. 63, Subpt. VVVVVV, Table 7 Table 7 to Subpart VVVVVV of Part 63—Partially Soluble HAP As required... partially soluble HAP listed in the following table. Partially soluble HAP name CAS No. 1. 1,1,1...

  2. Interlaboratory validation of small-scale solubility and dissolution measurements of poorly water-soluble drugs

    DEFF Research Database (Denmark)

    Andersson, Sara B. E.; Alvebratt, Caroline; Bevernage, Jan

    2016-01-01

    The purpose of this study was to investigate the interlaboratory variability in determination of apparent solubility (Sapp) and intrinsic dissolution rate (IDR) using a miniaturized dissolution instrument. Three poorly water-soluble compounds were selected as reference compounds and measured at m...

  3. Effect of cyclodextrin complexation on the aqueous solubility and solubility/dose ratio of praziquantel.

    Science.gov (United States)

    Maragos, Stratos; Archontaki, Helen; Macheras, Panos; Valsami, Georgia

    2009-01-01

    Praziquantel (PZQ), the primary drug of choice in the treatment of schistosomiasis, is a highly lipophilic drug that possesses high permeability and low aqueous solubility and is, therefore, classified as a Class II drug according to the Biopharmaceutics Classification System (BCS). In this work, beta-cyclodextrin (beta-CD) and hydroxypropyl-beta-cyclodextrin (HP-beta-CD) were used in order to determine whether increasing the aqueous solubility of a drug by complexation with CDs, a BCS-Class II compound like PZQ could behave as BCS-Class I (highly soluble/highly permeable) drug. Phase solubility and the kneading and lyophilization techniques were used for inclusion complex preparation; solubility was determined by UV spectroscopy. The ability of the water soluble polymer polyvinylpyrolidone (PVP) to increase the complexation and solubilization efficiency of beta-CD and HP-beta-CD for PZQ was examined. Results showed significant improvement of PZQ solubility in the presence of both cyclodextrins but no additional effect in the presence of PVP. The solubility/dose ratios values of PZQ-cyclodextrin complexes calculated considering the low (150 mg) and the high dose (600 mg) of PZQ, used in practice, indicate that PZQ complexation with CDs may result in drug dosage forms that would behave as a BCS-Class I depending on the administered dose.

  4. Evaluation on changes caused by volcanic activities in the groundwater environment as a natural barrier for the HLW disposal. Literature survey and groundwater observation conducted at Mt. Iwate

    International Nuclear Information System (INIS)

    Mahara, Yasunori; Nakata, Eiji; Tanaka, Kazuhiro

    2000-01-01

    It is very important in the site characterization for the HLW disposal to understand changes in geochemical performances caused by volcanic activities in the groundwater environment as the natural barrier. The various effects and its magnitude of changes were listed up and were filed from literature surveys of the correlation between volcanic activities and hydrological can geochemical changes (e.g. water temperature, water pressure, water level, dissolved gas concentration of He and Rn, isotopic ratio of He, and chloride concentration) in volcanic aquifer. However, it is difficult to evaluate the magnitude of impacts, which volcanic activities will give to the groundwater environment in the natural barrier, through only the literature surveys. We have started monitoring of groundwater level and changes in groundwater quality, since volcanic activities have enhanced at Mt. Iwate from June in 1998. Judging from variation of isotopic ratio of dissolved He in groundwater, a prompt and sharp signals indicating volcanic activities will easily be found in shallow groundwater and discharged ponds. On the other hands, geochemical conditions in deep groundwater surroundings from some 100 m to 1000 m deep will be very stable, if the area being more than 5 km apart from the volcanic active center. Consequently, our observed results suggest that the groundwater environment which is not directly disturbed by the underground magmatic activities spreads under the area that is connected to trench side of the volcanic front. (author)

  5. Verification study on technology for preliminary investigation for HLW geological disposal. Part 2. Verification of surface geophysical prospecting through establishing site descriptive models

    International Nuclear Information System (INIS)

    Kondo, Hirofumi; Suzuki, Koichi; Hasegawa, Takuma; Goto, Keiichiro; Yoshimura, Kimitaka; Muramoto, Shigenori

    2012-01-01

    The Yokosuka demonstration and validation project using Yokosuka CRIEPI site has been conducted since FY 2006 as a cooperative research between NUMO (Nuclear Waste Management Organization of Japan) and CRIEPI. The objectives of this project are to examine and to refine the basic methodology of the investigation and assessment of properties of geological environment in the stage of Preliminary Investigation for HLW geological disposal. Within Preliminary Investigation technologies, surface geophysical prospecting is an important means of obtaining information from deep geological environment for planning borehole surveys. In FY 2010, both seismic prospecting (seismic reflection and vertical seismic profiling methods) for obtaining information about geological structure and electromagnetic prospecting (magneto-telluric and time domain electromagnetic methods) for obtaining information about resistivity structure reflecting the distribution of salt water/fresh water boundary to a depth of over several hundred meters were conducted in the Yokosuka CRIEPI site. Through these surveys, the contribution of geophysical prospecting methods in the surface survey stage to improving the reliability of site descriptive models was confirmed. (author)

  6. Coupling diffusion and high-pH precipitation/dissolution in the near field of a HLW repository in clay by means of reactive solute transport models

    Science.gov (United States)

    Samper, J.; Font, I.; Yang, C.; Montenegro, L.

    2004-12-01

    The reference concept for a HLW repository in clay in Spain includes a 75 cm thick bentonite buffer which surrounds canisters. A concrete sustainment 20 cm thick is foreseen between the bentonite buffer and the clay formation. The long term geochemical evolution of the near field is affected by a high-pH hyperalkaline plume induced by concrete. Numerical models of multicomponent reactive transport have been developped in order to quantify the evolution of the system over 1 Ma. Water flow is negligible once the bentonite buffer is saturated after about 20 years. Therefore, solute transport occurs mainly by diffusion. Models account for aqueous complexation, acid-base and redox reactions, cation exchange, and mineral dissolution precipitation in the bentonite, the concrete and the clay formation. Numerical results obtained witth CORE2D indicate that the high-pH plume causes significant changes in porewater chemistry both in the bentonite buffer and the clay formation. Porosity changes caused by mineral dissolution/precipitation are extremely important. Therefore, coupled modes of diffusion and reactive transport accounting for changes in porosity caused by mineral precipitation are required in order to obtain realistic predictions.

  7. Reactive transport modelling of a heating and radiation experiment in the Boom clay (Belgium)

    International Nuclear Information System (INIS)

    Montenegro, L.; Samper, J.; Delgado, J.

    2003-01-01

    Most countries around the world consider Deep Geological Repositories (DGR) as the most safe option for the final disposal of high level radioactive waste (HLW). DGR is based on adopting a system of multiple barriers between the HLW and the biosphere. Underground laboratories provide information about the behaviour of these barriers at real conditions. Here we present a reactive transport model for the CERBERUS experiment performed at the HADES underground laboratory at Mol (Belgium) in order to characterize the thermal (T), hydrodynamic (H) and geochemical (G) behaviour of the Boon clay. This experiment is unique because it addresses the combined effect of heat and radiation produced by the storage of HLW in a DGR. Reactive transport models which are solved with CORE, are used to perform quantitative predictions of Boom clay thermo-hydro-geochemical (THG) behaviour. Numerical results indicate that heat and radiation cause a slight oxidation near of the radioactive source, pyrite dissolution, a pH decrease and slight changes in the pore water chemical composition of the Boom clay. (Author) 33 refs

  8. In-situ experiments for the determination of rock properties and behaviour at the Meuse/Haute Marne Centre

    Directory of Open Access Journals (Sweden)

    Conil N.

    2010-06-01

    Full Text Available Andra is in charge of studying the feasibility of a disposal facility for longlived high-level nuclear waste (LL-HLW in a deep geological environment. With this aim, dedicated experiments have been carried out for several years at the Meuse/Haute Marne Underground Research Laboratory excavated in a 500 m deep argillaceous rock formation. These experiments include determining the feasibility of the excavation of disposal cells for LL-HLW, consisting of 40 meter long, 70 cm in diameter, horizontal cased micro tunnels. The hydro mechanical impact of the excavation of such openings on the rock mass behaviour is continuously monitored as well as their mean term mechanical behaviour. Since LL-HLW produce heat, the impact of temperature on the surrounding rock mass and on the micro tunnel steel casing will also be studied. Specific instrumentation has been developed to study this impact. The first step of the microtunnel excavation tests, carried out in 2009, has led to improving the excavation method and the drilling machine. These improvements will be tested in the next step of the excavation tests planned for 2010. The THM experiment dedicated to studying the behaviour of the rock mass under thermal solicitation started early 2010. The behaviour of a steel casing in contact with the rock mass and under thermal solicitation will be studied in an experiment scheduled to start in September 2010.

  9. Solubilities of boric acid in heavy water

    International Nuclear Information System (INIS)

    Nakai, Shigetsugu; Aoi, Hideki; Hayashi, Ken-ichi; Katoh, Taizo; Watanabe, Takashi.

    1988-01-01

    A gravimetric analysis using meta-boric acid (HBO 2 or DBO 2 ) as a weighing form has been developed for solubility measurement. The method gave satisfactory results in preliminary measurement of solubilities of boric acid in light water. By using this method, the solubilities of 10 B enriched D 3 BO 3 in heavy water were measured. The results are as follows; 2.67 (7deg C), 3.52 (15deg C), 5.70 (30deg C), 8.87 (50deg C) and 12.92 (70deg C) w/o, respectively. These values are about 10% lower than those in light water. Thermodynamical consideration based on the data shows that boric acid is the water structure breaker. (author)

  10. Resveratrol cocrystals with enhanced solubility and tabletability.

    Science.gov (United States)

    Zhou, Zhengzheng; Li, Wanying; Sun, Wei-Jhe; Lu, Tongbu; Tong, Henry H Y; Sun, Changquan Calvin; Zheng, Ying

    2016-07-25

    Two new 1:1 cocrystals of resveratrol (RES) with 4-aminobenzamide (RES-4ABZ) and isoniazid (RES-ISN) were synthesized by liquid assisted grinding (LAG) and rapid solvent removal (RSR) methods using ethanol as solvent. Their physiochemical properties were characterized using PXRD, DSC, solid state and solution NMR, FT-IR, and HPLC. Pharmaceutically relevant properties, including tabletability, solubility, intrinsic dissolution rate, and hygroscopicity, were evaluated. Temperature-composition phase diagram for RES-ISN cocrystal system was constructed from DSC data. Both cocrystals show higher solubility than resveratrol over a broad range of pH. They are phase stable and non-hygroscopic even under high humidity conditions. Importantly, both cocrystals exhibit improved solubility and tabletability compared with RES, which make them more suitable candidates for tablet formulation development. Copyright © 2016 Elsevier B.V. All rights reserved.

