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Sample records for waste hlw solutions

  1. 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.

  2. 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.

  3. 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

  4. 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.)

  5. 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

  6. 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

  7. Radioactive waste management solutions

    International Nuclear Information System (INIS)

    Siemann, Michael

    2015-01-01

    One of the more frequent questions that arise when discussing nuclear energy's potential contribution to mitigating climate change concerns that of how to manage radioactive waste. Radioactive waste is produced through nuclear power generation, but also - although to a significantly lesser extent - in a variety of other sectors including medicine, agriculture, research, industry and education. The amount, type and physical form of radioactive waste varies considerably. Some forms of radioactive waste, for example, need only be stored for a relatively short period while their radioactivity naturally decays to safe levels. Others remain radioactive for hundreds or even hundreds of thousands of years. Public concerns surrounding radioactive waste are largely related to long-lived high-level radioactive waste. Countries around the world with existing nuclear programmes are developing longer-term plans for final disposal of such waste, with an international consensus developing that the geological disposal of high-level waste (HLW) is the most technically feasible and safe solution. This article provides a brief overview of the different forms of radioactive waste, examines storage and disposal solutions, and briefly explores fuel recycling and stakeholder involvement in radioactive waste management decision making

  8. 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

  9. 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

  10. 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).

  11. 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

  12. 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.

  13. 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

  14. 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)

  15. High-Level Waste (HLW) Feed Process Control Strategy

    International Nuclear Information System (INIS)

    STAEHR, T.W.

    2000-01-01

    The primary purpose of this document is to describe the overall process control strategy for monitoring and controlling the functions associated with the Phase 1B high-level waste feed delivery. This document provides the basis for process monitoring and control functions and requirements needed throughput the double-shell tank system during Phase 1 high-level waste feed delivery. This document is intended to be used by (1) the developers of the future Process Control Plan and (2) the developers of the monitoring and control system

  16. Novel waste forms for HLW and ILW immobilisation

    International Nuclear Information System (INIS)

    Lee, William E.; Milestone, Neil B.; Ojovan, Michael I.; Hyatt, Neil C.; Stennett, Martin C.; Setiadi, Anthony; Zhou, Qizhi

    2006-01-01

    The complex nature and heterogeneity of legacy wastes means that a toolbox of different host systems must be developed in which to immobilize them. New zirconolite ceramic, glass composite materials and novel cement systems including calcium sulpho aluminate cements and alkali activated slags being examined in the Immobilisation Science Laboratory at the University of Sheffield are described. (authors)

  17. 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

    This report describes the development and testing of new glass formulations for high aluminum waste streams that achieve high waste loadings while maintaining high processing rates. The testing was based on the compositions of Hanford High Level Waste (HLW) with limiting concentrations of aluminum specified by the Office of River Protection (ORP). The testing identified glass 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 and small scale melt rate screening tests. The results were used to select compositions for subsequent testing in a DuraMelter 100 (DM100) system. These tests were used to determine processing rates for the selected formulations as well as to examine the effects of increased glass processing temperature, and the form of aluminum in the waste simulant. Finally, one of the formulations was selected for large-scale confirmatory testing on the HLW Pilot Melter (DM1200), which is a one third scale prototype of the Hanford Tank Waste Treatment and Immobilization Plant (WTP) HLW melter and off-gas treatment system. This work builds on previous work performed at the Vitreous State Laboratory (VSL) for Department of Energy (DOE) to increase waste loading and processing rates for high-iron HLW waste streams as well as previous tests conducted for ORP on the same high-aluminum waste composition used in the present work and other Hanford HLW compositions. The scope of this study was outlined in a Test Plan that was prepared in response to an ORP-supplied statement of work. It is currently estimated that the number of HLW canisters to be produced in the WTP is about 13,500 (equivalent to 40,500 MT glass). This estimate is based upon the inventory of the tank wastes, the anticipated performance of the sludge treatment processes, and current understanding of the capability of the borosilicate glass waste form

  18. 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

    This report describes the development and testing of new glass formulations for high aluminum waste streams that achieve high waste loadings while maintaining high processing rates. The testing was based on the compositions of Hanford High Level Waste (HLW) with limiting concentrations of aluminum specified by the Office of River Protection (ORP). The testing identified glass 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 and small scale melt rate screening tests. The results were used to select compositions for subsequent testing in a DuraMelter 100 (DM100) system. These tests were used to determine processing rates for the selected formulations as well as to examine the effects of increased glass processing temperature, and the form of aluminum in the waste simulant. Finally, one of the formulations was selected for large-scale confirmatory testing on the HLW Pilot Melter (DM1200), which is a one third scale prototype of the Hanford Tank Waste Treatment and Immobilization Plant (WTP) HLW melter and off-gas treatment system. This work builds on previous work performed at the Vitreous State Laboratory (VSL) for Department of Energy (DOE) to increase waste loading and processing rates for high-iron HLW waste streams as well as previous tests conducted for ORP on the same high-aluminum waste composition used in the present work and other Hanford HLW compositions. The scope of this study was outlined in a Test Plan that was prepared in response to an ORP-supplied statement of work. It is currently estimated that the number of HLW canisters to be produced in the WTP is about 13,500 (equivalent to 40,500 MT glass). This estimate is based upon the inventory of the tank wastes, the anticipated performance of the sludge treatment processes, and current understanding of the capability of the borosilicate glass waste form

  19. High level waste (HLW) steam reducing station evaluation

    International Nuclear Information System (INIS)

    Gannon, R.E.

    1993-01-01

    Existing pressure equipment in High Level Waste does not have a documented technical baseline. Based on preliminary reviews, the existing equipment seems to be based on system required capacity instead of system capability. A planned approach to establish a technical baseline began September 1992 and used the Works Management System preventive maintenance schedule. Several issues with relief valves being undersized on steam reducing stations created a need to determine the risk of maintaining the steam in service. An Action Plan was developed to evaluate relief valves that did not have technical baselines and provided a path forward for continued operation. Based on Action Plan WER-HLE-931042, the steam systems will remain in service while the designs are being developed and implemented

  20. 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

  1. Optimization method for dimensioning a geological HLW waste repository

    International Nuclear Information System (INIS)

    Ouvrier, N.; Chaudon, L.; Malherbe, L.

    1990-01-01

    This method was developed by the CEA to optimize the dimensions of a geological repository by taking account of technical and economic parameters. It involves optimizing radioactive waste storage conditions on the basis of economic criteria with allowance for specified thermal 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. The optimum values found for the two media investigated (granite and salt) show that it is advisable to minimize the interim storage time, implying the containers must be separated by buffer material, whereas vertical spacing may not be required after a 30-year interim storage period. Moreover, the boreholes should be as deep as possible, on a close pitch in widely spaced handling drifts. These results depend to a considerable extent on the assumption of high interim storage costs

  2. 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

    This Test Plan describes work to support the development and testing of high waste loading glass formulations that achieve high glass melting rates for Hanford high aluminum high level waste (HLW). In particular, the present testing is designed to evaluate the effect of using low activity waste (LAW) waste streams as a source of sodium in place ofchemical additives, sugar or cellulose as a reductant, boehmite as an aluminum source, and further enhancements to waste processing rate while meeting all processing and product quality requirements. The work will include preparation and characterization of crucible melts in support of subsequent DuraMelter 100 (DM 100) tests designed to examine the effects of enhanced glass formulations, glass processing temperature, incorporation of the LAW waste stream as a sodium source, type of organic reductant, and feed solids content on waste processing rate and product quality. Also included is a confirmatory test on the HLW Pilot Melter (DM1200) with a composition selected from those tested on the DM100. This work builds on previous work performed at the Vitreous State Laboratory (VSL) for Department of Energy's (DOE's) Office of River Protection (ORP) to increase waste loading and processing rates for high-iron HLW waste streams as well as previous tests conducted for ORP on the same waste composition. This Test Plan is prepared in response to an ORP-supplied statement of work. It is currently estimated that the number of HLW canisters to be produced in the Hanford Tank Waste Treatment and Immobilization Plant (WTP) is about 12,500. This estimate is based upon the inventory ofthe tank wastes, the anticipated performance of the sludge treatment processes, and current understanding of the capability of the borosilicate glass waste form. 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

  3. 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

  4. Selection of Technical Solutions for the Management of Radioactive Waste

    International Nuclear Information System (INIS)

    2017-07-01

    The objectives of this publication are to identify and critically review the criteria to be considered while selecting waste management technologies; summarize, evaluate, rank and compare the different technical solutions; and offer a systematic approach for selecting the best matching solution. This publication covers the management of radioactive waste from all nuclear operations, including waste generated from research reactors, power reactors, and nuclear fuel cycle activities including high level waste (HLW) arising from reprocessing and spent nuclear fuel declared as waste (SFW), as well as low level waste (LLW) and intermediate level waste (ILW) arising from the production and use of radionuclides in industry, agriculture, medicine, education and research.

  5. 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.

  6. 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

  7. 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.

  8. 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

  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. 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

  11. 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

  12. 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

  13. 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

    This Test Plan describes work to support the development and testing of high waste loading glass formulations that achieve high glass melting rates for Hanford high aluminum high level waste (HLW). In particular, the present testing is designed to evaluate the effect of using low activity waste (LAW) waste streams as a source of sodium in place ofchemical additives, sugar or cellulose as a reductant, boehmite as an aluminum source, and further enhancements to waste processing rate while meeting all processing and product quality requirements. The work will include preparation and characterization of crucible melts in support of subsequent DuraMelter 100 (DM 100) tests designed to examine the effects of enhanced glass formulations, glass processing temperature, incorporation of the LAW waste stream as a sodium source, type of organic reductant, and feed solids content on waste processing rate and product quality. Also included is a confirmatory test on the HLW Pilot Melter (DM1200) with a composition selected from those tested on the DM100. This work builds on previous work performed at the Vitreous State Laboratory (VSL) for Department of Energy's (DOE's) Office of River Protection (ORP) to increase waste loading and processing rates for high-iron HLW waste streams as well as previous tests conducted for ORP on the same waste composition. This Test Plan is prepared in response to an ORP-supplied statement of work. It is currently estimated that the number of HLW canisters to be produced in the Hanford Tank Waste Treatment and Immobilization Plant (WTP) is about 12,500. This estimate is based upon the inventory ofthe tank wastes, the anticipated performance of the sludge treatment processes, and current understanding of the capability of the borosilicate glass waste form. 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

  14. An alternative waste form for the final disposal of high-level radioactive waste (HLW) on the basis of a survey of solidification and final disposal of HLW

    International Nuclear Information System (INIS)

    Bauer, C.

    1982-01-01

    The dissertation comprises two separate parts. The first part presents the basic conditions and concepts of the process leading to the development of a waste form, such as:origin, composition and characteristics of the high-level radioactive waste; evaluation of the methods available for the final disposal of radioactive waste, especially the disposal in a geological formation, including the resulting consequences for the conditions of state in the surroundings of the waste package; essential option for the conception of a waste form and presentation of the waste forms developed and examined on an international level up to now. The second part describes the production of a waste form on TiO 2 basis, in which calcined radioactive waste particles in the submillimeter range are embedded in a rutile matrix. That waste form is produced by uniaxial pressure sintering in the temperature range of 1223 K to 1423 K and pressures between 5 MPa and 20 MPa. Microstructure, mechanical properties and leaching rates of the waste form are presented. Moreover, a method is explained allowing compacting of the rutile matrix and also integration of a wasteless overpack of titanium or TiO 2 into the waste form. (orig.) [de

  15. 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

  16. 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

  17. 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

  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. 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

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

    International Nuclear Information System (INIS)

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

    1983-05-01

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

  1. Physical, Chemical and Structural Evolution of Zeolite - Containing Waste Forms Produced from Metakaolinite and Calcined HLW

    International Nuclear Information System (INIS)

    Grutzeck, Michael

    2005-01-01

    During the seventh year of the current grant (DE-FG02-05ER63966) we completed an exhaustive study of cold calcination and began work on the development of tank fill materials to fill empty tanks and control residuals. Cold calcination of low and high NOx low activity waste (LAW) SRS Tank 44 and Hanford AN-107 simulants, respectively with metallic Al + Si powders was evaluated. It was found that a combination of Al and Si powders could be used as reducing agents to reduce the nitrate and nitrite content of both low and high NOx LAW to low enough levels to allow the LAW to be solidified directly by mixing it with metakaolin and allowing it to cure at 90 C. During room temperature reactions, NOx was reduced and nitrogen was emitted as N2 or NH3. This was an important finding because now one can pretreat LAW at ambient temperatures which provides a low-temperature alternative to thermal calcination. The significant advantage of using Al and Si metals for denitration/denitrition of the LAW is the fact that the supernate could potentially be treated in situ in the waste tanks themselves. Tank fill materials based upon a hydroceramic binder have been formulated from mixtures of metakaolinite, Class F fly ash and Class C flue gas desulphurization (FGD) ash mixed with various concentrations of NaOH solution. These harden over a period of hours or days depending on composition. A systematic study of properties of the tank fill materials (leachability) and ability to adsorb and hold residuals is under way

  2. Results of Sludge Mobilization Testing at Hanford High Level Waste (HLW) Tank

    International Nuclear Information System (INIS)

    STAEHR, T.W.