  11. AW-101 entrained solids - Solubility versus temperature

    International Nuclear Information System (INIS)

    GJ Lumetta; RC Lettau; GF Piepel

    2000-01-01

    This report describes the results of a test conducted by Battelle to assess the solubility of the solids entrained in the diluted AW-101 low-activity waste (LAW) sample. BNFL requested Battelle to dilute the AW-1-1 sample using de-ionized water to mimic expected plant operating conditions. BNFL further requested Battelle to assess the solubility of the solids present in the diluted AW-101 sample versus temperature conditions of 30, 40, and 50 C. BNFL requested these tests to assess the composition of the LAW supernatant and solids versus expected plant-operating conditions. The work was conducted according to test plan BNFL-TP-29953-7, Rev. 0, Determination of the Solubility of LAW Entrained Solids. The test went according to plan, with no deviations from the test plan

  12. Solubility and stability of inorganic carbonates

    International Nuclear Information System (INIS)

    Taylor, P.

    1987-01-01

    The chemistry of inorganic carbonates is reviewed, with emphasis on solubility and hydrolytic stability, in order to identify candidate waste forms for immobilization and disposal of 14 C. At present, CaCO 3 and BaCO 3 are the two most widely favoured wasted forms, primarily because they are the products of proven CO 2 -scrubbing technology. However, they have relatively high solubilities in non-alkaline solutions, necessitating care in selecting and assessing an appropriate disposal environment. Three compounds with better solubility characteristics in near-neutral waters are identified: bismutite, (BiO) 2 CO 3 ; hydrocerussite, Pb 3 (OH) 2 (CO 3 ) 2 ; and rhodochrosite, MnCO 3 . Some of the limitations of each of these alternative waste forms are discussed

  13. A framework for API solubility modelling

    DEFF Research Database (Denmark)

    Conte, Elisa; Gani, Rafiqul; Crafts, Peter

    . In addition, most of the models are not predictive and requires experimental data for the calculation of the needed parameters. This work aims at developing an efficient framework for the solubility modelling of Active Pharmaceutical Ingredients (API) in water and organic solvents. With this framework......-SAFT) are used for solubility calculations when the needed interaction parameters or experimental data are available. The CI-UNIFAC is instead used when the previous models lack interaction parameters or when solubility data are not available. A new GC+ model for APIs solvent selection based...... on the hydrophobicity, hydrophilicity and polarity information of the API and solvent is also developed, for performing fast solvent selection and screening. Eventually, all the previous developments are integrated in a framework for their efficient and integrated use. Two case studies are presented: the first...

  14. Solubility of iron in liquid lead

    International Nuclear Information System (INIS)

    Ali-Khan, I.

    1981-01-01

    The use of liquid lead in high temperature chemical and metallurgical processes is well known. The structural materials applied for the containment of these processes are either iron base alloys or possess iron as an alloying element. Besides that, lead itself is alloyed in some steels to achieve some very useful properties. For understanding the effect of liquid lead in such structural materials, it is important to determine the solubility of iron in liquid lead which would also be indicative of the stability of these alloys. At the institute of reactor materials of KFA Juelich, investigations have been conducted to determine the solubility of iron in liquid lead up to a temperature of about 1000 0 C. In this presentation the data concerning the solubility of iron in liquid lead are brought up to date and discussed including the results of our previous investigations. (orig.)

  15. Equilibrium Solubility of CO2 in Alkanolamines

    DEFF Research Database (Denmark)

    Waseem Arshad, Muhammad; Fosbøl, Philip Loldrup; von Solms, Nicolas

    2014-01-01

    Equilibrium solubility of CO2 were measured in aqueous solutions of Monoethanolamine (MEA) and N,N-diethylethanolamine(DEEA). Equilibrium cells are generally used for these measurements. In this study, the equilibrium data were measured from the calorimetry. For this purpose a reaction calorimeter...... (model CPA 122 from ChemiSens AB, Sweden) was used. The advantage of this method is being the measurement of both heats of absorption and equilibrium solubility data of CO2 at the same time. The measurements were performed for 30 mass % MEA and 5M DEEA solutions as a function of CO2 loading at three...... different temperatures 40, 80 and 120 ºC. The measured 30 mass % MEA and 5M DEEA data were compared with the literature data obtained from different equilibrium cells which validated the use of calorimeters for equilibrium solubility measurements....

  16. FINAL REPORT INTEGRATED DM1200 MELTER TESTING OF REDOX EFFECTS USING HLW AZ-101 AND C-106/AY-102 SIMULANTS VSL-04R4800-1 REV 0 5/6/

    Energy Technology Data Exchange (ETDEWEB)

    KRUGER AA; MATLACK KS; GONG W; BARDAKCI T; D' ANGELO NA; LUTZE W; BIZOT PM; CALLOW RA; BRANDYS M; KOT WK; PEGG IL

    2011-12-29

    This report documents melter and off-gas performance results obtained on the DM1200 HLW Pilot Melter during processing of AZ-101 and C-106/AY-102 HLW simulants. The tests reported herein are a subset of three tests from a larger series of tests described in the Test Plan for the work; results from the remaining tests will be reported separately. Three nine day tests, one with AZ-101 and two with C-106/AY-102 feeds were conducted with variable amounts of added sugar to address the effects of redox. The test with AZ-101 included ruthenium spikes to also address the effects of redox on ruthenium volatility. One of tests addressed the effects of increased flow-sheet nitrate levels using C-106/AY-102 feeds. With high nitrate/nitrite feeds (such as WTP LAW feeds), reductants are required to prevent melt foaming and deleterious effects on glass production rates. Sugar is the baseline WTP reductant for this purpose. WTP HLW feeds typically have relatively low nitrate/nitrite content in comparison to the organic carbon content and, therefore, have typically not required sugar additions. However, HLW feed variability, particularly with respect to nitrate levels, may necessitate the use of sugar in some instances. The tests reported here investigate the effects of variable sugar additions to the melter feed as well as elevated nitrate levels in the waste. Variables held constant to the extent possible included melt temperature, bubbling rate, plenum temperature, cold cap coverage, the waste simulant composition, and the target glass composition. The principal objectives of the DM1200 melter testing were to determine the achievable glass production rates for simulated HLW feeds with variable amounts of added sugar and increased nitrate levels; characterize melter off-gas emissions; characterize the performance of the prototypical off-gas system components as well as their integrated performance; characterize the feed, glass product, and off-gas effluents; and perform pre- and

  17. Final Report Integrated DM1200 Melter Testing Of Redox Effects Using HLW AZ-101 And C-106/AY-102 Simulants VSL-04R4800-1, Rev. 0, 5/6/04

    International Nuclear Information System (INIS)

    Kruger, A.A.; Matlack, K.S.; Gong, W.; Bardakci, T.; D'Angelo, N.A.; Lutze, W.; Bizot, P.M.; Callow, R.A.; Brandys, M.; Kot, W.K.; Pegg, I.L.

    2011-01-01

    This report documents melter and off-gas performance results obtained on the DM1200 HLW Pilot Melter during processing of AZ-101 and C-106/AY-102 HLW simulants. The tests reported herein are a subset of three tests from a larger series of tests described in the Test Plan for the work; results from the remaining tests will be reported separately. Three nine day tests, one with AZ-101 and two with C-106/AY-102 feeds were conducted with variable amounts of added sugar to address the effects of redox. The test with AZ-101 included ruthenium spikes to also address the effects of redox on ruthenium volatility. One of tests addressed the effects of increased flow-sheet nitrate levels using C-106/AY-102 feeds. With high nitrate/nitrite feeds (such as WTP LAW feeds), reductants are required to prevent melt foaming and deleterious effects on glass production rates. Sugar is the baseline WTP reductant for this purpose. WTP HLW feeds typically have relatively low nitrate/nitrite content in comparison to the organic carbon content and, therefore, have typically not required sugar additions. However, HLW feed variability, particularly with respect to nitrate levels, may necessitate the use of sugar in some instances. The tests reported here investigate the effects of variable sugar additions to the melter feed as well as elevated nitrate levels in the waste. Variables held constant to the extent possible included melt temperature, bubbling rate, plenum temperature, cold cap coverage, the waste simulant composition, and the target glass composition. The principal objectives of the DM1200 melter testing were to determine the achievable glass production rates for simulated HLW feeds with variable amounts of added sugar and increased nitrate levels; characterize melter off-gas emissions; characterize the performance of the prototypical off-gas system components as well as their integrated performance; characterize the feed, glass product, and off-gas effluents; and perform pre- and

  18. Contribution of the European Commission to a European Strategy for HLW Management through Partitioning and Transmutation: Presentation of MYRRHA and its Role in the European P and T Strategy

    International Nuclear Information System (INIS)

    Abderrahim, H.A.; Van den Eynde, G.; Baeten, P.; Schyns, M.; Vandeplassche, D.; Kochetkov, A.

    2015-01-01

    To be able to answer the world's increasing demand for energy, nuclear energy must be part of the energy mix. As a consequence of the nuclear electricity generation, high-level nuclear waste (HLW) is produced. The HLW is presently considered to be managed through its burying in geological storage. Partitioning and transmutation (P and T) has been pointed out as the strategy to reduce the radiological impact of HLW. Transmutation can be achieved in an efficient way in fast neutron spectrum facilities, both in critical fast reactors as well as in accelerator driven systems (ADSs). For more than two decades, the European Commission has been co-funding various research and development projects conducted in many European research organisations and industries related to P and T as a complementary strategy for high-level waste management to the geological disposal. In 2005, a European strategy for the implementation of P and T for a large part of the HLW in Europe indicated the need for the demonstration of its feasibility at an 'engineering' level. The R and D activities of this strategy were arranged in four 'building blocks': 1. Demonstration of the capability to process a sizable amount of spent fuel from commercial light water reactors (LWRs) in order to separate plutonium, uranium and minor actinides. 2. Demonstration of the capability to fabricate at a semi-industrial level the dedicated fuel needed as load in a dedicated transmuter. 3. Design and construction of one or more dedicated transmuters. 4. Provision of a specific installation for processing of the dedicated fuel unloaded from the transmuter, which can be of a different type than the one used to process the original spent fuel unloaded from the commercial power plants, together with the fabrication of new dedicated fuel. MYRRHA contributes to the third building block. MYRRHA is an ADS under development at SCK.CEN in collaboration with a large number of European partners. One of

  19. Respiratory carcinogenicity assessment of soluble nickel compounds.

    Science.gov (United States)

    Oller, Adriana R

    2002-10-01

    The many chemical forms of nickel differ in physicochemical properties and biological effects. Health assessments for each main category of nickel species are needed. The carcinogenicity assessment of water-soluble nickel compounds has proven particularly difficult. Epidemiologic evidence indicates an association between inhalation exposures to nickel refinery dust containing soluble nickel compounds and increased risk of respiratory cancers. However, the nature of this association is unclear because of limitations of the exposure data, inconsistent results across cohorts, and the presence of mixed exposures to water-insoluble nickel compounds and other confounders that are known or suspected carcinogens. Moreover, well-conducted animal inhalation studies, where exposures were solely to soluble nickel, failed to demonstrate a carcinogenic potential. Similar negative results were seen in animal oral studies. A model exists that relates respiratory carcinogenic potential to the bioavailability of nickel ion at nuclear sites within respiratory target cells. This model helps reconcile human, animal, and mechanistic data for soluble nickel compounds. For inhalation exposures, the predicted lack of bioavailability of nickel ion at target sites suggests that water-soluble nickel compounds, by themselves, will not be complete human carcinogens. However, if inhaled at concentrations high enough to induce chronic lung inflammation, these compounds may enhance carcinogenic risks associated with inhalation exposure to other substances. Overall, the weight of evidence indicates that inhalation exposure to soluble nickel alone will not cause cancer; moreover, if exposures are kept below levels that cause chronic respiratory toxicity, any possible tumor-enhancing effects (particularly in smokers) would be avoided.

  20. SITE-94. Radionuclide solubilities for SITE-94

    Energy Technology Data Exchange (ETDEWEB)

    Arthur, R.; Apted, M. [QuantiSci, Denver, CO (United States)

    1996-12-01

    In this report, solubility constraints are evaluated on radioelement source-term concentrations supporting the SITE-94 performance assessment. Solubility models are based on heterogeneous-equilibrium, mass- and charge-balance constraints incorporated into the EQ3/6 geochemical software package, which is used to calculate the aqueous speciation behavior and solubilities of U, Th, Pu, Np, Am, Ni, Ra, Se, Sn, Sr, Tc and Zr in site groundwaters and near-field solutions. The chemical evolution of the near field is approximated using EQ3/6 in terms of limiting conditions at equilibrium, or steady state, in three closed systems representing fully saturated bentonite, Fe{sup o} corrosion products of the canister, and spent fuel. The calculations consider both low-temperature (15 deg C) and high-temperature (80 deg C) conditions in the near field, and the existence of either reducing or strongly oxidizing conditions in each of the bentonite, canister, and spent-fuel barriers. Heterogeneities in site characteristics are evaluated through consideration of a range of initial groundwaters and their interactions with engineered barriers. Aqueous speciation models for many radioelements are constrained by thermodynamic data that are estimated with varying degrees of accuracy. An important question, however, is how accurate do these models need to be for purposes of estimating source-term concentrations? For example, it is unrealistic to expect a high degree of accuracy in speciation models if such models predict solubilities that are below the analytical detection limit for a given radioelement. From a practical standpoint, such models are irrelevant if calculated solubilities cannot be tested by direct comparison to experimental data. In the absence of models that are both accurate and relevant for conditions of interest, the detection limit could define a pragmatic upper limit on radioelement solubility 56 refs, 25 tabs, 10 figs

  1. SITE-94. Radionuclide solubilities for SITE-94

    International Nuclear Information System (INIS)

    Arthur, R.; Apted, M.

    1996-12-01

    In this report, solubility constraints are evaluated on radioelement source-term concentrations supporting the SITE-94 performance assessment. Solubility models are based on heterogeneous-equilibrium, mass- and charge-balance constraints incorporated into the EQ3/6 geochemical software package, which is used to calculate the aqueous speciation behavior and solubilities of U, Th, Pu, Np, Am, Ni, Ra, Se, Sn, Sr, Tc and Zr in site groundwaters and near-field solutions. The chemical evolution of the near field is approximated using EQ3/6 in terms of limiting conditions at equilibrium, or steady state, in three closed systems representing fully saturated bentonite, Fe o corrosion products of the canister, and spent fuel. The calculations consider both low-temperature (15 deg C) and high-temperature (80 deg C) conditions in the near field, and the existence of either reducing or strongly oxidizing conditions in each of the bentonite, canister, and spent-fuel barriers. Heterogeneities in site characteristics are evaluated through consideration of a range of initial groundwaters and their interactions with engineered barriers. Aqueous speciation models for many radioelements are constrained by thermodynamic data that are estimated with varying degrees of accuracy. An important question, however, is how accurate do these models need to be for purposes of estimating source-term concentrations? For example, it is unrealistic to expect a high degree of accuracy in speciation models if such models predict solubilities that are below the analytical detection limit for a given radioelement. From a practical standpoint, such models are irrelevant if calculated solubilities cannot be tested by direct comparison to experimental data. In the absence of models that are both accurate and relevant for conditions of interest, the detection limit could define a pragmatic upper limit on radioelement solubility

  2. Hydrogen solubility in austenite of Fe-Ni-Cr alloys

    International Nuclear Information System (INIS)

    Zhirnova, V.V.; Mogutnov, B.M.; Tomilin, I.A.

    1981-01-01

    Hydrogen solubility in Fe-Ni-Cr alloys at 600-1000 deg C is determined. Hydrogen solubility in ternary alloys can not be predicted on the basis of the data on its solubility in binary Fe-Ni, Fe-Cr alloys. Chromium and nickel effect on hydrogen solubility in iron is insignificant in comparison with the effect of these elements on carbon or nitrogen solubility [ru

  3. Hydrogen terminal solubility in Zircaloy-4

    International Nuclear Information System (INIS)

    Vizcaino, Pablo; Banchik, Abrahan D.

    1999-01-01

    Terminal solubility temperature of hydrogen in zirconium and its alloys is an important parameter because hydrides precipitation embrittled these materials making them susceptible to the phenomenon known as retarded hydrogen cracking. This work continues the study presented in the 25 AATN Meeting. Within this framework, a study focused on determining these curves in recrystallized Zircaloy-4, using scanning differential calorimetric technique. Terminal solubility curves for Zircaloy-4 were constructed within a concentration range from 40 to 640 ppm in hydrogen weight and comparisons with results obtained by other authors were made. (author)

  4. Nitrogen solubility in nickel base multicomponent melts

    International Nuclear Information System (INIS)

    Bol'shov, L.A.; Stomakhin, A.Ya.; Sokolov, V.M.; Teterin, V.G.

    1984-01-01

    Applicability of various methods for calculation of nitrogen solubility in high-alloyed nickel base alloys, containing Cr, Fe, W, Mo, Ti, Nb, has been estimated. A possibility is shown to use the formUla, derived for the calculation of nitrogen solubility in iron on the basis of statistical theory for a grid model of solution which does not require limitations for the content of a solvent component. The calculation method has been used for nickel alloys, with the concentration of solvent, iron, being accepted equal to zero, and employing parameters of nitrogen interaction as determined for iron-base alloys

  5. Effect of amides on lithium tetraborate solubility

    Energy Technology Data Exchange (ETDEWEB)

    Tsekhanskij, R S; Skvortsov, V C; Molodkin, A K; Sadetdi-pov, Sh V [Chuvashskij Gosudarstvennyj Pedagogicheskij Inst., Cheboksary (USSR); Universitet Druzhby Narodov, Moscow (USSR))

    1983-03-01

    Using the methods of solubility, densi- and refractometry at 25 deg C, it has been established that the systems lithium tetraborate-formamide (acetamide, dimethyl-formamide)-water are of a simple eutonic type. Amides decrease the salt solubility. Lyotropic effect, as calculated for molar concentrations (-Lsub(M)) relative to the absolute value, increases from formamide to dimethyl-formamide. The sequence is determined by the fact that, when there is one or two hydrophilic methyl groups in amide molecules which are in contact with tetraborate, they decrease the hydration energy of lithium cations.

  6. Effect of amides on sodium tetraborate solubility

    International Nuclear Information System (INIS)

    Tsekhanskij, R.S.; Skvortsov, V.G.; Molodkin, A.K.; Sadetdinov, Sh.V.

    1986-01-01

    Methods of solubility and refractometry at 25 deg C were applied to investigate sodium tetraborate - formamide (dimethylformamide) - water systems. It is stated that they are of simple eutonic type as well as the earlier described sodium tetraborate-acetamide-water system. Amides reduce solubility of the salt. The effect of contact interaction between dissolved substances on salt cation hydration and thus on the value of liotropic amide effect is confirmed. This value is found to be also depend on the number of molecules of coordination water in the initial crystalline hydrate

  7. Effect of amides on lithium tetraborate solubility

    International Nuclear Information System (INIS)

    Tsekhanskij, R.S.; Skvortsov, V.C.; Molodkin, A.K.; Sadetdi- pov, Sh.V.

    1983-01-01

    Using the methods of solubility, densi- and refractometry at 25 deg C, it has been established that the systemS lithium tetraborate-formamide (acetamide, dimethyl-formamide)-water are of a simple eutonic type. Amides decrease the salt solubility. Lyotropic effect, as calculated for molar concentrations (-Lsub(M)) relative to the absolute value, increases from formamide to dimethylformamide. The sequence is determined by the fact that, when there is one or two hydrophilic methyl groups in amide molecules which are in contact with tetraborate, they decrease the hydration energy of lithium cations

  8. Effect of amides on sodium tetraborate solubility

    Energy Technology Data Exchange (ETDEWEB)

    Tsekhanskij, R S; Skvortsov, V G; Molodkin, A K; Sadetdinov, Sh V

    1986-11-01

    Methods of solubility and refractometry at 25 deg C were applied to investigate sodium tetraborate - formamide (dimethylformamide) - water systems. It is stated that they are of simple eutonic type as well as the earlier described sodium tetraborate-acetamide-water system. Amides reduce solubility of the salt. The effect of contact interaction between dissolved substances on salt cation hydration and thus on the value of liotropic amide effect is confirmed. This value is found to be also depend on the number of molecules of coordination water in the initial crystalline hydrate.

  9. Modeling of Salt Solubilities in Mixed Solvents

    DEFF Research Database (Denmark)

    Chiavone-Filho, O.; Rasmussen, Peter

    2000-01-01

    A method to correlate and predict salt solubilities in mixed solvents using a UNIQUAC+Debye-Huckel model is developed. The UNIQUAC equation is applied in a form with temperature-dependent parameters. The Debye-Huckel model is extended to mixed solvents by properly evaluating the dielectric...... constants and the liquid densities of the solvent media. To normalize the activity coefficients, the symmetric convention is adopted. Thermochemical properties of the salt are used to estimate the solubility product. It is shown that the proposed procedure can describe with good accuracy a series of salt...

  10. Effect of biosurfactants on the aqueous solubility of PCE and TCE.

    Science.gov (United States)

    Albino, John D; Nambi, Indumathi M

    2009-12-01

    The effect of biosurfactants on the solubility of tetrachloroethylene (PCE) and trichloroethylene (TCE) was studied in batch experiments pertaining to their use for solubilization and mobilization of such contaminants in surfactant enhanced aquifer remediation. Biosurfactants, rhamnolipid and surfactin used in solubility studies were synthesized in our laboratory by Pseudomonas aeruginosa (MTCC 2297) and Bacillus subtilis (MTCC 2423), respectively. The efficiency of the biosurfactants in solubilizing the chlorinated solvents was compared to that of synthetic surfactants. The Weight Solubilization Ratio (WSR) values for solubilization of PCE and TCE by biosurfactants were very high compared to the values obtained for synthetic surfactants. Surfactin proved to be a better surfactant over rhamnolipid. The WSR of surfactin on solubilization of PCE and TCE were 3.83 and 12.5, respectively, whereas the values obtained for rhamnolipid were 2.06 and 8.36. The solubility of the chlorinated solvents by biosurfactants was considerably affected by the changes in pH. The aqueous solubility of PCE and TCE increased tremendously with decrease in pH. The solubility of biosurfactants was observed to decrease with the pH, favoring partitioning of surfactants into the chlorinated solvents in significant amounts at lower pH. The excessive accumulation of biosurfactants at the interface facilitated interfacial tension reductions resulting in higher solubility of the chlorinated solvents at pH less than 7.