    2001-01-01

    Waste stored in the Tank 241-AZ-101 at the US DOE Hanford is scheduled as the initial feed for high-level waste vitrification. Tank 241-AZ-101 currently holds over 3,000,000 liters of waste made up of a settled sludge layer covered by a layer of liquid supernant. To retrieve the waste from the tank, it is necessary to mobilize and suspend the settled sludge so that the resulting slurry can be pumped from the tank for treatment and vitrification. Two 223.8-kilowatt mixer pumps have been installed in Tank 241-AZ-101 to mobilize the settled sludge layer of waste for retrieval. In May of 2000, the mixer pumps were subjected to a series of tests to determine (1) the extent to which the mixer pumps could mobilize the settle sludge layer of waste, (2) if the mixer pumps could function within operating parameters, and (3) if state-of-the-art monitoring equipment could effectively monitor and quantify the degree of sludge mobilization and suspension. This paper presents the major findings and results of the Tank 241-AZ-101 mixer pump tests, based on analysis of data and waste samples that were collected during the testing. Discussion of the results focuses on the effective cleaning radius achieved and the volume and concentration of sludge mobilized, with both one and two pumps operating in various configurations and speeds. The Tank 241-AZ-101 mixer pump tests were unique in that sludge mobilization parameters were measured using actual waste in an underground storage tank at the hanford Site. The methods and instruments that were used to measure waste mobilization parameters in Tank 241-AZ-101 can be used in other tanks. It can be concluded from the testing that the use of mixer pumps is an effective retrieval method for the mobilization of settled solids in Tank 241-AZ-101

  3. Derived Requirements for Double Shell Tank (DST) High Level Waste (HLW) Auxiliary Solids Mobilization

    Energy Technology Data Exchange (ETDEWEB)

    TEDESCHI, A.R.

    2000-02-28

    The potential need for auxiliary double-shell tank waste mixing and solids mobilization requires an evaluation of optional technologies. This document formalizes those operating and design requirements needed for further engineering evaluations.

  4. Derived Requirements for Double-Shell Tank (DST) High Level Waste (HLW) Auxiliary Solids Mobilization

    International Nuclear Information System (INIS)

    TEDESCHI, A.R.

    2000-01-01

    The potential need for auxiliary double-shell tank waste mixing and solids mobilization requires an evaluation of optional technologies. This document formalizes those operating and design requirements needed for further engineering evaluations

  5. High level waste containing granules coated and embedded in metal as an alternative to HLW glasses

    International Nuclear Information System (INIS)

    Neumann, W.

    1980-01-01

    Simulated high level waste containing granules were overcoated with pyrocarbon or nickel respectively. The coatings were performed by the use of chemical vapour deposition in a fluidized bed. The coated granules were embedded in an aluminium-silicon-alloy to improve the dissipation of radiation induced heat. The metal-granules-composites obtained were of improved product stability related to the high level waste containing glasses. (orig.) [de

  6. Hanford high level waste (HLW) tank mixer pump safe operating envelope reliability assessment

    International Nuclear Information System (INIS)

    Fischer, S.R.; Clark, J.

    1993-01-01

    The US Department of Energy and its contractor, Westinghouse Corp., are responsible for the management and safe storage of waste accumulated from processing defense reactor irradiated fuels for plutonium recovery at the Hanford Site. These wastes, which consist of liquids and precipitated solids, are stored in underground storage tanks pending final disposition. Currently, 23 waste tanks have been placed on a safety watch list because of their potential for generating, storing, and periodically releasing various quantities of hydrogen and other gases. Tank 101-SY in the Hanford SY Tank Farm has been found to release hydrogen concentrations greater than the lower flammable limit (LFL) during periodic gas release events. In the unlikely event that an ignition source is present during a hydrogen release, a hydrogen burn could occur with a potential to release nuclear waste materials. To mitigate the periodic gas releases occurring from Tank 101-SY, a large mixer pump currently is being installed in the tank to promote a sustained release of hydrogen gas to the tank dome space. An extensive safety analysis (SA) effort was undertaken and documented to ensure the safe operation of the mixer pump after it is installed in Tank 101-SY.1 The SA identified a need for detailed operating, alarm, and abort limits to ensure that analyzed safety limits were not exceeded during pump operations

  7. 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

  8. Performance of a buried radioactive high level waste (HLW) glass after 24 years

    International Nuclear Information System (INIS)

    Jantzen, Carol M.; Kaplan, Daniel I.; Bibler, Ned E.; Peeler, David K.; John Plodinec, M.

    2008-01-01

    A radioactive high level waste glass was made in 1980 with Savannah River Site (SRS) Tank 15 waste. This glass was buried in a lysimeter in the SRS burial ground for 24 years. Lysimeter leachate data was available for the first 8 years. The glass was exhumed in 2004. The glass was predicted to be very durable and laboratory tests confirmed this. Scanning electron microscopy of the glass burial surface showed no significant glass alteration consistent with results of other laboratory and field tests. Radionuclide profiling for alpha, beta, and 137 Cs indicated that Pu was not enriched in the soil while 137 Cs and 9 deg. C Sr were enriched in the first few centimeters surrounding the glass. Lysimeter leachate data indicated that 9 deg. C Sr and 137 Cs leaching from the glass was diffusion controlled

  9. Nuclear waste solutions

    Science.gov (United States)

    Walker, Darrel D.; Ebra, Martha A.

    1987-01-01

    High efficiency removal of technetium values from a nuclear waste stream is achieved by addition to the waste stream of a precipitant contributing tetraphenylphosphonium cation, such that a substantial portion of the technetium values are precipitated as an insoluble pertechnetate salt.

  10. 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

  11. Solutions for Waste Management

    International Nuclear Information System (INIS)

    2013-01-01

    To safely and securely dispose of highlevel and long-lived radioactive waste, this material needs to be stored for a period of time that is very long compared to our everyday experience. Underground disposal facilities need to be designed and constructed in suitable geological conditions that can be confidently demonstrated to contain and isolate the hazardous waste from our environment for hundreds of thousands of years. Over this period of time, during which the safety of an underground waste repository system must be assured, the waste's radioactivity will decay to a level that cannot pose a danger to people or the environment. The archaeological record can help in visualizing such a long period of time. Climates change, oceans rise and vanish, and species evolve in the course of a one hundred millennia. Rocks bear witness to all of these changes. Geologists in their search for safe repositories for the long-term disposal of high level radioactive waste have identified rock formations that have proven stable for millions of years. These geological formations are expected to remain stable for millions of years and can serve as host formations for waste repositories.

  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. Evaluation of engineering aspects of backfill placement for high level nuclear waste (HLW) deep geologic repositories

    International Nuclear Information System (INIS)

    Roberds, W.; Kleppe, J.; Gonano, L.

    1984-04-01

    This report includes the identification and subjective evaluation of alternative schemes for backfilling around waste packages and within emplacement rooms. The aspects of backfilling specifically considered in this study include construction and testing; costs have not been considered. However, because construction and testing are simply implementation and verification of design, a design basis for backfill is required. A generic basis has been developed for this study by first identifying qualitative performance objectives for backfill and then weighting each with respect to its potential influence on achieving the repository system performance objectives. Alternative backfill materials and additives have been identified and evaluated with respect to the perceived extent to which each combination can be expected to achieve the backfill design basis. Several distinctly different combinations of materials and additives which are perceived to have the highest potential for achieving the backfill design basis have been selected for further study. These combinations include zeolite/clinoptilolite, bentonite, muck, and muck mixed with bentonite. Feasible alternative construction and testing procedures for each selected combination have been discussed. Recommendations have been made regarding appropriate backfill schemes for hard rock (i.e., basalt at Hanford, Washington, tuff at Nevada Test Site, and generic granite) and salt (i.e., domal salt on the Gulf Coast and generic bedded salt). 27 references, 8 figures, 31 tables

  14. Reuse of hydroponic waste solution.

    Science.gov (United States)

    Kumar, Ramasamy Rajesh; Cho, Jae Young

    2014-01-01

    Attaining sustainable agriculture is a key goal in many parts of the world. The increased environmental awareness and the ongoing attempts to execute agricultural practices that are economically feasible and environmentally safe promote the use of hydroponic cultivation. Hydroponics is a technology for growing plants in nutrient solutions with or without the use of artificial medium to provide mechanical support. Major problems for hydroponic cultivation are higher operational cost and the causing of pollution due to discharge of waste nutrient solution. The nutrient effluent released into the environment can have negative impacts on the surrounding ecosystems as well as the potential to contaminate the groundwater utilized by humans for drinking purposes. The reuse of non-recycled, nutrient-rich hydroponic waste solution for growing plants in greenhouses is the possible way to control environmental pollution. Many researchers have successfully grown several plant species in hydroponic waste solution with high yield. Hence, this review addresses the problems associated with the release of hydroponic waste solution into the environment and possible reuse of hydroponic waste solution as an alternative resource for agriculture development and to control environmental pollution.

  15. Distribution of 14 elements from two solutions simulating Hanford HLW Tank 102-SY (acid-dissolved sludge and acidified supernate) on four cation exchange resins and five anion exchange resins having different functional groups

    International Nuclear Information System (INIS)

    Marsh, S.F.; Svitra, Z.V.; Bowen, S.M.

    1995-01-01

    As part of the Tank Waste Remediation System program at Los Alamos, we evaluated a series of cation exchange and anion exchange resins for their ability to remove hazardous components from radioactive high-level waste (HLW). The anion exchangers were Reillex TM HPQ, a polyvinyl pyridine resin, and four strong-base polystyrene resins having trimethyl, tri ethyl, tri propyl, and tributyl amine as their respective functional groups. The cation exchange resins included Amberlyst TM 15 and Amberlyst tM XN-1010 with sulfonic acid functionality, Duolite TM C-467 with phosphonic acid functionality, and poly functional Diphonix TM with di phosphonic acid, sulfonic acid, and carboxylic acid functionalities. We measured the distributions of 14 elements on these resins from solutions simulating acid-dissolved sludge (pH 0.6) and acidified supernate (pH 3.5) from underground storage tank 102-SY at the Hanford Reservation near Richland, Washington, USA. 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 the 252 element/resin/solution combinations, distribution coefficients (Kds) were measured for dynamic contact periods of 30 minutes, 2 hours, and 6 hours to obtain information about sorption kinetics from these complex media. Because we measured the sorption of many different elements, the tabulated results indicate which unwanted elements are most likely to interfere with the sorption of elements of special interest. On the basis of these 756 measured Kd values, we conclude that some of the tested resins appear suitable for partitioning hazardous components from Hanford HLW. (author). 10 refs., 11 tabs

  16. 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

  17. Post-test evaluations of Waste Isolation Pilot Plant - Savannah River simulated defense HLW canisters and waste form

    International Nuclear Information System (INIS)

    Molecke, M.A.; Sorensen, N.R.; Harbour, J.R.; Ferrara, D.M.

    1993-01-01

    Eighteen nonradioactive defense high-level waste (DHLW) canisters were emplaced in and subjected to accelerated overtest thermal conditions for about three years at the bedded salt Waste Isolation Pilot Plant (WIPP) facility. Post-test laboratory corrosion results of several stainless steel 304L waste canisters, cast steel overpacks, and associated instruments ranged from negligible to moderate. We found appreciable surface corrosion and corrosion products on the cast steel overpacks. Pieces of both 304L and 316 stainless steel test apparatus underwent extensive stress-corrosion cracking failure and nonuniform attack. One of the retrieved test packages contained nonradioactive glass waste form from the Savannah River Site. We conducted post-test analyses of this glass to determine the degree of resultant glass fracturing, and whether any respirable fines were present. Linear glass fracture density ranged from about 1 to 8 fractures intersecting every 5 cm (2 inch) segment along a diameter line of the canister cross-section. Glass fines between 1 and 10 microns in diameter were detected, but were not quantified

  18. 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.

  19. Physical, Chemical and Structural Evolution of Zeolite-Containing Waste Forms Produced from Metakaolinite and Calcined HLW

    International Nuclear Information System (INIS)

    Grutzeck, Michael; Jantzen, Carol M.

    1999-01-01

    Natural and synthetic zeolites are extremely versatile materials. They can adsorb a variety of liquids and gases, and also take part in cation exchange reactions. Zeolites are easy to synthesize from a wide variety of natural and man made materials. One combination of starting materials that exhibits a great deal of promise is a mixture of metakaolinite and/or Class F fly ash and concentrated sodium hydroxide solution. Once these ingredients are mixed and cured at elevated temperatures, they react to form a hard, dense, ceramic-like material that contains significant amounts of crystalline tectosilicates (zeolites and feldspathoids). Zeolites have the ability to sequester ions in lattice positions or within their networks of channels and voids. As such they are nearly perfect waste forms, the zeolites can host alkali, alkaline earth and a variety of higher valance cations. In addition to zeolites, it has been found that the zeolites are accompanied by an alkali aluminosilicate hydrate matrix that is a host, not only to the zeolites, but to residual amounts of insoluble hydroxide phases as well. A previous publication has established the fact that a mixture of a calcined equivalent ICPP waste (sodium aluminate/hydroxide solution containing ∼3:1 Na:Al) and fly ash and/or metakaolinite could be cured at various temperatures to produce a monolith containing Zeolite A (80 C) or Na-P1 plus hydroxy sodalite (130 C) crystals dispersed in an alkali aluminosilicate hydrate matrix. Dissolution tests have shown these materials (so-called hydroceramics) to have superior retention for alkali, alkaline earth and heavy metal ions. The zeolitization process is a simple one. Metakaolinite and/or Class F fly ash is mixed with a caustic sodium-bearing calcine and enough water to make a thick paste. The paste is transferred to a metal canister and ''soaked'' for a few hours at 70-80 C prior to steam autoclaving the sample at ∼200 C for 6-8 hours. The waste form produced in this

  20. 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)

  1. 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

  2. Geological aspects of the nuclear waste disposal problem

    International Nuclear Information System (INIS)

    Laverov, N.P.; Omelianenko, B.L.; Velichkin, V.I.