  11. Effects of iron(III)chelates on the solubility of heavy metals in calcareous soils

    Energy Technology Data Exchange (ETDEWEB)

    Ylivainio, Kari, E-mail: kari.ylivainio@mtt.f [Department of Applied Chemistry and Microbiology, FIN-00014 University of Helsinki (Finland)

    2010-10-15

    In this study I evaluated the effects of complexing agents on the solubility of heavy metals in an incubation experiment up to 56 days when complexing agents were applied as Fe-chelates (Fe-EDDS(S,S), Fe-EDDS(mix), Fe-EDTA and Fe-EDDHA) on calcareous soils at a level sufficient to correct Fe chlorosis (0.1 mmol kg{sup -1}). Of these ligands, EDDHA was the most efficient in keeping Fe in water-soluble form, and EDDS increased the solubility of Cu and Zn most, and only EDTA increased the solubility of Cd and Pb. EDTA increased the solubility of Ni steadily during the incubation period, equalling about 5-8% of the added EDTA concentration. [S,S]-EDDS was biodegraded within 56 days, whereas EDDS(mix) was less biodegradable. Ni-chelates were the most recalcitrant against biodegradation. The study shows that even a moderate input of chelates to soil increases the solubility of toxic heavy metals and their risk of leaching. - When correcting Fe chlorosis Fe-EDDS causes lower environmental concern than Fe-EDTA.

  12. Effects of iron(III)chelates on the solubility of heavy metals in calcareous soils

    International Nuclear Information System (INIS)

    Ylivainio, Kari

    2010-01-01

    In this study I evaluated the effects of complexing agents on the solubility of heavy metals in an incubation experiment up to 56 days when complexing agents were applied as Fe-chelates (Fe-EDDS(S,S), Fe-EDDS(mix), Fe-EDTA and Fe-EDDHA) on calcareous soils at a level sufficient to correct Fe chlorosis (0.1 mmol kg -1 ). Of these ligands, EDDHA was the most efficient in keeping Fe in water-soluble form, and EDDS increased the solubility of Cu and Zn most, and only EDTA increased the solubility of Cd and Pb. EDTA increased the solubility of Ni steadily during the incubation period, equalling about 5-8% of the added EDTA concentration. [S,S]-EDDS was biodegraded within 56 days, whereas EDDS(mix) was less biodegradable. Ni-chelates were the most recalcitrant against biodegradation. The study shows that even a moderate input of chelates to soil increases the solubility of toxic heavy metals and their risk of leaching. - When correcting Fe chlorosis Fe-EDDS causes lower environmental concern than Fe-EDTA.

  13. Effects of iron(III)chelates on the solubility of heavy metals in calcareous soils.

    Science.gov (United States)

    Ylivainio, Kari

    2010-10-01

    In this study I evaluated the effects of complexing agents on the solubility of heavy metals in an incubation experiment up to 56 days when complexing agents were applied as Fe-chelates (Fe-EDDS(S,S), Fe-EDDS(mix), Fe-EDTA and Fe-EDDHA) on calcareous soils at a level sufficient to correct Fe chlorosis (0.1 mmol kg(-1)). Of these ligands, EDDHA was the most efficient in keeping Fe in water-soluble form, and EDDS increased the solubility of Cu and Zn most, and only EDTA increased the solubility of Cd and Pb. EDTA increased the solubility of Ni steadily during the incubation period, equalling about 5-8% of the added EDTA concentration. [S,S]-EDDS was biodegraded within 56 days, whereas EDDS(mix) was less biodegradable. Ni-chelates were the most recalcitrant against biodegradation. The study shows that even a moderate input of chelates to soil increases the solubility of toxic heavy metals and their risk of leaching. Copyright (c) 2010 Elsevier Ltd. All rights reserved.

  14. Solubility investigation of ether and ester essential oils in water using spectrometry and GC/MS

    Directory of Open Access Journals (Sweden)

    B. Khodabandeloo

    2017-11-01

    Full Text Available Background and objectives: Essential oils (volatiles are aromatic oily liquids prepared from different parts of plants and demonstrate various therapeutic and cosmetic properties. The dissolution of essential oils are not desirable in water, therefore the aim of this research was evaluation and selection the best co-solvents for increasing their solubility and bio availability. Methods:The solubility of six  plants essential oils were investigated in presence of propylene glycol (PG, polyethylene glycol 300 (PEG, glycerin and ethanol as solvent and tween 80 or lecithin as co-solvent by observation and spectrophotometric assay. Chemical composition of the essential oils and supersaturated 50% ethanol (SSE and 50% PG or PEG (SSP solutions were analyzed by GC/MS, too. Results: Ester (Lavandula dentata, Heracleum persicum and, Elettaria cardamomum essential oils showed the best solubility in ethanol and PG, respectively. Ether (Foeniculum vulgare, Pimpinella anisum and Petroselinum crispum essential oils had the best solubility in ethanol and PEG, respectively. In ester class, mixture of ethanol/water was the best solvent according to solubility and total amounts of major compounds of the essential oils. In ether class, all samples had better solubility in mixtures of ethanol/water than PEG, but the amounts of total phenols or ethers in SSP of some samples were higher than SSE. Therefore selecting the best solvent for these class need more experiments. Conclusion: Selecting the solvent for essential oils changes their chemical composition; therefore the best solvent was different for various purposes.

  15. Buckminsterfullerene's (C60) octanol-water partition coefficient (Kow) and aqueous solubility.

    Science.gov (United States)

    Jafvert, Chad T; Kulkarni, Pradnya P

    2008-08-15

    To assess the risk and fate of fullerene C60 in the environment, its water solubility and partition coefficients in various systems are useful. In this study, the log Kow of C60 was measured to be 6.67, and the toluene-water partition coefficient was measured at log Ktw = 8.44. From these values and the respective solubilities of C60 in water-saturated octanol and water-saturated toluene, C60's aqueous solubility was calculated at 7.96 ng/L(1.11 x 10(-11) M) for the organic solvent-saturated aqueous phase. Additionally, the solubility of C60 was measured in mixtures of ethanol-water and tetrahydrofuran-water and modeled with Wohl's equation to confirm the accuracy of the calculated solubility value. Results of a generator column experiment strongly support the hypothesis that clusters form at aqueous concentrations below or near this calculated solubility. The Kow value is compared to those of other hydrophobic organic compounds, and bioconcentration factors for C60 were estimated on the basis of Kow.

  16. Effects of incubation on solubility and mobility of trace metals in two contaminated soils

    International Nuclear Information System (INIS)

    Ma, Lena Q.; Dong Yan

    2004-01-01

    Much research has focused on changes in solubility and mobility of trace metals in soils under incubation. In this experiment, changes in solubility and mobility of trace metals (Pb, Cu and As) and Fe in two contaminated soils from Tampa, Florida and Montreal, Canada were examined. Soils of 30 g were packed in columns and were incubated for 3-80 days under water-flooding incubation. Following incubation, metal concentrations in pore water (water soluble) and in 0.01 M CaCl 2 leachates (exchangeable+water soluble) were determined. While both soils were contaminated with Pb (1600-2500 mg kg -1 ), Tampa soil was also contaminated with As (230 mg kg -1 ). Contrast to the low pH (3.8) of Tampa soil, Montreal soil had an alkaline pH of 7.7 and high Ca of 1.6%. Concentrations of Fe(II) increased with incubation time in the Tampa soil mainly due to reductive Fe dissolution, but decreased in the Montreal soil possibly due to formation of FeCO 3 . The inverse relationship between concentrations of Pb and Fe(II) in pore water coupled with the fact that Fe(II) concentrations were much greater than those of Pb in pore water may suggest the importance of Fe(II) in controlling Pb solubility in soils. However, changes in concentrations of Fe(II), Pb, Cu and As in pore water with incubation time were similar to those in leachate, i.e. water soluble metals were positively related to exchangeable metals in the two contaminated soils. This research suggests the importance of Fe in controlling metal solubility and mobility in soils under water-flooded incubation. - Iron is important in controlling metal solubility and mobility in flooded soils

  17. Effect of fasting on the urinary excretion of water-soluble vitamins in humans and rats.

    Science.gov (United States)

    Fukuwatari, Tsutomu; Yoshida, Erina; Takahashi, Kei; Shibata, Katsumi

    2010-01-01

    Recent studies showed that the urinary excretion of the water-soluble vitamins can be useful as a nutritional index. To determine how fasting affects urinary excretion of water-soluble vitamins, a human study and an animal experiment were conducted. In the human study, the 24-h urinary excretion of water-soluble vitamins in 12 healthy Japanese adults fasting for a day was measured. One-day fasting drastically decreased urinary thiamin content to 30%, and increased urinary riboflavin content by 3-fold. Other water-soluble vitamin contents did not show significant change by fasting. To further investigate the alterations of water-soluble vitamin status by starvation, rats were starved for 3 d, and water-soluble vitamin contents in the liver, blood and urine were measured during starvation. Urinary excretion of thiamin, riboflavin, vitamin B(6) metabolite 4-pyridoxic acid, nicotinamide metabolites and folate decreased during starvation, but that of vitamin B(12), pantothenic acid and biotin did not. As for blood vitamin levels, only blood vitamin B(1), plasma PLP and plasma folate levels decreased with starvation. All water-soluble vitamin contents in the liver decreased during starvation, whereas vitamin concentrations in the liver did not decrease. Starvation decreased only concentrations of vitamin B(12) and folate in the skeletal muscle. These results suggest that water-soluble vitamins were released from the liver, and supplied to the peripheral tissues to maintain vitamin nutrition. Our human study also suggested that the effect of fasting should be taken into consideration for subjects showing low urinary thiamin and high urinary riboflavin.

  18. Revisiting Hansen Solubility Parameters by Including Thermodynamics

    NARCIS (Netherlands)

    Louwerse, Manuel J; Fernández-Maldonado, Ana María; Rousseau, Simon; Moreau-Masselon, Chloe; Roux, Bernard; Rothenberg, Gadi

    2017-01-01

    The Hansen solubility parameter approach is revisited by implementing the thermodynamics of dissolution and mixing. Hansen's pragmatic approach has earned its spurs in predicting solvents for polymer solutions, but for molecular solutes improvements are needed. By going into the details of entropy

  19. Solubility of hydrogen in delta iron

    International Nuclear Information System (INIS)

    Shapovalov, V.I.; Trofimenko, V.V.

    1979-01-01

    The solubility of hydrogen in iron (less than 0.002 % impurities) at temperatures of 800-1510 deg C and a pressure of 100 atm was measured. The heat of solution of hydrogen in delta-Fe, equal to 73 kJ/g-atom, is by far greater than the corresponding values for α- and γ-Fe

  20. Radiculography with water-soluble contraste medium

    International Nuclear Information System (INIS)

    Araujo Pinheiro, R.S. de

    1987-01-01

    The etiologic diagnosis of the lumbar pain is discussed. The radiculography with water-soluble contrast medium is used and 250 cases are studied. Some practical criteria of indication executation and interpretation of the examination are reported. (M.A.C.) [pt

  1. Solubility of heavy metals added to MSW

    International Nuclear Information System (INIS)

    Lo, H.M.; Lin, K.C.; Liu, M.H.; Pai, T.Z.; Lin, C.Y.; Liu, W.F.; Fang, G.C.; Lu, C.; Chiang, C.F.; Wang, S.C.; Chen, P.H.; Chen, J.K.; Chiu, H.Y.; Wu, K.C.

    2009-01-01

    This paper aims to investigate the six heavy metal levels (Cd, Cr, Cu, Pb, Ni and Zn) in municipal solid waste (MSW) at different pHs. It intends to provide the baseline information of metals solubility in MSW co-disposed or co-digested with MSW incinerator ashes in landfill or anaerobic bioreactors or heavy metals contaminated in anaerobic digesters. One milliliter (equal to 1 mg) of each metal was added to the 100 ml MSW and the batch reactor test was carried out. The results showed that higher HNO 3 and NaOH were consumed at extreme pH of 1 and 13 compared to those from pH 2 to 11 due to the comparably higher buffer capacity. Pb was found to have the least soluble level, highest metal adsorption (%) and highest partitioning K d (l g -1 ) between pH 3 and 12. In contrast, Ni showed the highest soluble level, lowest metal adsorption (%) and lowest K d (l g -1 ) between pH 4 and 12. Except Ni and Cr, other four metals seemed to show the amphibious properties as comparative higher solubility was found in the acidic and basic conditions

  2. Solubility of heavy metals added to MSW

    Energy Technology Data Exchange (ETDEWEB)

    Lo, H.M. [Department of Environmental Engineering and Management, Chaoyang University of Technology, 168 Gifong E. Road, Wufong, Taichung County 41349, Taiwan (China)], E-mail: hmlo@cyut.edu.tw; Lin, K.C. [Department of Occupational Safety and Health, Chung Shan Medical University, 110, Sec. 1, Jiangguo N. Rd., Taichung 402, Taiwan (China); Liu, M.H.; Pai, T.Z. [Department of Environmental Engineering and Management, Chaoyang University of Technology, 168 Gifong E. Road, Wufong, Taichung County 41349, Taiwan (China); Lin, C.Y. [Department of Soil and Water Conservation, Chung Hsing University, 250 Kuokuang Road, Taichung 402, Taiwan (China); Liu, W.F. [Department of Electronical Engineering, Feng Chia University, 100 Wenhwa Road, Taichung 407, Taiwan (China); Fang, G.C. [Department of Environmental Engineering, Hungkuang University, 34 Chung-Chie Road, Sha Lu, Taichung 433, Taiwan (China); Lu, C. [Department of Environmental Engineering, Chung Hsing University, 250 Kuokuang Road, Taichung 402, Taiwan (China); Chiang, C.F. [Department of Health Risk Management, China Medical University, No. 91 Hsueh-Shih Road, Taichung 40402, Taiwan (China); Wang, S.C.; Chen, P.H.; Chen, J.K.; Chiu, H.Y.; Wu, K.C. [Department of Environmental Engineering and Management, Chaoyang University of Technology, 168 Gifong E. Road, Wufong, Taichung County 41349, Taiwan (China)

    2009-01-15

    This paper aims to investigate the six heavy metal levels (Cd, Cr, Cu, Pb, Ni and Zn) in municipal solid waste (MSW) at different pHs. It intends to provide the baseline information of metals solubility in MSW co-disposed or co-digested with MSW incinerator ashes in landfill or anaerobic bioreactors or heavy metals contaminated in anaerobic digesters. One milliliter (equal to 1 mg) of each metal was added to the 100 ml MSW and the batch reactor test was carried out. The results showed that higher HNO{sub 3} and NaOH were consumed at extreme pH of 1 and 13 compared to those from pH 2 to 11 due to the comparably higher buffer capacity. Pb was found to have the least soluble level, highest metal adsorption (%) and highest partitioning K{sub d} (l g{sup -1}) between pH 3 and 12. In contrast, Ni showed the highest soluble level, lowest metal adsorption (%) and lowest K{sub d} (l g{sup -1}) between pH 4 and 12. Except Ni and Cr, other four metals seemed to show the amphibious properties as comparative higher solubility was found in the acidic and basic conditions.

  3. Anomalous Solubility Behavior of Several Acidic Drugs

    Directory of Open Access Journals (Sweden)

    Alex Avdeef

    2014-04-01

    Full Text Available The “anomalous solubility behavior at higher pH values” of several acidic drugs originally studied by Higuchi et al. in 1953 [1], but hitherto not fully rationalized, has been re-analyzed using a novel solubility-pH analysis computer program, pDISOL-XTM. The program internally derives implicit solubility equations, given a set of proposed equilibria and constants (iteratively refined by weighted nonlinear regression, and does not require explicit Henderson-Hasselbalch equations. The re-analyzed original barbital, phenobarbital, oxytetracycline, and sulfathiazole solubility-pH data of Higuchi et al. is consistent with the presence of dimers in saturated solutions. In the case of barbital, phenobarbital and sulfathiazole, anionic dimers, reaching peak concentrations near pH 8. However, oxytetracycline indicated a pronounced tendency to form a cationic dimer, peaking near pH 2. Under the conditions of the original study, only barbital indicated a slight tendency to form a salt precipitate at pH > 6.8, with a highly unusual stoichiometry (consistent with a slope of 0.55 in the log S – pH plot: K+ + A2H- + 3HA D KA5H4(s. Thus the “anomaly” in the Higuchi data can be rationalized by invoking specific aggregated species.

  4. Changes in protein solubility, fermentative capacity, viscoelasticity ...

    African Journals Online (AJOL)

    Frozen dough should be stored for fewer than 21 days; time in which the loaf volume of bread made from frozen dough was approximately 40.84% smaller than that of fresh bread dough formulation. Keywords: French type bread, frozen dough, protein solubility, baking quality, viscoelasticity. African Journal of Biotechnology ...

  5. Solubility of Tc(IV) oxides

    International Nuclear Information System (INIS)

    Liu, D.J.; Fan, X.H.

    2005-01-01

    Full text of publication follows: The deep geological disposal of the high level radioactive wastes is expected to be a safer disposal method in most countries. The long-lived fission product 99 Tc is present in large quantities in nuclear wastes and its chemical behavior in aqueous solution is of considerable interest. Under the reducing conditions, expected to exist in a deep geological repository, it is generally predicted that technetium will be present as TcO 2 .nH 2 O. The solubility of Tc(IV) is used as a source term in performance assessment of radioactive waste repository. Technetium oxide was prepared by reduction of a technetate solution with Sn 2+ . The solubility of Tc(IV) oxide has been determined in simulated groundwater and re-distilled water under aerobic and anaerobic conditions. The effects of pH and CO 3 2- concentration of solution on solubility of Tc(IV) oxide were studied. The concentration of total technetium and Tc(IV) species in the solutions were periodically determined by separating the oxidized and reduced technetium species using a solvent extraction procedure and counting the beta activity of the 99 Tc with a liquid scintillation counter. The experimental results show that the rate of oxidation of Tc(IV) in simulated groundwater and re-distilled water is about (1.49∼1.86) x 10 -9 mol/(L.d) under aerobic conditions, but Tc(IV) in simulated groundwater and re-distilled water is not oxidized under anaerobic conditions. Under aerobic or anaerobic conditions the solubility of Tc(IV) oxide in simulated groundwater and re-distilled water is equal on the whole after centrifugation or ultrafiltration. The solubility of Tc(IV) oxide decreases with the increase of pH at pH 10 and is pH independent in the range 2 -8 to 10 -9 mol/L at 2 3 2- concentration. These data could be used to estimate the Tc(IV) solubility for cases where solubility limits transport of technetium in reducing environments of high-level waste repositories. (authors)

  6. Scoring function to predict solubility mutagenesis

    Directory of Open Access Journals (Sweden)

    Deutsch Christopher

    2010-10-01

    Full Text Available Abstract Background Mutagenesis is commonly used to engineer proteins with desirable properties not present in the wild type (WT protein, such as increased or decreased stability, reactivity, or solubility. Experimentalists often have to choose a small subset of mutations from a large number of candidates to obtain the desired change, and computational techniques are invaluable to make the choices. While several such methods have been proposed to predict stability and reactivity mutagenesis, solubility has not received much attention. Results We use concepts from computational geometry to define a three body scoring function that predicts the change in protein solubility due to mutations. The scoring function captures both sequence and structure information. By exploring the literature, we have assembled a substantial database of 137 single- and multiple-point solubility mutations. Our database is the largest such collection with structural information known so far. We optimize the scoring function using linear programming (LP methods to derive its weights based on training. Starting with default values of 1, we find weights in the range [0,2] so that predictions of increase or decrease in solubility are optimized. We compare the LP method to the standard machine learning techniques of support vector machines (SVM and the Lasso. Using statistics for leave-one-out (LOO, 10-fold, and 3-fold cross validations (CV for training and prediction, we demonstrate that the LP method performs the best overall. For the LOOCV, the LP method has an overall accuracy of 81%. Availability Executables of programs, tables of weights, and datasets of mutants are available from the following web page: http://www.wsu.edu/~kbala/OptSolMut.html.