    1994-06-01

    For the successful solution of the high-level waste (HLW) problem in Russia one must take into account such factors as the existence of the great volume of accumulated HLW, the large size and variety of geological conditions in the country, and the difficult economic conditions. The most efficient method of HLW disposal consists in the maximum use of protective capacities of the geological environment and in using inexpensive natural minerals for engineered barrier construction. In this paper, the principal trends of geological investigation directed toward the solution of HLW disposal are considered. One urgent practical aim is the selection of sites in deep wells in regions where the HLW is now held in temporary storage. The aim of long-term investigations into HLW disposal is to evaluate geological prerequisites for regional HLW repositories

  3. 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)

  4. Progress and future direction for the interim safe storage and disposal of Hanford high level waste (HLW)

    International Nuclear Information System (INIS)

    Wodrich, D.D.

    1996-01-01

    This paper describes the progress made at the largest environmental cleanup program in the United States. Substantial advances in methods to start interim safe storage of Hanford Site high-level wastes, waste characterization to support both safety- and disposal-related information needs, and proceeding with cost-effective disposal by the US DOE and its Hanford Site contractors, have been realized. Challenges facing the Tank Waste Remediation System Program, which is charged with the dual and parallel missions of interim safe storage and disposal of the high-level tank waste stored at the Hanford Site, are described

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

    International Nuclear Information System (INIS)

    Grutzeck, Michael W.

    2003-01-01

    Zeolites can adsorb liquids and gases, take part in catalytic reactions and serve as cation exchange media. They are commercially available as finely divided powders. Using zeolites to manage radioactive waste is not new, but a process by which zeolites can be made to act both as a host phase and a cementing agent is. It is notable that zeolites occur in nature as well consolidated/cemented deposits. The Romans used blocks of Neapolitan zeolitized tuff as a building material and some of these buildings are still standing. Zeolites are easy to synthesize from a wide range of both natural and man-made precursor materials. The method of making a ''hydroceramic'' is derived from a process in which metakaolinite (thermally dehydroxylated kaolinite) is slurried with a dilute sodium hydroxide (NaOH) solution and then reacted for hours to days at mildly elevated temperatures (60-200 C). The zeolites that form in solution are finely divided powders containing micrometer sized crystals. However, if the process is changed and only enough concentrated sodium hydroxide solution (e.g. 12 M) is added to the metakaolinite to give the mixture a putty-like consistency and the mixture is then cured under similar conditions, the mixture becomes a very hard ceramic-like material containing distinct tectosilicate crystallites (zeolites and feldspathoids) imbedded in an X-ray amorphous sodium aluminosilicate hydrate matrix. Due to the material's vitreous character, the composite has been called a hydroceramic. Similar to zeolite/feldspathoid powders, a hydroceramic is able to sequester cations and a wide range of salt molecules (e.g., nitrate, nitrite and sulfate) in lattice positions and within structural channels and voids thus rendering them ''insoluble'' and making them an ideal contingency waste form for solidifying radioactive waste. The obvious similarities between a hydroceramic waste form and a waste form based on solidified Portland-cement grout are superficial because their

  6. 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)

  7. PHYSICAL, CHEMICAL, AND STRUCTURAL EVOLUTION OF ZEOLITE-CONTAINING WASTE FORMS PRODUCED FROM METAKAOLINITE AND CALCINED HLW

    International Nuclear Information System (INIS)

    Pareizs, J. M.; Jantzenm, C.M.

    2000-01-01

    Natural and synthetic zeolites are extremely versatile materials. They can adsorb a variety of liquids and gases, and also take part in cation exchange reactions. Zeolites have the ability to sequester ions in lattice positions or within their networks of channels and voids. The zeolites can host alkali, alkaline earth and a variety of higher valance cations. As such they may be a viable alternative for immobilization of low activity waste (LAW) salts and calcines. The process for synthesizing zeolites is well documented for pure starting materials. A reactive aluminosilicate is reacted with an alkaline hydroxide at low temperature (<300 C) to form a zeolite. Processing time and temperature and specific reactants determine the type of zeolite formed. Zeolites are easy to make, and can be synthesized from a wide variety of natural and man made materials. However, relatively little is known about the process if one of the starting materials is a poorly characterized complex mixture of oxides (waste) containing nearly every element in the periodic table. The purpose of this work is to develop a clearer understanding of the advantages and limitations of producing a zeolite waste form from radioactive waste. Dr. M. W. Grutzeck at the Pennsylvania State University is investigating the production of a zeolite waste form using nonradioactive simulants. Dr. C. M. Jantzen and J. M. Pareizs at the Savannah River Technology Center will use the results from simulant work as a starting point for producing a zeolite waste form from an actual Savannah River Site radioactive waste stream

  8. Physical, chemical, and structural evolution of zeolite-containing waste forms produced from metakaolinite and calcined HLW

    International Nuclear Information System (INIS)

    Pareizs, J.M.

    2000-01-01

    Natural and synthetic zeolites are extremely versatile materials. They can adsorb a variety of liquids and gases, and also take part in cation exchange reactions. Zeolites have the ability to sequester ions in lattice positions or within their networks of channels and voids. The zeolites can host alkali, alkaline earth and a variety of higher valence cations. As such they may be a viable alternative for immobilization of low activity waste (LAW) salts and calcines. The process for synthesizing zeolites is well documented for pure starting materials. A reactive aluminosilicate is reacted with an alkaline hydroxide at low temperature to form a zeolite. Processing time and temperature and specific reactants determine the type of zeolite formed. Zeolites are easy to make, and can be synthesized from a wide variety of natural and man made materials. However, relatively little is known about the process if one of the starting materials is a poorly characterized complex mixture of oxides (waste) containing nearly every element in the periodic table. The purpose of this work is to develop a clearer understanding of the advantages and limitations of producing a zeolite waste form from radioactive waste. Dr. M. W. Grutzeck at the Pennsylvania State University is investigating the production of a zeolite waste form using non-radioactive simulants. Dr. C. M. Jantzen and J. M. Pareizs at the Savannah River Technology Center will use the results from simulant work as a starting point for producing a zeolite waste form from an actual Savannah River Site radioactive waste stream

  9. 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.

  10. 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'

  11. 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)

  12. Immobilized High-Level Waste (HLW) Interim Storage Alternative Generation and analysis and Decision Report - second Generation Implementing Architecture

    International Nuclear Information System (INIS)

    CALMUS, R.B.

    2000-01-01

    Two alternative approaches were previously identified to provide second-generation interim storage of Immobilized High-Level Waste (IHLW). One approach was retrofit modification of the Fuel and Materials Examination Facility (FMEF) to accommodate IHLW. The results of the evaluation of the FMEF as the second-generation IHLW interim storage facility and subsequent decision process are provided in this document

  13. Geologic disposal as optimal solution of managing the spent nuclear fuel and high-level radioactive waste

    International Nuclear Information System (INIS)

    Ilie, P.; Didita, L.; Ionescu, A.; Deaconu, V.

    2002-01-01

    To date there exist three alternatives for the concept of geological disposal: 1. storing the high-level waste (HLW) and spent nuclear fuel (SNF) on ground repositories; 2. solutions implying advanced separation processes including partitioning and transmutation (P and T) and eventual disposal in outer space; 3. geological disposal in repositories excavated in rocks. Ground storing seems to be advantageous as it ensures a secure sustainable storing system over many centuries (about 300 years). On the other hand ground storing would be only a postponement in decision making and will be eventually followed by geological disposal. Research in the P and T field is expected to entail a significant reduction of the amount of long-lived radioactive waste although the long term geological disposal will be not eliminated. Having in view the high cost, as well as the diversity of conditions in the countries owning power reactors it appears as a reasonable regional solution of HLW disposal that of sharing a common geological disposal. In Romania legislation concerning of radioactive waste is based on the Law concerning Spent Nuclear Fuel and Radioactive Waste Management in View of Final Disposal. One admits at present that for Romania geological disposal is not yet a stressing issue and hence intermediate ground storing of SNF will allow time for finding a better final solution

  14. 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

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

    International Nuclear Information System (INIS)

    Grutzeck, Michael W.

    2004-01-01

    Zeolites are extremely versatile. They can adsorb liquids and gases and serve as cation exchange media. They occur in nature as well cemented deposits. The Romans used blocks of zeolitized tuff as a building material. Using zeolites for the management of radioactive waste is not new, but a process by which the zeolites can be made to act as a cementing agent is. Zeolitic materials are relatively easy to synthesize from a wide range of both natural and man-made precursors. The process under study is derived from a well known method in which metakaolin (thermally dehydroxylated kaolin a mixture of kaolinite and smaller amounts of quartz and mica that has been heated to ∼700 C) is mixed with sodium hydroxide (NaOH) and water and reacted in slurry form (for a day or two) at mildly elevated temperatures. The zeolites form as finely divided powders containing micrometer ((micro)m) sized crystals. However, if the process is changed slightly and just enough concentrated sodium hydroxide solution is added to the metakaolinite to make a thick paste and then the paste is cured under mild hydrothermal conditions (60-200 C), the mixture forms a concrete-like ceramic material made up of distinct crystalline tectosilicate minerals (zeolites and feldspathoids) imbedded in an X-ray amorphous hydrated sodium aluminosilicate matrix. Due to its vitreous character we have chosen to call this composite a ''hydroceramic''. Similar to zeolite powders, a hydroceramic is able to sequester cations in both lattice positions and within the channels and voids present in its tectosilicate framework structure. It can also accommodate a wide range of salt molecules (e.g., sodium nitrate) within these same openings thus rendering them insoluble. Due to its fine crystallite size and cementing character, the matrix develops significant physical strength. The obvious similarities between a hydroceramic waste form and a waste form based on solidified Portland cement grout are only superficial because

  16. Prediction of geological and mechanical processes while disposing of high-level waste (HLW) into the earth crust

    International Nuclear Information System (INIS)

    Kedrovsky, O.L.; Morozov, V.N.

    1992-01-01

    Prediction of geological and mechanical processes while disposing of high-level waste of atomic industry into the earth crust is the fundamental base for ecological risk assessment (possible consequences) while developing repository designs. The subject of this paper is the analytical estimate of possibilities of rock fracturing mechanisms to predict isolation properties loss by massif beginning from crystal lattice of minerals up to large fracture disturbances under conditions of long-term influence of pressure, temperature, and radiation. To solve the problem possibilities of kinetic

  17. Conditioning of radioactive waste solutions by cementation

    International Nuclear Information System (INIS)

    Vejmelka, P.; Rudolph, G.; Kluger, W.; Koester, R.

    1992-02-01

    For the cementation of the low and intermediate level evaporator concentrates resulting from the reprocessing of spent fuel numerous experiments were performed to optimize the waste form composition and to characterize the final waste form. Concerning the cementation process, properties of the waste/cement suspension were investigated. These investigations include the dependence of viscosity, bleeding, setting time and hydration heat from the waste cement slurry composition. For the characterization of the waste forms, the mechanical, thermal and chemical stability were determined. For special cases detailed investigations were performed to determine the activity release from waste packages under defined mechanical and thermal stresses. The investigations of the interaction of the waste forms with aqueous solutions include the determination of the Cs/Sr release, the corrosion resistance and the release of actinides. The Cs/Sr release was determined in dependence of the cement type, additives, setting time and sample size. (orig./DG) [de

  18. 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

  19. Waste processing of chemical cleaning solutions

    International Nuclear Information System (INIS)

    Peters, G.A.

    1991-01-01

    This paper reports on chemical cleaning solutions containing high concentrations of organic chelating wastes that are difficult to reduce in volume using existing technology. Current methods for evaporating low-level radiative waste solutions often use high maintenance evaporators that can be costly and inefficient. The heat transfer surfaces of these evaporators are easily fouled, and their maintenance requires a significant labor investment. To address the volume reduction of spent, low-level radioactive, chelating-based chemical cleaning solutions, ECOSAFE Liquid Volume Reduction System (LVRS) has been developed. The LVRS is based on submerged combustion evaporator technology that was modified for treatment of low-level radiative liquid wastes. This system was developed in 1988 and was used to process 180,000 gallons of waste at Oconee Nuclear Station

  20. Thermal analysis in the near field for geological disposal of high-level radioactive waste. Establishment of the disposal tunnel spacing and waste package pitch on the 2nd progress report for the geological disposal of HLW in Japan

    International Nuclear Information System (INIS)

    Taniguchi, Wataru; Iwasa, Kengo

    1999-11-01

    For the underground facility of the geological disposal of high-level radioactive waste (HLW), the space is needed to set the engineered barrier, and the set engineered barrier and rock-mass of near field are needed to satisfy some conditions or constraints for their performance. One of the conditions above mentioned is thermal condition arising from heat outputs of vitrified waste and initial temperature at the disposal depth. Hence, it is needed that the temperature of the engineered barrier and rock mass is less degree than the constraint temperature of each other. Therefore, the design of engineered barrier and underground facility is conducted so that the temperature of the engineered barrier and rock mass is less degree than the constraint temperature of each other. One of these design is establishment of the disposal tunnel spacing and waste package pitch. In this report, thermal analysis is conducted to establish the disposal tunnel spacing and waste package pitch to satisfy the constraint temperature in the near field. Also, other conditions or constraints for establishment of the disposal tunnel spacing and waste package pitch are investigated. Then, design of the disposal tunnel spacing and waste package pitch, considering these conditions or constraints, is conducted. For the near field configuration using the results of the design above mentioned, the temperature with time dependency is studied by analysis, and then the temperature variation due to the gaps, that will occur within the engineered barrier and between the engineered barrier and rock mass in setting engineered barrier in the disposal tunnel or pit, is studied. At last, the disposal depth variation is studied to satisfy the temperature constraint in the near field. (author)

  1. Physical, Chemical and Structural Evolution of Zeolite-Containing Waste Forms Produced from Metakaolinite and Calcined Sodium Bearing Waste (HLW and/or LLW)

    International Nuclear Information System (INIS)

    Grutzeck, Michael W.

    2005-01-01

    Zeolites are extremely versatile. They can adsorb liquids and gases and serve as cation exchange media. They occur in nature as well cemented deposits. The ancient Romans used blocks of zeolitized tuff as a building material. Using zeolites for the management of radioactive waste is not a new idea, but a process by which the zeolites can be made to act as a cementing agent is. Zeolitic materials are relatively easy to synthesize from a wide range of both natural and man-made substances. The process under study is derived from a well known method in which metakaolin (an impure thermally dehydroxylated kaolinite heated to ∼700 C containing traces of quartz and mica) is mixed with sodium hydroxide (NaOH) and reacted in slurry form (for a day or two) at mildly elevated temperatures. The zeolites form as finely divided powders containing micrometer ((micro)m) sized crystals. However, if the process is changed slightly and only just enough concentrated sodium hydroxide solution is added to the metakaolinite to make a thick crumbly paste and then the paste is compacted and cured under mild hydrothermal conditions (60-200 C), the mixture will form a hard ceramic-like material containing distinct crystalline tectosilicate minerals (zeolites and feldspathoids) imbedded in an X-ray amorphous hydrated sodium aluminosilicate matrix. Due to its lack of porosity and vitreous appearance we have chosen to call this composite a ''hydroceramic''

  2. HLW disposal dilemma

    International Nuclear Information System (INIS)

    Andrei, V.; Glodeanu, F.

    2003-01-01

    ' 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)

  3. Radioactive wastes. The groundwork of current solutions

    International Nuclear Information System (INIS)

    Grevoz, A.; Boullis, B.; Devezeaux de Lavergne, J.G.; Butez, M.; Bordier, G.; Vitart, X.; Hablot, I.; Chastagnet, F.

    2005-01-01

    Today the groundwork laid down by research has made processes available for the durable treatment and conditioning of all types of radioactive waste. This document illustrates the today situations in five presentations. Now standing as a national reference, the french inventory of radioactive waste, drawn up by ANDRA, has not only expanded to cover recoverable material but also features predictions of waste arisings for 2010 and 2020, including waste from the decommissioning of current installations. The current process used for spent fuel reprocessing allows extraction for recycling purpose, of uranium and plutonium, with very high recovery and purification rates. Advances in characterization and decontamination allow improvements in sorting and retrieval and conditioning to be considered for older wastes. The french National radioactive waste management agency (ANDRA) is already providing optimum industrial solutions for all short-lived, low and very low level waste on its Soulaines and Morvillers sites. For several decades, Areva has been reprocessing spent fuel and conditioning ultimate waste in its La Hague plants. (A.L.B.)

  4. 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

  5. Radioactive waste management turning options into solution

    International Nuclear Information System (INIS)

    Neubauer, J.

    2000-10-01

    Most of the statements from representatives of different countries and institutions focused on the status of high level radioactive waste management, including spent fuel repositories. Speakers dealing with such topics were representatives from countries applying nuclear power for electricity production. They all reported about there national programs on technical and safety aspects of radioactive waste management. The panel discussion extended to questions on political sensitivities and public acceptance; in this respect, interesting developments are taking place in Finland and Sweden. It is expected that Finland will operate a final repository for spent fuel in 10 - 15 years from now, followed close by Sweden. Other countries, however, face decisions by policy makers and elected officials to postpone dealing with waste disposal concerns. In this connection there is relevant experience in our country, too - even in the absence of spent fuel or other high level waste to be dealt with. During personal discussions with representatives of other countries not using nuclear power it was confirmed that there are similar or shared experiences. Development of publicly -accepted solutions to radioactive waste management remains an important issue. Independent of the amount or the activity of radioactive waste, the public at large remains skeptical despite the agreement among experts that disposal can be safe, technically feasible and environmentally sound. In countries not using nuclear power there are only small quantities of low and intermediate level radioactive waste. Therefore, international co-operation among such countries should be an option. There was common understanding by representatives from Norway, Italy and Austria that international co-operation should be developed for treatment and disposal of such waste. For the moment however it has to be accepted that, for political reasons, it is not possible. Forced to deal with the lack of near-term solutions, the

  6. 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.'

  7. Specific transport and storage solutions: Waste management facing current and future stakes of the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Deniau, Helene; Gagner, Laurent; Gendreau, Francoise; Presta, Anne

    2006-01-01

    development process; - Current solutions proposed by COGEMA LOGISTICS; - Transport of Low Level and Very Low Level waste (LLW and VLLW); - Shipment of large volumes; - Non-reusable transport packaging; - Packaging transport permit the storage in final disposal area; - A steep increase of transport flows in the next 10 years; - Future challenges, waste from dismantling operations; - Transport packaging of IP 2 type and IP 2 package; - Some examples of standardized packages; - DV78 Package; - CC102 over pack; - A solution for liquid waste from NPPs, TNTMCIEL; - Transport of Intermediate Level Waste (ILW) and High Level Waste (HLW); - Alpha waste (or TRU waste); - TNTMGEMINI; - RD 26 package; - Bituminized waste (ILW): the TNTM833; - Technological waste in CBFC'2 (ILW): the TNTM837; - Universal Canisters for Vitrified waste (HLW) and Compacted waste (ILW); - Vitrified waste (HLW); - Transport solutions for CSD-V canisters: the TNTM28 VT; - Dual purpose solutions for CSD-V canisters, the TS 28; - CSD-V programme of return; - Compacted waste (ILW): the TNTM843; - Some future stakes and new developments. The following conclusions completes the paper. COGEMA LOGISTICS experience in designing packages and transporting nuclear materials is part of the COGEMA / AREVA waste management policy, which aims to waste volume reduction through waste sorting and packaging in universal canisters or standard waste containers. For the design and manufacture of special shipping casks, as for nuclear transportation and storage, COGEMA LOGISTICS tailors its solutions to its international customers' requirements. Personnel safety, transportation safety and environmental protection are put first. These reasons have lead COGEMA LOGISTICS to become the world's leading designer of nuclear fuel packages, with unique expertise in designing packages for back-end transportation and storage of nuclear materials. Beyond being a transport company and packaging designer, COGEMA LOGISTICS is a partner implementing

  8. 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

  9. 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

  10. Processing of waste solutions from electrochemical decontamination

    International Nuclear Information System (INIS)

    Charlot, L.A.; Allen, R.P.; Arrowsmith, H.W.; Hooper, J.L.

    1979-09-01

    The use of electropolishing as a decontamination technique will be effective only if we can minimize the amount of secondary waste requiring disposal and economically recycle part of the decontamination electrolyte. Consequently, a solution purification method is needed to remove the dissolved contamination and metal in the electrolyte. This report describes the selection of a purification method for a phosphoric acid electrolyte from the following possible acid reclamation processes: ion exchange, solvent extraction, precipitation, distillation, electrolysis, and membrane separation

  11. MIIT: International in-situ testing of simulated HLW forms - performance of SRS simulated waste glass after 6 mos., 1 yr., 2 yrs. and 5 yrs. of burial at WIPP

    International Nuclear Information System (INIS)

    Wicks, G.G.; Lodding, A.R.; Macedo, P.B.; Clark, D.E.

    1991-01-01

    The first field test, involving burial of simulated high-level waste (HLW) forms and package components, to be conducted in the United States, was begun in July of 1986. This program, called the Materials Interface Interactions Test or MIIT, comprises the largest 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 7 countries. Also included are about 300 potential canister or overpack metal samples along with more than 500 geologic and backfill specimens. There are almost 2000 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. The MIIT program represents a joint effort managed by Sandia National Laboratories in Albuquerque, N.M., and Savannah River Laboratory in Aiken, S.C. and sponsored by the US Department of Energy. Also involved in MIIT are participants from various laboratories and universities in France, Germany, Belgium, Canada, Japan, Sweden, the United Kingdom, and the United States. In July of 1991, the experimental portion of the 5-yr. MIIT program was completed. Although only about 5% of all MIIT samples have been assessed thus far, there are already interesting findings that have emerged. The present paper will discuss results obtained for SRS 165/TDS waste glass after burial of 6 mo., 1 yr. and 2 yrs., along with initial analyses of 5 yr. samples

  12. Characterization of high level waste for minor actinides by chemical separation and alpha spectrometry

    International Nuclear Information System (INIS)

    Murali, M.S.; Bhattacharayya, A.; Kar, A.S.; Tomar, B.S.; Manchanda, V.K.

    2010-01-01

    Quantification of minor actinides present in of High Level Waste (HLW) solutions originating from the power reactors is important in view of management of radioactive wastes and actinide partitioning. Several methods such as ICP-MS, X-ray fluorescence methods, ICP-AES, alpha spectrometry are used in characterizing such types of wastes. As alpha spectrometry is simple and reliable, this technique has been used for the estimation of minor actinides after devising steps of separation for estimating Np and Pu present in HLW solutions of PHWR origin. Using a wealth of knowledge appropriate to the solution chemistry of actinides, the task of separation, though appears easy, it is challenging job for a radiochemist handling high-dose HLW samples, for obtaining clean alpha peaks for Np and Pu. This paper reports on the successful attempt made to quantify 241 Am, 244 Cm, Pu (239 mainly) and 237 Np present in HLW-PHWR obtained from PREFRE, Tarapur

  13. Recovery of uranium from analytical waste solution

    International Nuclear Information System (INIS)

    Kumar, Pradeep; Anitha, M.; Singh, D.K.

    2016-01-01

    Dispersion fuels are considered as advance fuel for the nuclear reactor. Liquid waste containing significant quantity of uranium gets generated during chemical characterization of dispersion fuel. The present paper highlights the effort in devising a counter current solvent extraction process based on the synergistic mixture of D2EHPA and Cyanex 923 to recover uranium from such waste solutions. A typical analytical waste solution was found to have the following composition: U 3 O 8 (∼3 g/L), Al: 0.3 g/L, V: 15 ppm, Phosphoric acid: 3M, sulphuric acid : 1M and nitric acid : 1M. The aqueous solution is composed of mixture of either 3M phosphoric acid and 1M sulphuric acid or 1M sulphuric acid and 1M nitric acid, keeping metallic concentrations in the above mentioned range. Different organic solvents were tested. Based on the higher extraction of uranium with synergistic mixture of 0.5M D2EHPA + 0.125M Cyanex 923, it was selected for further investigation in the present work

  14. Treatment for hydrazine-containing waste water solution

    Science.gov (United States)

    Yade, N.

    1986-01-01

    The treatment for waste solutions containing hydrazine is presented. The invention attempts oxidation and decomposition of hydrazine in waste water in a simple and effective processing. The method adds activated charcoal to waste solutions containing hydrazine while maintaining a pH value higher than 8, and adding iron salts if necessary. Then, the solution is aerated.

  15. 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)

  16. Nuclear waste disposal: Can there be a resolution? Past problems and future solutions

    Energy Technology Data Exchange (ETDEWEB)

    Ahearne, J [Scientific Research Society, Sigma Xi, Research Triangle Park, NC (United States)

    1990-07-01

    Why does the high level waste problem have to be solved now? There are perhaps three answers to that question. First, to have a recovery of nuclear power. But a lack of resolution of the high level waste problem is not the principal reason that nuclear power has foundered and, consequently, solving it will not automatically revive nuclear power. However, if the nuclear industry is adamantly convinced that this is the key to reviving nuclear power, then the nuclear industry should demonstrate its conviction by putting much greater effort into resolving the high level waste problem technically, not through public relations. For example, a substantial effort on the actinide burning approach might demonstrate, in the old American phrase, 'putting your money where your mouth is'. Second, the high level waste problem must be solved now because it is a devil's brew. However, chemical wastes last longer, as we all know, than do the radioactive wastes. As one expert has noted: 'There is real risk in nuclear power, just as there is real risk in coal power.... For some of [these risks], like the greenhouse effect, the potential damage is devastating. While for others, like nuclear accidents, the risk is limited, but imaginations are not. For still others, like the risk posed by a high-level waste repository, there is essentially nothing outside the imagination of the gullible.' Furthermore, any technical solution or any solution to a risky problem requires one to think carefully. It is often better to do it right than quickly. A third reason for requiring it to be solved right now is that HLW disposal is a major technical problem blocking a potentially valuable energy source. But we need a new solution. The current solutions are not working. I believe that we ought to recognize the failure of the geologic repository approach. I believe the federal government should identify, with industry's assistance, the best techniques for surface storage. Some federal locations should be

  17. Nuclear waste disposal: Can there be a resolution? Past problems and future solutions

    International Nuclear Information System (INIS)

    Ahearne, J.

    1990-01-01

    Why does the high level waste problem have to be solved now? There are perhaps three answers to that question. First, to have a recovery of nuclear power. But a lack of resolution of the high level waste problem is not the principal reason that nuclear power has foundered and, consequently, solving it will not automatically revive nuclear power. However, if the nuclear industry is adamantly convinced that this is the key to reviving nuclear power, then the nuclear industry should demonstrate its conviction by putting much greater effort into resolving the high level waste problem technically, not through public relations. For example, a substantial effort on the actinide burning approach might demonstrate, in the old American phrase, 'putting your money where your mouth is'. Second, the high level waste problem must be solved now because it is a devil's brew. However, chemical wastes last longer, as we all know, than do the radioactive wastes. As one expert has noted: 'There is real risk in nuclear power, just as there is real risk in coal power.... For some of [these risks], like the greenhouse effect, the potential damage is devastating. While for others, like nuclear accidents, the risk is limited, but imaginations are not. For still others, like the risk posed by a high-level waste repository, there is essentially nothing outside the imagination of the gullible.' Furthermore, any technical solution or any solution to a risky problem requires one to think carefully. It is often better to do it right than quickly. A third reason for requiring it to be solved right now is that HLW disposal is a major technical problem blocking a potentially valuable energy source. But we need a new solution. The current solutions are not working. I believe that we ought to recognize the failure of the geologic repository approach. I believe the federal government should identify, with industry's assistance, the best techniques for surface storage. Some federal locations should be

  18. 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

  19. Scientific Solutions to Nuclear Waste Environmental Challenges

    International Nuclear Information System (INIS)

    Johnson, Bradley R.