  7. Solubility of gases in 1-alkyl-3methylimidazolium alkyl sulfate ionic liquids: Experimental determination and modeling

    International Nuclear Information System (INIS)

    Bermejo, María Dolores; Fieback, Tobias M.; Martín, Ángel

    2013-01-01

    Highlights: ► The solubility of CO 2 , CH 4 and C 2 H 6 in [emim][EtSO 4 ] is measured with a magnetic suspension balance. ► New data and literature results have been modeled with a Group Contribution equation of state. ► A specific group definition is required to model data of ionic liquids with a [MeSO 4 ] anion. ► Deviations between model and experiments are lower than 10% in most cases. ► Deviations of 34% are observed in the case of the solubility of ethane in the ionic liquid. -- Abstract: The solubility of different gases (carbon dioxide, methane, ethane, carbon monoxide and hydrogen) in ionic liquids with an alkyl sulfate anion has been modeled with the Group Contribution equation of state developed by Skjold-Jørgensen. New gas solubility measurements have been carried out with a high pressure magnetic suspension balance in order to cover pressure and temperature ranges not considered in previous studies and to obtain more experimental information for the correlation of parameters of the equation of state. New solubility measurements include the solubility of carbon dioxide in 1-ethyl 3-methyl imidazolium ethyl sulfate [emim][EtSO 4 ] at temperatures of 298 K and 348 K and pressures ranging from 0.3 MPa to 6.5 MPa, the solubility of methane in [emim][EtSO 4 ] at a temperature of 293 K and pressures ranging from 0.2 MPa to 10.2 MPa, and the solubility of ethane in [emim][EtSO 4 ] at temperatures of 323 K and 350 K and pressures ranging from 0.2 MPa to 4 MPa. Results show that the Group Contribution equation of state can be used to describe the solubility of gases in alkyl sulfate ionic liquids as well as infinite dilution coefficients of alkanes in the ionic liquids, with average deviations between experiments and calculations ranging from 1% to 10% in the case of mixtures with CO 2 , CO, CH 4 and H 2 with the alkyl sulfate ionic liquids to up to 34% in the case of the solubility of ethane in [emim][EtSO 4

  8. Thermodynamic model of Ni(II) solubility, hydrolysis and complex formation with ISA

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez-Siso, Maria Rosa; Duro, Lara; Bruno, Jordi [Amphos21, Barcelona (Spain); Gaona, Xavier; Altmaier, Marcus [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany). Inst. for Nuclear Waste Disposal

    2018-04-01

    The solubility of β-Ni(OH){sub 2}(cr) was investigated at T=(22±2) C in the absence and presence of α-isosaccharinic acid (ISA), the main degradation product of cellulose under alkaline pH conditions. Batch solubility experiments were performed from undersaturation conditions under inert gas (Ar) atmosphere. Solubility experiments in the absence of ISA were conducted in 0.5 and 3.0 M NaCl-NaOH solutions at 7.5 ≤ pH{sub m} ≤ 13 (with pH{sub m} = -log{sub 10}[H{sup +}]). XRD analyses of selected solid phases collected after completing the solubility experiments (∼300 days) confirmed that β-Ni(OH){sub 2}(cr) remains as solid phase controlling the solubility of Ni(II) in all investigated conditions. Based on the slope analysis (log{sub 10}[Ni] vs. pH{sub m}) of the solubility data and solid phase characterization, the equilibrium reactions β-Ni(OH){sub 2}(cr)+2 H{sup +} <=> Ni{sup 2+}+2 H{sub 2}O(l) and β-Ni(OH){sub 2}(cr) <=> Ni(OH){sub 2}(aq) were identified as controlling the solubility of Ni(II) within the investigated pH{sub m} region. The conditional equilibrium constants determined from the solubility experiments at different ionic strengths were evaluated with the specific ion interaction theory (SIT). In contrast to the current thermodynamic selection in the NEA-TDB, solubility data collected in the present work does not support the formation of the anionic hydrolysis species Ni(OH){sub 3}{sup -} up to pH{sub m} ≤ 13.0. Solubility experiments in the presence of ISA were conducted in 0.5 M NaCl-NaOH-NaISA solutions with 0.01 M ≤ [NaISA] ≤ 0.2 M and 9 ≤ pH{sub m} ≤ 13. XRD analyses confirmed that β-Ni(OH){sub 2}(cr) is also the solid phase controlling the solubility of Ni(II) in the presence of ISA. Solubility data of all investigated systems can be properly explained with chemical and thermodynamic models including the formation of the complexes NiOHISA(aq), Ni(OH){sub 2}ISA{sup -} and Ni(OH){sub 3}ISA{sup 2-}. The reported data confirm

  9. Nuclide Release Behavior from a Repository for a Pyro-process HLW and SF due to Variation of the MWCF Properties

    International Nuclear Information System (INIS)

    Lee, Youn Myoung; Hwang, Yong Soo

    2009-01-01

    An assessment program for an optional evaluation of a repository both for disposal of such high-level wastes (HLWs) from various steps of pyro-processes of PWR spent nuclear fuel (SF) and for direct disposal of PWR and CANDU SFs has been developed by utilizing general purpose GoldSim developing tool, by which nuclide transports in the near- and far-field of a repository as well as a transport through a biosphere under various natural and manmade disruptive events affecting a nuclide release could be modeled and evaluated. KAERI has been in charge of modeling and developing assessment tools by which the above mentioned repository system could be assessed in accordance with various features, events, and processes (FEPs) that could happen in and around the repository system. To cope with such various natural and manmade disruptive FEPs as well as normal release scenarios, all the possible cases in view of the Korean circumstances should be modeled and have been evaluated even though we have not yet have any repository. A possible case, among many others, with the variation of such physical properties as the fracture width and the rock matrix diffusion depth, associated with the natural fractures in the geological rock media, along which nuclide could be transported preferentially with the flow of groundwater is considered in the current study. Due to whatever the reason, such as e,g., the earthquake or human intrusion, it is assumed that the physical properties of the major water conducting fault (MWCF) is changed resulting in the size of fracture width and the matrix diffusion depth. For such case another illustration is made for probabilistic evaluation of a hypothetical Korean HLW repository, as similarly done in the previous studies

  10. Project Entsorgungsnachweis, 'Demonstration of disposal feasibility for SF/HLW/ILW in the Opalinus Clay of the Zuercher Weinland', Background, Objectives and Overview

    International Nuclear Information System (INIS)

    Schneider, Juerg; Zuidema, P.

    2004-01-01

    Juerg Schneider (Nagra, Switzerland) described the project on the Opalinus Clay (Project Entsorgungsnachweis, demonstration of disposal feasibility for SF/HLW/ILW in the Opalinus Clay of the Zuercher Weinland) for which the main objective is to demonstrate disposal feasibility and to provide input to the decision how to proceed. The report structure was described, the focus of the presentation being the report that aimed to provide a comprehensive assessment of long-term safety. The current situation was described in the presentation as follows: - The key need is to provide arguments for having proposed a good system for which there is sufficient understanding to allow a credible safety evaluation. - Alternative options exist, on which attention is maintained by a task-force. However, Nagra is confident in its results on Project Entsorgungsnachweis, given the knowledge base that currently exists, and has put forward a proposal, for consideration by the Swiss Government, to focus future work on the Opalinus Clay (OPA) of the Zuercher Weinland. - Making the safety case requires a proper integration of science, engineering and safety assessment. - Three key issues were identified in making a safety case: completeness, sufficient safety, and robustness to diminish the importance of uncertainties. - A safety case needs to be adequate to support a decision to proceed to the next stage in the programme, with multiple arguments including the existence of reserve FEP's. - The interacting functions of the relevant teams were viewed as a key component of the process of preparing a safety case: management; science; safety assessment; bias audit. During the discussion, the role of the bias team was recognised as being helpful to ensure completeness, as well as using the NEA FEP database as a check list. When speaking about sufficient safety, it should not imply predictive capability but rather that there is enough confidence in the current level of understanding to

  11. Solubility behavior and biopharmaceutical classification of novel high-solubility ciprofloxacin and norfloxacin pharmaceutical derivatives.

    Science.gov (United States)

    Breda, Susana A; Jimenez-Kairuz, Alvaro F; Manzo, Ruben H; Olivera, María E

    2009-04-17

    The hydrochlorides of the 1:3 aluminum:norfloxacin and aluminum:ciprofloxacin complexes were characterized according to the Biopharmaceutics Classification System (BCS) premises in comparison with their parent compounds. The pH-solubility profiles of the complexes were experimentally determined at 25 and 37 degrees C in the range of pH 1-8 and compared to that of uncomplexed norfloxacin and ciprofloxacin. Both complexes are clearly more soluble than the antibiotics themselves, even at the lowest solubility pHs. The increase in solubility was ascribed to the species controlling solubility, which were analyzed in the solid phases at equilibrium at selected pHs. Additionally, permeability was set as low, based on data reported in the scientific literature regarding oral bioavailability, intestinal and cell cultures permeabilities and also considering the influence of stoichiometric amounts of aluminum. The complexes fulfill the BCS criterion to be classified as class 3 compounds (high solubility/low permeability). Instead, the active pharmaceutical ingredients (APIs) currently used in solid dosage forms, norfloxacin and ciprofloxacin hydrochloride, proved to be BCS class 4 (low solubility/low permeability). The solubility improvement turns the complexes as potential biowaiver candidates from the scientific point of view and may be a good way for developing more dose-efficient formulations. An immediate release tablet showing very rapid dissolution was obtained. Its dissolution profile was compared to that of the commercial ciprofloxacin hydrochloride tablets allowing to dissolution of the complete dose at a critical pH such as 6.8.

  12. Effect of Cyclodextrin Complexation on the Aqueous Solubility and Solubility/Dose Ratio of Praziquantel

    OpenAIRE

    Maragos, Stratos; Archontaki, Helen; Macheras, Panos; Valsami, Georgia

    2009-01-01

    Praziquantel (PZQ), the primary drug of choice in the treatment of schistosomiasis, is a highly lipophilic drug that possesses high permeability and low aqueous solubility and is, therefore, classified as a Class II drug according to the Biopharmaceutics Classification System (BCS). In this work, β-cyclodextrin (β-CD) and hydroxypropyl-β-cyclodextrin (HP-β-CD) were used in order to determine whether increasing the aqueous solubility of a drug by complexation with CDs, a BCS-Class II compound ...

  13. Synthesis of LiBOB Fine Powder to Increase Solubility

    Directory of Open Access Journals (Sweden)

    Etty Marti Wigayati

    2017-04-01

    Full Text Available Lithium bis (oxalate borate or LiBOB compound has captured interest of researchers, because it is potentially viable to be used as electrolyte salt in lithium-ion battery system. This compound is easy to synthesize and considered to be more environmentally friendly compared to conventional electrolyte salt because LiBOB does not contain halogen element. This research focused on the synthesis of LiBOB fine powder, which main purpose is improving LiBOB salt solubility in liquid electrolyte solution. This will aid the ion transfer between electrodes which in turn will increase the electrolyte performance. Solid state reaction was employed in this experiment. Synthesis of LiBOB compound was performed by reacting oxalic acid dihydrate, lithium hydroxide monohydrate, and boric acid. The resulting powder was then processed into fine powder using ball milling technique with varying milling time (0, 6, 10, and 13 hour. Microstructure of the sample was then analyzed to obtain information regarding phase formation, functional groups, grain surface morphology, surface area, pore volume, solubility, and ionic conductivity. The analysis shown that LiBOB and LiBOB hydrate phase was formed during the reaction, there was no changed in existing phase during milling process, crystallinity index was shifted to lower value but there was no difference in functional groups. Highest value in surface area was found to be 83.11 m2/g, with pore volume of 1.21311e+02 A at 10 hours milling. Smaller powder size resulted in higher solubility, unfortunately the ionic conductivity was found to be decreased.

  14. Novel electrosprayed nanospherules for enhanced aqueous solubility and oral bioavailability of poorly water-soluble fenofibrate.