    2014-01-01

    The Hidden Cost of Nuclear Weapons The Cold War arms race drove an intense plutonium production program in the U.S. This campaign produced approximately 100 tons of plutonium over 40 years. The epicenter of plutonium production in the United States was the Hanford site, a 586 square mile reservation owned by the Department of Energy and located on the Colombia River in Southeastern Washington. Plutonium synthesis relied on nuclear reactors to convert uranium to plutonium within the reactor fuel rods. After a sufficient amount of conversion occurred, the rods were removed from the reactor and allowed to cool. They were then dissolved in an acid bath and chemically processed to separate and purify plutonium from the rest of the constituents in the used reactor fuel. The acidic waste was then neutralized using sodium hydroxide and the resulting mixture of liquids and precipitates (small insoluble particles) was stored in huge underground waste tanks. The byproducts of the U.S. plutonium production campaign include over 53 million gallons of high-level radioactive waste stored in 177 large underground tanks at Hanford and another 34 million gallons stored at the Savannah River Site in South Carolina. This legacy nuclear waste represents one of the largest environmental clean-up challenges facing the world today. The nuclear waste in the Hanford tanks is a mixture of liquids and precipitates that have settled into sludge. Some of these tanks are now over 60 years old and a small number of them are leaking radioactive waste into the ground and contaminating the environment. The solution to this nuclear waste challenge is to convert the mixture of solids and liquids into a durable material that won't disperse into the environment and create hazards to the biosphere. What makes this difficult is the fact that the radioactive half-lives of some of the radionuclides in the waste are thousands to millions of years long. (The half-life of a radioactive substance is the amount

  20. 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.

  1. Radioactive waste management

    International Nuclear Information System (INIS)

    Tsoulfanidis, N.

    1991-01-01

    The management of radioactive waste is a very important part of the nuclear industry. The future of the nuclear power industry depends to a large extent on the successful solution of the perceived or real problems associated with the disposal of both low-level waste (LLW) and high-level waste (HLW). All the activities surrounding the management of radioactive waste are reviewed. The federal government and the individual states are working toward the implementation of the Nuclear Waste Policy Act and the Low-Level Waste Policy Act. The two congressional acts are reviewed and progress made as of early 1990 is presented. Spent-fuel storage and transportation are discussed in detail as are the concepts of repositories for HLW. The status of state compacts for LLW is also discussed. Finally, activities related to the decommissioning of nuclear facilities are also described

  2. 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.

  3. PAIRWISE BLENDING OF HIGH LEVEL WASTE

    International Nuclear Information System (INIS)

    CERTA, P.J.

    2006-01-01

    The primary objective of this study is to demonstrate a mission scenario that uses pairwise and incidental blending of high level waste (HLW) to reduce the total mass of HLW glass. Secondary objectives include understanding how recent refinements to the tank waste inventory and solubility assumptions affect the mass of HLW glass and how logistical constraints may affect the efficacy of HLW blending

  4. 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

  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. 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.'

  7. 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).

  8. 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)

  9. 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

  10. Nuclear waste management. Pioneering solutions from Finland

    International Nuclear Information System (INIS)

    Rasilainen, Kari

    2016-01-01

    Presentation outline: Background: Nuclear energy in Finland; Nuclear Waste Management (NWM) Experiences; Low and Intermediate Level Waste (LILW); High Level Waste - Deep Geological Repository (DGR); NWM cost estimate in Finland; Conclusions: World-leading expert services

  11. 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)

  12. 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

  13. 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

  14. 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.

  15. 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.

  16. The different solutions for the waste storage

    International Nuclear Information System (INIS)

    Fillion, E.

    2001-01-01

    Created in 1979, the National agency for the management of radioactive waste (A.N.D.R.A.) is a public establishment in charge of the management of radioactive waste produced in France. It is independent from waste producers and watches over the long term protection of man and his environment, at any step of radioactive waste management. It has for mission to check the waste quality and to conceive, to establish, to build and to manage storage centers where waste are stored according their characteristics. (N.C.)

  17. Carbon Market and Integrated Waste Solutions : a Case Study of ...

    International Development Research Centre (IDRC) Digital Library (Canada)

    Carbon Market and Integrated Waste Solutions : a Case Study of Indonesia ... dual purpose of helping developing countries achieve sustainable development ... with a view to devising integrated waste management solutions in urban centres ... and disseminate them through national, regional and international networks.

  18. Treatment of organic waste solutions containing tributyl phosphate

    International Nuclear Information System (INIS)

    Drobnik, S.

    The two processes developed in the laboratory for treating waste solutions containing TBP, namely TBP separation with phosphoric acid and saponification were tested on a semi-industrial scale. A waste solution from the first phase of the Karlsruhe reprocessing plant was used

  19. Energy from waste: a wholly acceptable waste-management solution

    International Nuclear Information System (INIS)

    Porteous, A.

    1997-01-01

    This paper briefly reviews the 'waste management hierarchy' and why it should be treated as a checklist and not a piece of unquestioning dogma. The role of energy from waste (EfW) is examined in depth to show that it is a rigorous and environmentally sound waste-management option which complements other components of the waste-management hierarchy and assists resource conservation. (Copyright (c) 1997 Elsevier Science B.V., Amsterdam. All rights reserved.)

  20. Radioactive waste management - a safe solution

    International Nuclear Information System (INIS)

    1993-01-01

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

  1. Low-level waste management - suggested solutions for problem wastes

    International Nuclear Information System (INIS)

    Pechin, W.H.; Armstrong, K.M.; Colombo, P.

    1984-01-01

    Problem wastes are those wastes which are difficult or require unusual expense to place into a waste form acceptable under the requirements of 10 CFR 61 or the disposal site operators. Brookhaven National Laboratory has been investigating the use of various solidification agents as part of the DOE Low-Level Waste Management Program for several years. Two of the leading problem wastes are ion exchange resins and organic liquids. Ion exchange resins can be solidified in Portland cement up to about 25 wt % resin, but waste forms loaded to this degree exhibit significantly reduced compressive strength and may disintegrate when immersed in water. Ion exchange resins can also be incorporated into organic agents. Mound Laboratory has been investigating the use of a joule-heated glass melter as a means of disposing of ion exchange resins and organic liquids in addition to other combustible wastes

  2. Waste management outlook for mountain regions: Sources and solutions.

    Science.gov (United States)

    Semernya, Larisa; Ramola, Aditi; Alfthan, Björn; Giacovelli, Claudia

    2017-09-01

    Following the release of the global waste management outlook in 2015, the United Nations Environment Programme (UN Environment), through its International Environmental Technology Centre, is elaborating a series of region-specific and thematic waste management outlooks that provide policy recommendations and solutions based on current practices in developing and developed countries. The Waste Management Outlook for Mountain Regions is the first report in this series. Mountain regions present unique challenges to waste management; while remoteness is often associated with costly and difficult transport of waste, the potential impact of waste pollutants is higher owing to the steep terrain and rivers transporting waste downstream. The Outlook shows that waste management in mountain regions is a cross-sectoral issue of global concern that deserves immediate attention. Noting that there is no 'one solution fits all', there is a need for a more landscape-type specific and regional research on waste management, the enhancement of policy and regulatory frameworks, and increased stakeholder engagement and awareness to achieve sustainable waste management in mountain areas. This short communication provides an overview of the key findings of the Outlook and highlights aspects that need further research. These are grouped per source of waste: Mountain communities, tourism, and mining. Issues such as waste crime, plastic pollution, and the linkages between exposure to natural disasters and waste are also presented.

  3. 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

  4. Electrochemical processing of nitrate waste solutions

    International Nuclear Information System (INIS)

    Genders, D.; Weinberg, N.; Hartsough, D.

    1992-01-01

    The second phase of research performed at The Electrosynthesis Co., Inc. has demonstrated the successful removal of nitrite and nitrate from a synthetic effluent stream via a direct electrochemical reduction at a cathode. It was shown that direct reduction occurs at good current efficiencies in 1,000 hour studies. The membrane separation process is not readily achievable for the removal of nitrites and nitrates due to poor current efficiencies and membrane stability problems. A direct reduction process was studied at various cathode materials in a flow cell using the complete synthetic mix. Lead was found to be the cathode material of choice, displaying good current efficiencies and stability in short and long term tests under conditions of high temperature and high current density. Several anode materials were studied in both undivided and divided cell configurations. A divided cell configuration was preferable because it would prevent re-oxidation of nitrite by the anode. The technical objective of eliminating electrode fouling and solids formation was achieved although anode materials which had demonstrated good stability in short term divided cell tests corroded in 1,000 hour experiments. The cause for corrosion is thought to be F - ions from the synthetic mix migrating across the cation exchange membrane and forming HF in the acid anolyte. Other possibilities for anode materials were explored. A membrane separation process was investigated which employs an anion and cation exchange membrane to remove nitrite and nitrate, recovering caustic and nitric acid. Present research has shown poor current efficiencies for nitrite and nitrate transport across the anion exchange membrane due to co-migration of hydroxide anions. Precipitates form within the anion exchange membranes which would eventually result in the failure of the membranes. Electrochemical processing offers a highly promising and viable method for the treatment of nitrate waste solutions

  5. The best solution to our Nation's waste management problem: Education

    International Nuclear Information System (INIS)

    Mikel, C.J.

    1992-01-01

    In addition to the Waste Isolation Pilot Plant (WIPP) being the best solution today to the Nation's problem of permanent storage of transuranic radioactive waste produced by the defense industry, WIPP is also involved in finding the solution for another national problem: the education of our youth. The youth of America have grown up thinking that science and math are too hard, or not interesting. We, the parents of our Nation's leaders of tomorrow, must find a solution to this dilemma. It is the mission of the Waste Isolation Division Educational Programs to create programs to promote quality education in the classroom and to enhance each student's interest in mathematics and the sciences

  6. 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

  7. 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

  8. Importance of waste composition for Life Cycle Assessment of waste management solutions

    DEFF Research Database (Denmark)

    Bisinella, Valentina; Götze, Ramona; Conradsen, Knut

    2017-01-01

    The composition of waste materials has fundamental influence on environmental emissions associated with waste treatment, recycling and disposal, and may play an important role also for the Life Cycle Assessment (LCA) of waste management solutions. However, very few assessments include effects...... of the waste composition and waste LCAs often rely on poorly justified data from secondary sources. This study systematically quantifiesy the influence and uncertainty on LCA results associated with selection of waste composition data. Three archetypal waste management scenarios were modelled with the waste...... LCA model EASETECH based on detailed waste composition data from the literature. The influence from waste composition data on the LCA results was quantified with a step-wise Global Sensitivity Analysis (GSA) approach involving contribution, sensitivity, uncertainty and discernibility analyses...

  9. Long-term integrity of waste package final closure for HLW geological disposal, (2). Applicability of TIG welding method to overpack final closure

    International Nuclear Information System (INIS)

    Asano, Hidekazu; Sawa, Shuusuke; Aritomi, Masanori

    2005-01-01

    Overpack, a high-level radioactive waste package for geological disposal, seals vitrified waste and in line with Japan's waste management program is required to isolate it from contact with groundwater for 1,000 years. In this study, TIG (Tungsten Inert Gas) welding method, a typical arc welding method and widely used in various industries, was examined for its applicability to seal a carbon steel overpack lid with a thickness of 190 mm. Welding conditions and welding parameters were examined for multi-layer welding in a narrow gap for four different groove depths. Weld joint tests were conducted and weld flaws, macro- and microstructure, and mechanical properties were assessed within tentatively applied criteria for weld joints. Measurement and numerical calculation for residual stress were also conducted and the tendency of residual stress distribution was discussed. These test results were compared with the basic requirements of the welding method for overpack which were pointed out in our first report. It is assessed that the TIG welding method has the potential to provide the necessary requirements to complete the final closure of overpack with a maximum thickness of 190 mm. (author)

  10. Technical solutions for waste treatment in the Belene project

    International Nuclear Information System (INIS)

    Büttner, K.; Eichhorn, H.

    2011-01-01

    Outline: In June 2010 NUKEM Technologies GmbH was awarded a contract from ATOMSTROYEXPORT JSC to perform the complete work package related to designing and completion of the equipment for treatment of radioactive waste on the turn-key basis for Belene NPP. Technical Solutions: Waste Streams and Technologies at UKC and UKS; Concentration Plant; Thermal Treatment of Resins Sorting Facility; Biological Waste Water Treatment; Conditioning – Cementation • Sorting of Radwaste; Plasma Facility; Grouting; Filter Press; Monitoring and Tracking

  11. Electrochemical processing of low-level waste solutions

    International Nuclear Information System (INIS)

    Hobbs, D.T.; Ebra, M.A.