    Science.gov (United States)

    Yousaf, Abid Mehmood; Mustapha, Omer; Kim, Dong Wuk; Kim, Dong Shik; Kim, Kyeong Soo; Jin, Sung Giu; Yong, Chul Soon; Youn, Yu Seok; Oh, Yu-Kyoung; Kim, Jong Oh; Choi, Han-Gon

    2016-01-01

    The purpose of the present research was to develop a novel electrosprayed nanospherule providing the most optimized aqueous solubility and oral bioavailability for poorly water-soluble fenofibrate. Numerous fenofibrate-loaded electrosprayed nanospherules were prepared with polyvinylpyrrolidone (PVP) and Labrafil(®) M 2125 as carriers using the electrospray technique, and the effect of the carriers on drug solubility and solvation was assessed. The solid state characterization of an optimized formulation was conducted by scanning electron microscopy, powder X-ray diffraction, differential scanning calorimetry, and Fourier transform infrared spectroscopic analyses. Oral bioavailability in rats was also evaluated for the formulation of an optimized nanospherule in comparison with free drug and a conventional fenofibrate-loaded solid dispersion. All of the electrosprayed nanospherule formulations had remarkably enhanced aqueous solubility and dissolution compared with free drug. Moreover, Labrafil M 2125, a surfactant, had a positive influence on the solubility and dissolution of the drug in the electrosprayed nanospherule. Increases were observed as the PVP/drug ratio increased to 4:1, but higher ratios gave no significant increases. In particular, an electrosprayed nanospherule composed of fenofibrate, PVP, and Labrafil M 2125 at the weight ratio of 1:4:0.5 resulted in a particle size of water-soluble fenofibrate.

  15. Ingestion of guar gum hydrolysate, a soluble fiber, increases calcium absorption in totally gastrectomized rats.

    Science.gov (United States)

    Hara, H; Suzuki, T; Kasai, T; Aoyama, Y; Ohta, A

    1999-01-01

    Gastrectomy induces osteopenia. We examined the effects of feeding a diet containing soluble dietary fiber, guar gum hydrolysate (GGH, 50 g/kg diet), on intestinal calcium absorption and bone mineralization in totally gastrectomized (Roux-en-Y esophagojejunostomy) rats by comparing them with those in two control groups (laparotomized and bypassed rats). In the bypassed rats, chyme bypassed the duodenum and upper jejunum without gastrectomy. In a second separate experiment, we compared calcium absorption and bone mineralization in the gastrectomized rats fed diets containing soluble and insoluble calcium salts and in bypassed rats fed insoluble calcium. In Experiment 1, apparent absorption of calcium supplied as a water-insoluble salt was more than 50% lower in gastrectomized rats than in the intact (laparotomized) or bypassed rats 3 wk after the start of feeding the test diets (P Calcium absorption was higher (P Experiment 2, absorption of soluble calcium in the gastrectomized rats did not differ from the absorption of calcium from calcium carbonate by bypassed rats. The soluble calcium pool in the cecal contents was significantly lower in gastrectomized rats (Experiment 1) than in intact or bypassed control rats, and was higher (P calcium absorption correlated most closely (r = 0.787, P calcium content was significantly lower in gastrectomized rats fed insoluble calcium than in bypassed rats fed the same diet, but was partially restored in the rats fed soluble calcium (Experiment 2). Bone calcium was not increased by feeding GGH in gastrectomized rats (Experiment 1). We conclude that the severely diminished calcium absorption following total gastrectomy is totally due to a decrease in calcium solubilization, and feeding GGH partially restores calcium absorption. The decrease in bone calcium that occurs as a result of gastrectomy is mainly due to diminished intestinal calcium absorption.

  16. Combined experimental and computational modelling studies of the solubility of nickel in strontium titanate

    NARCIS (Netherlands)

    Beale, A.M.; Paul, M.; Sankar, G.; Oldman, R.J.; Catlow, R.A.; French, S.; Fowles, M.

    2009-01-01

    A combination of X-ray techniques and atomistic computational modelling has been used to study the solubility of Ni in SrTiO3 in relation to the application of this material for the catalytic partial oxidation of methane. The experiments have demonstrated that low temperature, hydrothermal synthesis

  17. Studies on solubility, glass-adsorption, extraction and evaporation of DDT using 36Cl-DDT

    International Nuclear Information System (INIS)

    Duursma, E.K.; Vas, D.; Marchand, M.

    1974-01-01

    A series of experiments are reported to investigate the behaviour of DDT in seawater using 36 Cl-DDT to determine DDT solubility in seawater, its adsorption to walls of glass vessels, the reproductibility of DDT extraction from seawater by organic solvents and its volatility when brought to dryness from solution

  18. Proposed siting of a HLW Repository to the Municipality of Oskarshamn - Dialogue with the public and how it responds

    Energy Technology Data Exchange (ETDEWEB)

    Carlsson, Torsten; Hallberg, Krister [Municipality of Oskarshamn (Sweden); Andersson, Kjell [Karinta-Konsult, Taeby (Sweden); Aahagen, Harald [Aahagen and Co., Vaederstad (Sweden); Braakenhielm, Carl-Reinhold [Univ. of Uppsala (Sweden). Dept. of Theology

    2001-07-01

    The Swedish High Level Nuclear Waste programme is in a decision making phase where the issue is if three site investigations in three municipalities are to be initiated or not. In Oskarshamn the answer to this question is currently being prepared by three newly formed working groups. The task for these groups is to prepare reports to the Municipality Council with recommendations how this question should be answered. The work is conducted following the 'Oskarshamn Model'. The municipality is expected to answer this question early 2002. The major challenge during 2001 for the municipality is to form a dialogue with the region, the Oskarshamn public and in particular the land owners and neighbours (the most effected parties) to the proposed area of investigation. With the dialogue, the municipality should be able to continue to take decisions and post possible conditions that are supported and recognised by a large majority of the people in Oskarshamn. Should site investigations be accepted, the experience from the work conducted to date and the 'Oskarshamn Model' will continue to form a strong base for the municipality participation in the six to seven year long investigation phase.

  19. Proposed siting of a HLW Repository to the Municipality of Oskarshamn - Dialogue with the public and how it responds

    International Nuclear Information System (INIS)

    Carlsson, Torsten; Hallberg, Krister; Andersson, Kjell; Aahagen, Harald; Braakenhielm, Carl-Reinhold

    2001-01-01

    The Swedish High Level Nuclear Waste programme is in a decision making phase where the issue is if three site investigations in three municipalities are to be initiated or not. In Oskarshamn the answer to this question is currently being prepared by three newly formed working groups. The task for these groups is to prepare reports to the Municipality Council with recommendations how this question should be answered. The work is conducted following the 'Oskarshamn Model'. The municipality is expected to answer this question early 2002. The major challenge during 2001 for the municipality is to form a dialogue with the region, the Oskarshamn public and in particular the land owners and neighbours (the most effected parties) to the proposed area of investigation. With the dialogue, the municipality should be able to continue to take decisions and post possible conditions that are supported and recognised by a large majority of the people in Oskarshamn. Should site investigations be accepted, the experience from the work conducted to date and the 'Oskarshamn Model' will continue to form a strong base for the municipality participation in the six to seven year long investigation phase

  20. Solubility of hydrogen in metals and its effect of pore-formation and embrittlement. Ph.D. Thesis

    Science.gov (United States)

    Shahani, H. R.

    1984-01-01

    The effect of alloying elements on hydrogen solubility were determined by evaluating solubility equations and interaction coefficients. The solubility of dry hydrogen at one atmosphere was investigated in liquid aluminum, Al-Ti, Al-Si, Al-Fe, liquid gold, Au-Cu, and Au-Pd. The design of rapid heating and high pressure casting furnaces used in meta foam experiments is discussed as well as the mechanism of precipitation of pores in melts, and the effect of hydrogen on the shrinkage porosity of Al-Cu and Al-Si alloys. Hydrogen embrittlement in iron base alloys is also examined.

  1. Solubility of unirradiated UO2 fuel in aqueous solutions. Comparison between experimental and calculated (EQ3/6) data

    International Nuclear Information System (INIS)

    Ollila, K.

    1995-11-01

    The solubility behaviour of unirradiated UO 2 pellets was studied under oxic (air-saturated) and anoxic (N 2 ) conditions in deionized water, in sodium bicarbonate solutions with varying bicarbonate content (60 - 600 ppm), in Allard groundwater simulating granitic fresh groundwater conditions, and in bentonite water simulating the effects of bentonite on granitic fresh groundwater (25 deg C). The release of uranium was measured during static batch dissolution experiments of long duration (2-6 years). A comparison was made with the theoretical solubility data calculated with the geochemical code EQ3/6 in order to evaluate solubility (steady state) limiting factors. (orig.) (26 refs., 32 figs., 13 tabs.)

  2. Safety aspects of long-term dry interim storage of type-B spent fuel and HLW transport casks

    International Nuclear Information System (INIS)

    Wolff, D.; Probst, U.; Voelzke, H.; Droste, B.; Roedel, R.

    2004-01-01

    Based on the German decision to minimise transports of spent fuel casks between nuclear power plants, reprocessing plants and central storage facilities several on-site storage facilities have been licensed till the end of 2003. Because of the large amount of type-B transport casks which are going to be used for long-term interim storage the question of time limited type-B license maintenance during the storage period of up to 40 years has been discussed under different aspects. This paper describes present technical aspects of the discussion. A main aspect of transport cask qualification for interim storage is the long-term behaviour of the metallic seal lid system. Concerning this results from current experimental long-term tests with metallic ''Helicoflex''-seals in which pool water is enclosed are presented. The test series has been performed by the Federal Institute for Materials Research and Testing (BAM) on behalf of the Federal Office for Radiation Protection (BfS) since 2001. Finally, the paper presents a German concept for an authorities' and technical experts' exchange of experience, know-how and state of the art referring to cask dispatch in nuclear facilities. BAM has taken over a central role in this so-called ''co-ordinating institution for cask dispatching information'' (''KOBAF'') which contains an online data base and a technical working group meeting twice a year. The goal is to keep comparable technical standards for all nuclear sites and storage facilities which are going to load and dispatch casks of the same or similar types under the responsibility of different German state governments for the next decades

  3. Solubility of cobalt in primary circuit solutions

    International Nuclear Information System (INIS)

    Lambert, I.; Joyer, F.