    1987-01-01

    The feasibility of treating low-level Savannah River Plant (SRP) waste solutions by an electrolytic process has been demonstrated. Although the economics of the process are marginal at the current densities investigated at the laboratory scale, there are a number of positive environmental benefits. These benefits include: (1) reduction in the levels of nitrate and nitrite in the waste, (2) further decontamination of 99 Tc and 106 Ru, and (3) reduction in the volume of waste

  12. Industrial Water Waste, Problems and the Solution

    Directory of Open Access Journals (Sweden)

    Alif Noor Anna

    2004-01-01

    Full Text Available Recently, the long term development in Indonesia has changed agricultural sector to the industrial sector. This development can apparently harm our own people. This is due to the waste that is produced from factories. The waste from various factories seems to have different characteristics. This defference encourages us to be able to find out different of methods of managing waste so that cost can be reduced, especially in water treatment. In order that industrial development and environmental preservation can run together in balance, many institutions involved should be consider, especially in the industrial chain, the environment, and human resource, these three elements can be examined in terms of their tolerance to waste.

  13. Environment friendly solutions of plastics waste management

    International Nuclear Information System (INIS)

    Pirzada, F.N.; Riffat, T.; Pirzada, M.D.S.

    1997-01-01

    The use of plastics is growing worldwide. Consequently, the volume of plastic waste is also increasing. Presently, more than 100 million tons per year of plastic is being produced globally. In U.S. alone more than 10 million tons of plastic is being dumped in landfills as waste, where it can persist for decades. This has resulted in exhausting old landfills. Public awareness on environment is also making it difficult to find new sites for landfills. This has led to increased emphasis on treatment and recycling of plastic wastes. Volume reduction of plastic waste has some unique problems. They arise from the intrinsic chemical inertness of polymeric materials and toxic nature of their degradation byproducts. The paper reviews the present state of plastic waste management including land filling, incineration and recycling technologies. The technical problems associated with each of these processes have been discussed. There is also brief description of ongoing R and D for finding improved methods of plastic waste handling with their promises and problems. The role of tougher legislation in developing better recycling methods and degradable plastics has also been evaluated. The claims made by the proponents of degradable polymers have also been critically reviewed. (authors)

  14. 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

  15. 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.

  16. 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

  17. Engineering solutions to the management of solid radioactive waste

    International Nuclear Information System (INIS)

    1991-01-01

    The management of radioactive waste, its safe handling and ultimate disposal, is of vital concern to engineers in the nuclear industry. The international conference 'Engineering Solutions to the Management of Solid Radioactive Waste', organized by the Institution of Mechanical Engineers and held in Manchester in November 1991, provided a forum for the discussion and comparison of the different methods of waste management used in Europe and America. Papers presented and discussed included: the interaction between the design of containers for low level radioactive waste and the design of a deep repository, commercial low level waste disposal sites in the United States, and the development of radioactive waste monitoring systems at the Sellafield reprocessing complex. This volume is a collection of 22 papers presented at the conference. All are indexed separately. (author)

  18. Development of iron phosphate ceramic waste form to immobilize radioactive waste solution

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Jongkwon [Division of Advanced Nuclear Engineering, Pohang University of Science and Technology (POSTECH), San 31, Hyoja-Dong, Pohang (Korea, Republic of); Um, Wooyong, E-mail: wooyong.um@pnnl.gov [Division of Advanced Nuclear Engineering, Pohang University of Science and Technology (POSTECH), San 31, Hyoja-Dong, Pohang (Korea, Republic of); Pacific Northwest National Laboratory, Richland, WA 99354 (United States); Choung, Sungwook [Division of Advanced Nuclear Engineering, Pohang University of Science and Technology (POSTECH), San 31, Hyoja-Dong, Pohang (Korea, Republic of)

    2014-09-15

    The objective of this research was to develop an iron phosphate ceramic (IPC) waste form using converter slag obtained as a by-product of the steel industry as a source of iron instead of conventional iron oxide. Both synthetic off-gas scrubber solution containing technetium-99 (or Re as a surrogate) and LiCl–KCl eutectic salt, a final waste solution from pyrochemical processing of spent nuclear fuel, were used as radioactive waste streams. The IPC waste form was characterized for compressive strength, reduction capacity, chemical durability, and contaminant leachability. Compressive strengths of the IPC waste form prepared with different types of waste solutions were 16 MPa and 19 MPa for LiCl–KCl eutectic salt and the off-gas scrubber simulant, respectively, which meet the minimum compressive strength of 3.45 MPa (500 psi) for waste forms to be accepted into the radioactive waste repository. The reduction capacity of converter slag, a main dry ingredient used to prepare the IPC waste form, was 4136 meq/kg by the Ce(IV) method, which is much higher than those of the conventional Fe oxides used for the IPC waste form and the blast furnace slag materials. Average leachability indexes of Tc, Li, and K for the IPC waste form were higher than 6.0, and the IPC waste form demonstrated stable durability even after 63-day leaching. In addition, the Toxicity Characteristic Leach Procedure measurements of converter slag and the IPC waste form with LiCl–KCl eutectic salt met the universal treatment standard of the leachability limit for metals regulated by the Resource Conservation and Recovery Act. This study confirms the possibility of development of the IPC waste form using converter slag, showing its immobilization capability for radionuclides in both LiCl–KCl eutectic salt and off-gas scrubber solutions with significant cost savings.

  19. 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

  20. 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.

  1. Sustainable solutions for solid waste management in Southeast Asian countries

    International Nuclear Information System (INIS)

    Uyen Nguyen Ngoc; Schnitzer, Hans

    2009-01-01

    Human activities generate waste and the amounts tend to increase as the demand for quality of life increases. Today's rate in the Southeast Asian Nations (ASEANs) is alarming, posing a challenge to governments regarding environmental pollution in the recent years. The expectation is that eventually waste treatment and waste prevention approaches will develop towards sustainable waste management solutions. This expectation is for instance reflected in the term 'zero emission systems'. The concept of zero emissions can be applied successfully with today's technical possibilities in the agro-based processing industry. First, the state-of-the-art of waste management in Southeast Asian countries will be outlined in this paper, followed by waste generation rates, sources, and composition, as well as future trends of waste. Further on, solutions for solid waste management will be reviewed in the discussions of sustainable waste management. The paper emphasizes the concept of waste prevention through utilization of all wastes as process inputs, leading to the possibility of creating an ecosystem in a loop of materials. Also, a case study, focusing on the citrus processing industry, is displayed to illustrate the application of the aggregated material input-output model in a widespread processing industry in ASEAN. The model can be shown as a closed cluster, which permits an identification of opportunities for reducing environmental impacts at the process level in the food processing industry. Throughout the discussion in this paper, the utilization of renewable energy and economic aspects are considered to adapt to environmental and economic issues and the aim of eco-efficiency. Additionally, the opportunities and constraints of waste management will be discussed.

  2. Recovery of fission products from acidic waste solutions thereof

    International Nuclear Information System (INIS)

    Carlin, W.W.; Darlington, W.B.; Dubois, D.W.

    1975-01-01

    Fission products, e.g., palladium, ruthenium and technetium, are removed from aqueous, acidic waste solutions thereof. The acidic waste solution is electrolyzed in an electrolytic cell under controlled cathodic potential conditions and technetium, ruthenium, palladium and rhodium are deposited on the cathode. Metal deposit is removed from the cathode and dissolved in acid. Acid insoluble rhodium metal is recovered, dissolved by alkali metal bisulfate fusion and purified by electrolysis. In one embodiment, the solution formed by acid dissolution of the cathode metal deposit is treated with a strong oxidizing agent and distilled to separate technetium and ruthenium (as a distillate) from palladium. Technetium is separated from ruthenium by organic solvent extraction and then recovered, e.g., as an ammonium salt. Ruthenium is disposed of as waste by-product. Palladium is recovered by electrolysis of an acid solution thereof under controlled cathodic potential conditions. Further embodiments wherein alternate metal recovery sequences are used are described. (U.S.)

  3. 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

  4. 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.

  5. 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

  6. 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

  7. 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

  8. Partitioning of actinides from high active waste solution of Purex origin counter-current extraction studies using TBP and CMPO

    International Nuclear Information System (INIS)

    Chitnis, R.R.; Dhami, P.S.; Gopalkrishnan, V.; Wattal, P.K.; Ramanujam, A.; Murali, M.S.; Mathur, J.N.; Bauri, A.K.; Chattopadhyay, S.

    2000-10-01

    A solvent extraction scheme has been formulated for the partitioning of actinides from Purex high level waste (HLW). The scheme is based on the results of earlier studies carried out with simulated waste solutions. In the present studies, the scheme was tested with high active waste (HAW) solution generated during the reprocessing of spent fuel from research reactors using laboratory scale mixer-settlers. The proposed process involved two-step extraction using tri-n-butyl phosphate (TBP) and octyl (phenyl)-N,N-diisobutylcarbamolylmethylphosphine oxide (CMPO). In the first step, uranium, neptunium and plutonium were removed from the waste using TBP as extractant. The minor actinides left in the raffinate were extracted using a mixture of CMPO and TBP in the second step. The results showed complete extraction of actinides from the waste solution. Plutonium and neptunium extracted in TBP, were stripped together using a mixture of hydrogen peroxide and ascorbic acid in 2 M nitric acid medium, leaving uranium in the organic phase. Uranium can later be stripped using dilute nitric acid. Actinides extracted in CMPO-TBP phase were stripped using a mixture of formic acid, hydrazine, hydrate and citric acid. The stripping was quantitative in both the stripping runs. An additional extraction step for the preferential recovery of uranium, neptunium and plutonium from the waste solution using TBP is a modification over the conventional Truex process. Selective stripping of neptunium and plutonium from large quantities of uranium. The extraction of uranium using TBP eliminates the possibility of third phase and undesired loading of CMPO-TBP in the following step. Use of citrate-containing strippant allows the recovery of actinides from loaded CMPO-TBP mixture without causing any reflux of the actinides during stripping. The process has been developed with due consideration to minimising the generation of secondary wastes. The proposed strippants are effective even in presence of

  9. Overview on the Multinational Collaborative Waste Storage and Disposal Solutions

    International Nuclear Information System (INIS)

    MARGEANU, C.A.

    2013-01-01

    The main drivers for a Safe, Secure and Global Energy future become clear and unequivocal: Security of supply for energy sources, Low-carbon electricity generation and Extended nuclear power assuring economic nuclear energy production, safe nuclear facilities and materials, safe and secure radioactive waste management and public acceptance. Responsible use of nuclear power requires that – in addition to safety, security and environmental protection associated with NPPs operation – credible solutions to be developed for dealing with the radioactive waste produced and especially for a responsible long term radioactive waste management. The paper deals with the existing multinational initiative in nuclear fuel cycle and the technical documents sustaining the multinational/regional disposal approach. Meantime, the paper far-reaching goal is to highlight on: What is offering the multinational waste storage and disposal solutions in terms of improved nuclear security ‽

  10. Photochemical oxidation: A solution for the mixed waste dilemma

    Energy Technology Data Exchange (ETDEWEB)

    Prellberg, J.W.; Thornton, L.M.; Cheuvront, D.A. [Vulcan Peroxidation Systems, Inc., Tucson, AZ (United States)] [and others

    1995-12-31

    Numerous technologies are available to remove organic contamination from water or wastewater. A variety of techniques also exist that are used to neutralize radioactive waste. However, few technologies can satisfactorily address the treatment of mixed organic/radioactive waste without creating unacceptable secondary waste products or resulting in extremely high treatment costs. An innovative solution to the mixed waste problem is on-site photochemical oxidation. Liquid-phase photochemical oxidation has a long- standing history of successful application to the destruction of organic compounds. By using photochemical oxidation, the organic contaminants are destroyed on-site leaving the water, with radionuclides, that can be reused or disposed of as appropriate. This technology offers advantages that include zero air emissions, no solid or liquid waste formation, and relatively low treatment cost. Discussion of the photochemical process will be described, and several case histories from recent design testing, including cost analyses for the resulting full-scale installations, will be presented as examples.

  11. Removal of radioactive materials from waste solutions via magnetic ferrites

    International Nuclear Information System (INIS)

    Boyd, T.E.; Kochen, R.L.; Price, M.Y.

    1982-01-01

    Ferrite waste treatment was found to be effective in removing actinides from simulated Rocky Flats process waste solutions. With a one-stage ferrite treatment, plutonium concentrations were consistently reduced from 10 -4 g/l to less than 10 -8 g/l, and americium concentrations were lowered from 10 -7 g/l to below 10 -10 g/l. In addition, siginficantly less solid was produced as compared with the flocculant precipitation technique now employed at Rocky Flats. Aging of ferrite solids and elevated beryllium and phosphate concentrations were identified as interferences in the ferrite treatment of process waste, but neither appeeared serious enough to prevent implementation in plant operations

  12. Methods for removing transuranic elements from waste solutions

    International Nuclear Information System (INIS)

    Slater, S.A.; Chamberlain, D.B.; Connor, C.; Sedlet, J.; Srinivasan, B.; Vandegrift, G.F.

    1994-11-01

    This report outlines a treatment scheme for separating and concentrating the transuranic (TRU) elements present in aqueous waste solutions stored at Argonne National Laboratory (ANL). The treatment method selected is carrier precipitation. Potential carriers will be evaluated in future laboratory work, beginning with ferric hydroxide and magnetite. The process will result in a supernatant with alpha activity low enough that it can be treated in the existing evaporator/concentrator at ANL. The separated TRU waste will be packaged for shipment to the Waste Isolation Pilot Plant

  13. 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

  14. China's Scientific Investigation for Liquid Waste Treatment Solutions

    International Nuclear Information System (INIS)

    Liangjin, B.; Meiqiong, L.; Kelley, D.

    2006-01-01

    Post World War II created the nuclear age with several countries developing nuclear technology for power, defense, space and medical applications. China began its nuclear research and development programs in 1950 with the establishment of the China Institute of Atomic Energy (CIAE) located near Beijing. CIAE has been China's leader in nuclear science and technical development with its efforts to create advanced reactor technology and upgrade reprocessing technology. In addition, with China's new emphasis on environmental safety, CIAE is focusing on waste treatment options and new technologies that may provide solutions to legacy waste and newly generated waste from the full nuclear cycle. Radioactive liquid waste can pose significant challenges for clean up with various treatment options including encapsulation (cement), vitrification, solidification and incineration. Most, if not all, nuclear nations have found the treatment of liquids to be difficult, due in large part to the high economic costs associated with treatment and disposal and the failure of some methods to safely contain or eliminate the liquid. With new environmental regulations in place, Chinese nuclear institutes and waste generators are beginning to seek new technologies that can be used to treat the more complex liquid waste streams in a form that is safe for transport and for long-term storage or final disposal. [1] In 2004, CIAE and Pacific Nuclear Solutions, a division of Pacific World Trade, USA, began discussions about absorbent technology and applications for its use. Preliminary tests were conducted at CIAE's Department of Radiochemistry using generic solutions, such as lubricating oil, with absorbent polymers for solidification. Based on further discussions between both parties, it was decided to proceed with a more formal test program in April, 2005, and additional tests in October, 2005. The overall objective of the test program was to apply absorbent polymers to various waste streams

  15. 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

  16. Solutions for energy recovery of animal waste from leather industry

    International Nuclear Information System (INIS)

    Lazaroiu, Gheorghe; Pană, Constantin; Mihaescu, Lucian; Cernat, Alexandru; Negurescu, Niculae; Mocanu, Raluca; Negreanu, Gabriel

    2017-01-01

    Highlights: • Animal fats in blend with diesel fuel for energy valorification through combustion. • Animal waste from tanneries as fuel and for biogas production. • Experimental tests using animal fats as fuel for diesel engines. • Experimental tests modifying the characteristic parameters. - Abstract: Secondary products from food and leather industries are regarded as animal wastes. Conversion of these animal wastes into fuels represents an energy recovery solution not only because of their good combustion properties, but also from the viewpoint of supply stability. A tannery factory usually processes 60–70 t/month of crude leathers, resulting in 12–15 t/month of waste. Fats, which can be used as the input fuel for diesel engines (in crude state or as biodiesel), represent 10% of this animal waste, while the rest are proteins that can be used to generate biogas through anaerobic digestion. Herein, we analyse two approaches to the use of animal waste from tanneries: as fuel for diesel engines and for biogas generation for heat production. Diesel fuelling and fuelling by animal wastes are compared in terms of the engine performance and pollutant emissions. The effects of animal waste usage on the pollutant emissions level, exhaust gas temperature, indicated mean effective pressure, maximum pressure, and engine efficiency are analysed. The energy recovery technologies for animal waste, which are analysed in this work, can be easily implemented and can simultaneously solve the problem posed by animal wastes by using them as an alternative to fossil fuels. Animal fats can be considered an excellent alternative fuel for diesel engines without major constructive modifications.

  17. Radioactive Waste...The Problem and Some Possible Solutions

    Science.gov (United States)

    Olivier, Jean-Pierre

    1977-01-01

    Nuclear safety is a highly technical and controversial subject that has caused much heated debate and political concern. This article examines the problems involved in managing radioactive wastes and the techniques now used. Potential solutions are suggested and the need for international cooperation is stressed. (Author/MA)

  18. Food waste in Central Europe – challenges and solutions

    Directory of Open Access Journals (Sweden)

    den Boer Jan

    2017-01-01

    Full Text Available Food waste is an important issue in the global economy. In the EU many activities aimed at this topic are carried out, however in Central Europe is still quite pristine. There is lack of reliable data on food waste quantities in this region, and not many preventive actions are taken. To improve this situation the STREFOWA (Strategies to Reduce and Manage Food Waste in Central Europe was initiated. It is an international project (Austria, Czech Republic, Hungary, Italy, Poland, founded by the Interreg Central Europe programme, running from July 2016 to June 2019. Its main purpose is to provide solutions to prevent and manage food waste throughout the entire food supply chain. The results of STREFOWA will have positive economical, social and environmental impacts.

  19. Food waste in Central Europe - challenges and solutions

    Science.gov (United States)

    den Boer, Jan; Kobel, Przemysław; Dyjakon, Arkadiusz; Urbańska, Klaudia; Obersteiner, Gudrun; Hrad, Marlies; Schmied, Elisabeth; den Boer, Emilia

    2017-11-01

    Food waste is an important issue in the global economy. In the EU many activities aimed at this topic are carried out, however in Central Europe is still quite pristine. There is lack of reliable data on food waste quantities in this region, and not many preventive actions are taken. To improve this situation the STREFOWA (Strategies to Reduce and Manage Food Waste in Central Europe) was initiated. It is an international project (Austria, Czech Republic, Hungary, Italy, Poland), founded by the Interreg Central Europe programme, running from July 2016 to June 2019. Its main purpose is to provide solutions to prevent and manage food waste throughout the entire food supply chain. The results of STREFOWA will have positive economical, social and environmental impacts.

  20. Cement encapsulation of low-level waste liquids. Final report

    International Nuclear Information System (INIS)

    Baker, M.N.; Houston, H.M.

    1999-01-01

    Pretreatment of liquid high-level radioactive waste at the West Valley Demonstration Project (WVDP) was essential to ensuring the success of high-level waste (HLW) vitrification. By chemically separating the HLW from liquid waste, it was possible to achieve a significant reduction in the volume of HLW to be vitrified. In addition, pretreatment made it possible to remove sulfates, which posed several processing problems, from the HLW before vitrification took place

  1. 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

  2. Necessary contents of public outreach for high level radioactive waste disposal

    International Nuclear Information System (INIS)

    Kanzaki, Noriko; Okamoto, Koji

    2011-01-01

    Nuclear power generation is one of the solutions for global warming. However, the nuclear power generation technology can not be completed unless the disposal method of the radioactive waste is decided. Various actions are performed about the High Level Radioactive Waste (HLW) disposal in particular in each country. However, planning of HLW disposal site was not successful, except Finland and Sweden. In Japan, geological disposal of HLW was selected. The operating body and the capital management body are also decided. Up to the present, no municipality apply the disposal site candidate. An important social element for HLW disposal is careful explanation and communication for municipality. For this purpose, a symposium to explain necessity of HLW is held in each district in Japan. The symposium is not successful, because of lack of carefulness to local situation considered. In this study, we evaluates the questionnaire by the symposium attendee to extract the idea and requests by the local people. With these questionnaire, the responsibility of the government should be more enhanced. Also, the detail answer to the people's questions are needed. Using these knowledge, the HLW disposal social acceptance has been discussed. (author)

  3. 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

  4. 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.

  5. Removal of fluoride ions from aqueous solution by waste mud

    International Nuclear Information System (INIS)

    Kemer, Baris; Ozdes, Duygu; Gundogdu, Ali; Bulut, Volkan N.; Duran, Celal; Soylak, Mustafa

    2009-01-01

    The present study was carried out to assess the ability of original waste mud (o-WM) and different types of activated waste mud which are acid-activated (a-WM) and precipitated waste mud (p-WM), in order to remove excess of fluoride from aqueous solution by using batch technique. The p-WM exhibited greater performance than the others. Adsorption studies were conducted as a function of pH, contact time, initial fluoride concentration, adsorbent concentration, temperature, etc. Studies were also performed to understand the effect of some co-existing ions present in aqueous solutions. Adsorption process was found to be almost independent of pH for all types of waste mud. Among the kinetic models tested for p-WM, pseudo-second-order model fitted the kinetic data well with a perfect correlation coefficient value of 1.00. It was found that the adequate time for the adsorption equilibrium of fluoride was only 1 h. Thermodynamic parameters including the Gibbs free energy (ΔG o ), enthalpy (ΔH o ), and entropy (ΔS o ) revealed that adsorption of fluoride ions on the p-WM was feasible, spontaneous and endothermic in the temperature range of 0-40 deg. C. Experimental data showed a good fit with the Langmuir and Freundlich adsorption isotherm models. Results of this study demonstrated the effectiveness and feasibility of WM for removal of fluoride ions from aqueous solution.

  6. Removal of fluoride ions from aqueous solution by waste mud

    Energy Technology Data Exchange (ETDEWEB)

    Kemer, Baris; Ozdes, Duygu; Gundogdu, Ali; Bulut, Volkan N.; Duran, Celal [Karadeniz Technical University, Faculty of Arts and Sciences, Department of Chemistry, 61080 Trabzon (Turkey); Soylak, Mustafa, E-mail: soylak@erciyes.edu.tr [Erciyes University, Faculty of Arts and Sciences, Department of Chemistry, 38039 Kayseri (Turkey)

    2009-09-15

    The present study was carried out to assess the ability of original waste mud (o-WM) and different types of activated waste mud which are acid-activated (a-WM) and precipitated waste mud (p-WM), in order to remove excess of fluoride from aqueous solution by using batch technique. The p-WM exhibited greater performance than the others. Adsorption studies were conducted as a function of pH, contact time, initial fluoride concentration, adsorbent concentration, temperature, etc. Studies were also performed to understand the effect of some co-existing ions present in aqueous solutions. Adsorption process was found to be almost independent of pH for all types of waste mud. Among the kinetic models tested for p-WM, pseudo-second-order model fitted the kinetic data well with a perfect correlation coefficient value of 1.00. It was found that the adequate time for the adsorption equilibrium of fluoride was only 1 h. Thermodynamic parameters including the Gibbs free energy ({Delta}G{sup o}), enthalpy ({Delta}H{sup o}), and entropy ({Delta}S{sup o}) revealed that adsorption of fluoride ions on the p-WM was feasible, spontaneous and endothermic in the temperature range of 0-40 deg. C. Experimental data showed a good fit with the Langmuir and Freundlich adsorption isotherm models. Results of this study demonstrated the effectiveness and feasibility of WM for removal of fluoride ions from aqueous solution.

  7. The waste management at research laboratories - problems and solutions

    International Nuclear Information System (INIS)

    Dellamano, Jose Claudio; Vicente, Roberto

    2011-01-01

    The radioactive management in radioactive installations must be planned and controlled. However, in the case of research laboratories, that management is compromised due to the common use of materials and installations, the lack of trained personnel and the nonexistence of clear and objective orientations by the regulator organism. Such failures cause an increasing of generated radioactive wastes and the imprecision or nonexistence of record of radioactive substances, occasioning a financial wastage, and the cancelling of licences for use of radioactive substances. This paper discusses and proposes solutions for the problems found at radioactive waste management in research laboratories

  8. Mixed waste: An alternative solution. The utility perspective

    International Nuclear Information System (INIS)

    Seizert, R.D.

    1988-01-01

    The issue of mixed waste is one of significant interest to the utility industry. The interest is focused on the current regulatory scheme of dual regulation. A fundamental concern of the commercial nuclear utilities resulting from dual regulation is that there are currently no facilities in the US to dispose of mixed low-level radioactive and hazardous waste. The lack of available sites renders mixed waste an orphan, requiring generators of such material to store the waste on-site. This in turn causes commercial nuclear power plants to be subjected to the full gamut of Environmental Protection Agency (EPA) Resource Conservation and Recovery Act (RCRA) regulation in addition to the existing Nuclear Regulatory Commission (NRC) regulations. Superimposing dual regulatory schemes will have impacts which extend far beyond the mere management of mixed waste. Certainly the burdens, complexities and costs of complying with the overlapping regulatory schemes will not have a commensurate increase in protection from the real risks being addressed. For these reasons, the commercial nuclear utility industry is working toward an alternative solution which will protect the public health and the environment from all hazards of mixed waste and will minimize the impacts on both the regulators and the regulated community

  9. Encapsulating of high-level radioactive waste with use of pyrocarbon and silicon carbide coatings

    International Nuclear Information System (INIS)

    Chernikov, A.

    2007-01-01

    It is known that high-level radioactive waste (HLW) constitute a real danger to biosphere, especially that their part, which contains transuranium and long-lived radionuclides resulting during reprocessing of nuclear fuel industrial and power reactors. Such waste contains approximately 99 % of long-lived fission products and transplutonium elements. At present, the concept of multi barrier protection of biosphere from radioactive waste is generally acknowledged. The main barriers are the physicochemical form of waste and enclosing strata of geological formation at places of waste-disposal. Applied methods of solidification of HLW with preparation of phosphatic and borosilicate glasses do not guarantee in full measure safety of places of waste-disposal of solidified waste in geological formations during thousand years. One promising way to improve HLW handling safety is placing of radionuclides in mineral-like matrixes similar to natural materials. The other possible way to increase safety of HLW disposal places is suggested for research by experts of Russian research institutes, for example, in the proposal for the Project of ISTC and considered in the present report, is to introduce an additional barrier on a radionuclides migration path by coating of HLW particles. Unique protective properties of pyrocarbon and silicon carbide such as low coefficients of diffusion of gaseous and solid fission products and high chemical and radiation stability [1] attract attention to these materials for coating of solidified HLW. The objective of the Project is the development of method of HLW encapsulating with use of pyrocarbon and silicon carbide coatings. To gain this end main direction of researches, including analysis of various encapsulation processes of fractionated HLW, and expected results are presented. Realization of the Project will allow to prove experimentally the efficiency of the proposed approach in the solution of the problem of HLW conditioning and ecological

  10. Intermediate Level Waste Research Programme: Progress report for 1986/87 from the Waste Treatment and Disposal Working Party covering Joint Funded Work

    International Nuclear Information System (INIS)

    Claxton, D.G.S.A.