    1992-01-01

    The solubility of cobalt ferrite (CoFe 2 O 4 ) was measured in PWR primary circuit conditions, in the temperature range 250-350 deg C, and the results were compared with the ones obtained on magnetite and nickel ferrite. As in the former cases, it was found that, in the prevailing primary circuit conditions, the solubility of the cobalt ferrite was minimum at temperatures around 300 deg C, for cobalt as well as for iron. The equilibrium iron concentration is significantly lower than in the case of magnetite. The results are discussed in relation with the POTHY code, based only on thermodynamic laws and data, used for the prediction of the primary circuit chemistry

  4. Biochemical synthesis of water soluble conducting polymers

    Science.gov (United States)

    Bruno, Ferdinando F.; Bernabei, Manuele

    2016-05-01

    An efficient biomimetic route for the synthesis of conducting polymers/copolymers complexed with lignin sulfonate and sodium (polystyrenesulfonate) (SPS) will be presented. This polyelectrolyte assisted PEG-hematin or horseradish peroxidase catalyzed polymerization of pyrrole (PYR), 3,4 ethyldioxithiophene (EDOT) and aniline has provided a route to synthesize water-soluble conducting polymers/copolymers under acidic conditions. The UV-vis, FTIR, conductivity and cyclic voltammetry studies for the polymers/copolymer complex indicated the presence of a thermally stable and electroactive polymers. Moreover, the use of water-soluble templates, used as well as dopants, provided a unique combination of properties such as high electronic conductivity, and processability. These polymers/copolymers are nowadays tested/evaluated for antirust features on airplanes and helicopters. However, other electronic applications, such as photovoltaics, for transparent conductive polyaniline, actuators, for polypyrrole, and antistatic films, for polyEDOT, will be proposed.

  5. Biochemical synthesis of water soluble conducting polymers

    Energy Technology Data Exchange (ETDEWEB)

    Bruno, Ferdinando F., E-mail: Ferdinando-Bruno@uml.edu [US Army Natick Soldier Research, Development and Engineering Center, Natick, MA 01760 (United States); Bernabei, Manuele [ITAF, Test Flight Centre, Chemistry Dept. Pratica di Mare AFB, 00071 Pomezia (Rome), Italy (UE) (Italy)

    2016-05-18

    An efficient biomimetic route for the synthesis of conducting polymers/copolymers complexed with lignin sulfonate and sodium (polystyrenesulfonate) (SPS) will be presented. This polyelectrolyte assisted PEG-hematin or horseradish peroxidase catalyzed polymerization of pyrrole (PYR), 3,4 ethyldioxithiophene (EDOT) and aniline has provided a route to synthesize water-soluble conducting polymers/copolymers under acidic conditions. The UV-vis, FTIR, conductivity and cyclic voltammetry studies for the polymers/copolymer complex indicated the presence of a thermally stable and electroactive polymers. Moreover, the use of water-soluble templates, used as well as dopants, provided a unique combination of properties such as high electronic conductivity, and processability. These polymers/copolymers are nowadays tested/evaluated for antirust features on airplanes and helicopters. However, other electronic applications, such as photovoltaics, for transparent conductive polyaniline, actuators, for polypyrrole, and antistatic films, for polyEDOT, will be proposed.

  6. Biochemical synthesis of water soluble conducting polymers

    International Nuclear Information System (INIS)

    Bruno, Ferdinando F.; Bernabei, Manuele

    2016-01-01

    An efficient biomimetic route for the synthesis of conducting polymers/copolymers complexed with lignin sulfonate and sodium (polystyrenesulfonate) (SPS) will be presented. This polyelectrolyte assisted PEG-hematin or horseradish peroxidase catalyzed polymerization of pyrrole (PYR), 3,4 ethyldioxithiophene (EDOT) and aniline has provided a route to synthesize water-soluble conducting polymers/copolymers under acidic conditions. The UV-vis, FTIR, conductivity and cyclic voltammetry studies for the polymers/copolymer complex indicated the presence of a thermally stable and electroactive polymers. Moreover, the use of water-soluble templates, used as well as dopants, provided a unique combination of properties such as high electronic conductivity, and processability. These polymers/copolymers are nowadays tested/evaluated for antirust features on airplanes and helicopters. However, other electronic applications, such as photovoltaics, for transparent conductive polyaniline, actuators, for polypyrrole, and antistatic films, for polyEDOT, will be proposed.

  7. Soluble organic nanotubes for catalytic systems

    Science.gov (United States)

    Xiong, Linfeng; Yang, Kunran; Zhang, Hui; Liao, Xiaojuan; Huang, Kun

    2016-03-01

    In this paper, we report a novel method for constructing a soluble organic nanotube supported catalyst system based on single-molecule templating of core-shell bottlebrush copolymers. Various organic or metal catalysts, such as sodium prop-2-yne-1-sulfonate (SPS), 1-(2-(prop-2-yn-1-yloxy)ethyl)-1H-imidazole (PEI) and Pd(OAc)2 were anchored onto the tube walls to functionalize the organic nanotubes via copper-catalyzed azide-alkyne cycloaddition (CuAAC) reaction. Depending on the ‘confined effect’ and the accessible cavity microenvironments of tubular structures, the organic nanotube catalysts showed high catalytic efficiency and site-isolation features. We believe that the soluble organic nanotubes will be very useful for the development of high performance catalyst systems due to their high stability of support, facile functionalization and attractive textural properties.

  8. Soluble organic nanotubes for catalytic systems.

    Science.gov (United States)

    Xiong, Linfeng; Yang, Kunran; Zhang, Hui; Liao, Xiaojuan; Huang, Kun

    2016-03-18

    In this paper, we report a novel method for constructing a soluble organic nanotube supported catalyst system based on single-molecule templating of core–shell bottlebrush copolymers. Various organic or metal catalysts, such as sodium prop-2-yne-1-sulfonate (SPS), 1-(2-(prop-2-yn-1-yloxy)ethyl)-1H-imidazole (PEI) and Pd(OAc)2 were anchored onto the tube walls to functionalize the organic nanotubes via copper-catalyzed azide-alkyne cycloaddition (CuAAC) reaction. Depending on the 'confined effect' and the accessible cavity microenvironments of tubular structures, the organic nanotube catalysts showed high catalytic efficiency and site-isolation features. We believe that the soluble organic nanotubes will be very useful for the development of high performance catalyst systems due to their high stability of support, facile functionalization and attractive textural properties.

  9. Hydrogen solubility measurements of analyzed tall oil fractions and a solubility model

    International Nuclear Information System (INIS)

    Uusi-Kyyny, Petri; Pakkanen, Minna; Linnekoski, Juha; Alopaeus, Ville

    2017-01-01

    Highlights: • Hydrogen solubility was measured in four tall oil fractions between 373 and 597 K. • Continuous flow synthetic isothermal and isobaric method was used. • A Henry’s law model was developed for the distilled tall oil fractions. • The complex composition of the samples was analyzed and is presented. - Abstract: Knowledge of hydrogen solubility in tall oil fractions is important for designing hydrotreatment processes of these complex nonedible biobased materials. Unfortunately measurements of hydrogen solubility into these fractions are missing in the literature. This work reports hydrogen solubility measured in four tall oil fractions between 373 and 597 K and at pressures from 5 to 10 MPa. Three of the fractions were distilled tall oil fractions their resin acids contents are respectively 2, 20 and 23 in mass-%. Additionally one fraction was a crude tall oil (CTO) sample containing sterols as the main neutral fraction. Measurements were performed using a continuous flow synthetic isothermal and isobaric method based on the visual observation of the bubble point. Composition of the flow was changed step-wise for the bubble point composition determination. We assume that the tall oil fractions did not react during measurements, based on the composition analysis performed before and after the measurements. Additionally the densities of the fractions were measured at atmospheric pressure from 293.15 to 323.15 K. A Henry’s law model was developed for the distilled tall oil fractions describing the solubility with an absolute average deviation of 2.1%. Inputs of the solubility model are temperature, total pressure and the density of the oil at 323.15 K. The solubility of hydrogen in the CTO sample can be described with the developed model with an absolute average deviation of 3.4%. The solubility of hydrogen increases both with increasing pressure and/or increasing temperature. The more dense fractions of the tall oil exhibit lower hydrogen

  10. Solubility and Permeability Studies of Aceclofenac in Different Oils

    African Journals Online (AJOL)

    The solubility and permeability of aceclofenac were compared with the hydroalcoholic solution of ... the use of lipid based systems such as micro- or .... carriers/vehicles for enhanced solubility and permeability ... modifications: A recent review.

  11. Soluble L-selectin levels predict survival in sepsis

    DEFF Research Database (Denmark)

    Seidelin, Jakob B; Nielsen, Ole H; Strøm, Jens

    2002-01-01

    To evaluate serum soluble L-selectin as a prognostic factor for survival in patients with sepsis.......To evaluate serum soluble L-selectin as a prognostic factor for survival in patients with sepsis....

  12. Soluble polymer conjugates for drug delivery.

    Science.gov (United States)

    Minko, Tamara

    2005-01-01

    The use of water-soluble polymeric conjugates as drug carriers offers several possible advantages. These advantages include: (1) improved drug pharmacokinetics; (2) decreased toxicity to healthy organs; (3) possible facilitation of accumulation and preferential uptake by targeted cells; (4) programmed profile of drug release. In this review, we will consider the main types of useful polymeric conjugates and their role and effectiveness as carriers in drug delivery systems.: © 2005 Elsevier Ltd . All rights reserved.

  13. Thermal degradation of organo-soluble polyimides

    Institute of Scientific and Technical Information of China (English)

    黄俐研; 史燚; 金熹高

    1999-01-01

    The thermal degradation behavior of two organo-soluble polyimides was investigated by high resolution pyrolysis-gas chromatography/mass spectrometry. The pyrolyzates of the polymers at various temperatures were identified and characterized quantitatively. The relationship between the polymer structure and pyrolyzate distribution was discussed. The kinetic parameters of the thermal degradation were calculated based on thermogravimetric measurements. Finally, the thermal degradation mechanism for the polymers was suggested.

  14. Measurement of Soluble Biomarkers by Flow Cytometry

    OpenAIRE

    Antal-Szalm?s, P?ter; Nagy, B?la; Debreceni, Ildik? Beke; Kappelmayer, J?nos

    2013-01-01

    Microparticle based flow cytometric assays for determination of the level of soluble biomarkers are widely used in several research applications and in some diagnostic setups. The major advantages of these multiplex systems are that they can measure a large number of analytes (up to 500) at the same time reducing assay time, costs and sample volume. Most of these assays are based on antigen-antibody interactions and work as traditional immunoassays, but nucleic acid alterations ? by using spe...

  15. Water Soluble Polymers for Pharmaceutical Applications

    OpenAIRE

    Veeran Gowda Kadajji; Guru V. Betageri

    2011-01-01

    Advances in polymer science have led to the development of novel drug delivery systems. Some polymers are obtained from natural resources and then chemically modified for various applications, while others are chemically synthesized and used. A large number of natural and synthetic polymers are available. In the present paper, only water soluble polymers are described. They have been explained in two categories (1) synthetic and (2) natural. Drug polymer conjugates, block copolymers, hydrogel...

  16. Solubility of plutonium and waste evaporation

    International Nuclear Information System (INIS)

    Karraker, D.G.

    1993-01-01

    Chemical processing of irradiated reactor elements at the Savannah River Site separates uranium, plutonium and fission products; fission products and process-added chemicals are mixed with an excess of NaOH and discharged as a basic slurry into large underground tanks for temporary storage. The slurry is composed of base-insoluble solids that settle to the bottom of the tank; the liquid supemate contains a mixture of base-soluble chemicals--nitrates, nitrites aluminate, sulfate, etc. To conserve space in the waste tanks, the sup