    1988-06-01

    The Waste Treatment and Disposal Working Party (WTDWP) covered the areas of: ILW Product Evaluation; ILW and HLW Disposal Studies, and ILW and HLW Quality Checking. The objectives of the programme were to evaluate potential waste products arising from the treatment of ILW/HLW, and to develop appropriate techniques which could be used to check the quality of the finished waste product. (author)

  11. Dissolution of Simulated and Radioactive Savannah River Site High-Level Waste Sludges with Oxalic Acid & Citric Acid Solutions

    International Nuclear Information System (INIS)

    STALLINGS, MARY

    2004-01-01

    This report presents findings from tests investigating the dissolution of simulated and radioactive Savannah River Site sludges with 4 per cent oxalic acid and mixtures of oxalic and citric acid previously recommended by a Russian team from the Khlopin Radium Institute and the Mining and Chemical Combine (MCC). Testing also included characterization of the simulated and radioactive waste sludges. Testing results showed the following: Dissolution of simulated HM and PUREX sludges with oxalic and citric acid mixtures at SRTC confirmed general trends reported previously by Russian testing. Unlike the previous Russian testing six sequential contacts of a mixture of oxalic acid citric acids at a 2:1 ratio (v/w) of acid to sludge did not produce complete dissolution of simulated HM and PUREX sludges. We observed that increased sludge dissolution occurred at a higher acid to sludge ratio, 50:1 (v/w), compared to the recommended ratio of 2:1 (v/w). We observed much lower dissolution of aluminum in a simulated HM sludge by sodium hydroxide leaching. We attribute the low aluminum dissolution in caustic to the high fraction of boehmite present in the simulated sludge. Dissolution of HLW sludges with 4 per cent oxalic acid and oxalic/citric acid followed general trends observed with simulated sludges. The limited testing suggests that a mixture of oxalic and citric acids is more efficient for dissolving HM and PUREX sludges and provides a more homogeneous dissolution of HM sludge than oxalic acid alone. Dissolution of HLW sludges in oxalic and oxalic/citric acid mixtures produced residual sludge solids that measured at higher neutron poison to equivalent 235U weight ratios than that in the untreated sludge solids. This finding suggests that residual solids do not present an increased nuclear criticality safety risk. Generally the neutron poison to equivalent 235U weight ratios of the acid solutions containing dissolved sludge components are lower than those in the untreated

  12. 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

  13. Processing of nuclear waste

    International Nuclear Information System (INIS)

    Hennelly, E.J.

    1981-01-01

    The processing of nuclear waste to transform the liquid waste from fuel reprocessing activities is well defined. Most solid waste forms, if they are cooled and contain diluted waste, are compatible with many permanent storage environments. The public acceptance of methods for disposal is being delayed in the US because of the alternatives studies of waste forms and repositories now under way that give the impression of indecision and difficulty for the disposal of HLW. Conservative programs that dilute and cool solid waste are under way in France and Sweden and demonstrate that a solution to the problem is available now. Research and development should be directed toward improving selected methods rather than seeking a best method, which at best, may always be illusory

  14. DTDGA impregnated XAD-16 beads for separation and recovery of palladium from acidic wastes

    International Nuclear Information System (INIS)

    Singh, Krishan Kant; Kanagare, Anant B.; Bairwa, K.K.; Manmohan Kumar; Bajaj, Parma N.; Ruhela, Ritesh; Singh, Ajoy K.; Hubli, Rajendra C.

    2014-01-01

    In the present study DTDGA extractant impregnated XAD16 polymeric beads (DTDGA- XAD16) were synthesized by wet chemical method and the beads were characterized by various techniques viz. FT-IR, Optical microscopy, SEM and TGA. The DTDGA-XAD 16 beads were evaluated for separation and recovery of palladium from high level waste solutions (HLW)

  15. Women, e-waste, and technological solutions to climate change.

    Science.gov (United States)

    McAllister, Lucy; Magee, Amanda; Hale, Benjamin

    2014-06-14

    In this paper, we argue that a crossover class of climate change solutions (which we term "technological solutions") may disproportionately and adversely impact some populations over others. We begin by situating our discussion in the wider climate discourse, particularly with regard to the Millennium Development Goals (MDGs) and the Basel Convention. We then suggest that many of the most attractive technological solutions to climate change, such as solar energy and electric car batteries, will likely add to the rapidly growing stream of electronic waste ("e-waste"). This e-waste may have negative downstream effects on otherwise disenfranchised populations. We argue that e-waste burdens women unfairly and disproportionately, affecting their mortality/morbidity and fertility, as well as the development of their children. Building on this, we claim that these injustices are more accurately captured as problems of recognition rather than distribution, since women are often institutionally under-acknowledged both in the workplace and in the home. Without institutional support and representation, women and children are deprived of adequate safety equipment, health precautions, and health insurance. Finally, we return to the question of climate justice in the context of the human right to health and argue for greater inclusion and recognition of women waste workers and other disenfranchised groups in forging future climate agreements. Copyright © 2014 McAllister, Magee. This is an open access article distributed under the terms of the Creative Commons Attribution Non-Commercial License (http://creativecommons.org/licenses/by-nc/3.0/), which permits unrestricted non-commercial use, distribution, and reproduction in any medium, provided the original author and source are credited.

  16. An eco friendly solution to the food waste disposal

    Science.gov (United States)

    Babu, G. Reddy; Kumar, G. Madhav

    2017-07-01

    In recent years, waste disposal at workmen camp is one of the major problems being faced by many nations across the world. In the workmen colony at Chittapur, a series of kitchens were built for cooking purpose and a number of small canteens are also functioning. Considerable quantity of food waste is collected daily from these eateries and disposed at a faraway place. Food waste is highly degradable in nature, if not disposed properly it causes problems related to environmental pollution. Hence, it is very important to identify an environment friendly process rather than opt for land filling or any disposal method. We worked together to find a suitable eco-friendly solution for the food waste disposal at Chittapur site and suggested that biogas production through anaerobic digestion is a solution for the disposal and utilization of food waste for better purpose. This resulted in setting up a 500 kg per day food waste treatment biogas plant at Chittapur. This establishment is the first time in the construction industry at workmen camp in India. Anaerobic Digestion has been recognized as one of the best options that is available for treating food waste, as it generates two valuable end products, biogas and compost. Biogas is a mixture of CH4 and CO2 about (55:45). Biogas generated can be used for thermal applications such as cooking or for generating electricity. The digested slurry is a well stabilized organic manure and can be used as soil fertilizer. Plant design is to handle 500 kg of food waste /day. 27 kg LPG is obtained from 500kg of kitchen waste. The Value of 27 kg of LPG is Rs.2700/day. Daily 1000 litres of digested effluent was obtained. It is good organic manure with plant micro nutrients and macro nutrients. This can be used for growing plants and in agriculture. The value of manure per day is Rs.250/-. The annual revenue is Rs.10.62 lakhs and the annual expenditure is 1.8 lakhs. The net benefit is 8.82 lakhs. Payback period is 2.1 years. This process

  17. Distillation as a pretreatment process of waste scintillation solutions

    International Nuclear Information System (INIS)

    Dellamano, J.C.

    1988-05-01

    A process to pretreat scintillation solutions composed basically of PPO, POPOP, TOLUENE and ANTAROX, utilized by radioimmunoassay laboratories, is described. The technique employed is distillation which permits a waste reduction to about 40% of the initial volume with the recovery of the solvent (toluene). The recovered toluene can be resued for the same purpose, since it is free of radioactive material as assured by quality control procedures. (author) [pt

  18. Process for denitrating waste solutions containing nitrates and actinides with simultaneous separation of the actinides

    International Nuclear Information System (INIS)

    Gompper, K.

    1986-01-01

    The invention is intended to reduce the acid and nitrate content of nitrate waste solutions, to reduce the total salt content of the waste solution, to remove the actinides contained in it by precipitation, without any danger of violent reactions or an increase in the volume of the waste solution. The invention achieves this by mixing the waste solution with diethyl oxalate at room temperature and heating the mixture to at least 80 0 C. (orig./PW) [de

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  20. 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)

  1. 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.

  2. 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)

  3. 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)

  4. Decontamination of waste radioactive polluted solutions in radiation treatment

    International Nuclear Information System (INIS)

    Simova, G.; Boyadzhiev, A.; Mikhajlov, M.G.; Shopov, N.

    1979-01-01

    The decontamination capacity of solutions of the trivial cleaning Bulgarian preparations ''Mipro'', ''Sana'', ''Synthek'' and ''Univer'' for different surfaces (steel, glass, PVC and linoleum) contaminated with cesium-134, strontium-85 or cerium-144 chlorides, was studied. Concentrations from 5 to 15 g/l of the solutions used in this study displayed a degree of cleaning over 90%. Higher concentration of the solution does not improve its cleaning capacity. For evaluation of foam formation by the solutions, the so called ''foam column stability coefficient'' has been adopted. This coefficient represents the ratio between the height of the foam column and the time of its half life, referred to the time for the foam column formation when blown through with a constant air current. On the basis of this index, solutions of the preparation ''Mipro'' proved to be the best ones for decontamination - in the whole investigated concentration span, the foam column stability coefficient for the solutions of this preparation is with two orders lower than the respective coefficient of the other preparations. It was experimentally established that radiation treatment of radio-contaminated solutions reduces the foam column stability coefficient. Radiation treatment should be carried out in a gamma field, realizing at least one megarad within an acceptable for the liquid wastes time period. (A.B.)

  5. 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.

  6. 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

  7. Geopolymerisation of fly ashes with waste aluminium anodising etching solutions.

    Science.gov (United States)

    Ogundiran, M B; Nugteren, H W; Witkamp, G J

    2016-10-01

    Combined management of coal combustion fly ash and waste aluminium anodising etching solutions using geopolymerisation presents economic and environmental benefits. The possibility of using waste aluminium anodising etching solution (AES) as activator to produce fly ash geopolymers in place of the commonly used silicate solutions was explored in this study. Geopolymerisation capacities of five European fly ashes with AES and the leaching of elements from their corresponding geopolymers were studied. Conventional commercial potassium silicate activator-based geopolymers were used as a reference. The geopolymers produced were subjected to physical, mechanical and leaching tests. The leaching of elements was tested on 28 days cured and crushed geopolymers using NEN 12457-4, NEN 7375, SPLP and TCLP leaching tests. After 28 days ambient curing, the geopolymers based on the etching solution activator showed compressive strength values between 51 and 84 MPa, whereas the commercial potassium silicate based geopolymers gave compressive strength values between 89 and 115 MPa. Based on the regulatory limits currently associated with the used leaching tests, all except one of the produced geopolymers (with above threshold leaching of As and Se) passed the recommended limits. The AES-geopolymer geopolymers demonstrated excellent compressive strength, although less than geopolymers made from commercial activator. Additionally, they demonstrated low element leaching potentials and therefore can be suitable for use in construction works. Copyright © 2016. Published by Elsevier Ltd.

  8. Solidification of radioactive waste solutions by pelletization technique

    International Nuclear Information System (INIS)

    Akbar, A.H.; Koester, R.; Rudolph, G.

    1980-04-01

    A possible way of performing the cement fixation of radioactive wastes is the incorporation into cement pellets on a pan pelletizer, followed by embedding the pellets into an inactive cement matrix. This procedure is suitable for various types of waste, particularly for medium level liquid wastes, and can be used both at drum disposal and at in-situ solidification. This report describes some initial studies on the pelletization technique using a laboratory pelletizer. Formation and size of the pellets have been found to be determined by speed, angle, and load of the pan, ratio and mode of addition of the liquid and solid components, ect. Pellets in various compositions have been produced from cement and water or simulated waste solution, in some cases with the addition of bentonite for improving cesium retention. Some mechanical properties of the pellets such as fall height of fresh pellets, development of hardness (crush test), impact and abrasion resistance, have been determined. Some preliminary experiments were done on backfilling the void space between the pellets - about 40 per cent of the bulk volume - with cement grouts of appropriate compositions. (orig.) [de

  9. The waste management program VUB-AZ: An integrated solution for nuclear biomedical waste management

    International Nuclear Information System (INIS)

    Covens, P.; Sonck, M.; Eggermont, G.; Meert, D.

    2001-01-01

    unit will be compared with the MDA obtained by different handheld monitors. All results will be finally correlated to the different proposed clearance levels. These clearance levels can easily be met through on-site storage for radionuclides with half-life less than 1 year. For a waste stream of 1000 packages or more a year, a management software is indispensable. The software 'WasteMan' was developed on-site. This user-friendly software takes care of the entire storage procedure and allows a complete bookkeeping of the daily nuclear waste streams. Based on the sophisticated waste collection procedure, the WasteMan software allows both a complete inventory of the storage facility and a full traceability of all waste packages from production to either clearance or disposal. At the same time all necessary documents for either clearance or disposal are generated automatically. The data-exchange between several interfaces enables timesaving administration. In addition to these technical aspects a general analysis of the economic impact of such an on- site decay program will be made for a medium sized university with hospital, yielding a serious reduction of waste handling costs. This waste storage program, including the complete measurement set-up and the necessary management software, was recently installed in a second university, proving the general applicability of the whole concept for biomedical nuclear waste. Many hospitals and other biomedical centres however produce small quantities of nuclear waste for which investments, like measurement equipment and decay rooms, are not cost-effective. The installation of a regional centre for nuclear biomedical waste will be presented here as an alternative solution for this problem

  10. 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

  11. 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

  12. Measurement of Solute Diffusion Behavior in Fractured Waste Glass Media

    International Nuclear Information System (INIS)

    Saripalli, Kanaka P.; Lindberg, Michael J.; Meyer, Philip D.

    2008-01-01

    Determination of aqueous phase diffusion coefficients of solutes through fractured media is essential for understanding and modeling contaminants transport at many hazardous waste disposal sites. No methods for earlier measurements are available for the characterization of diffusion in fractured glass blocks. We report here the use of time-lag diffusion experimental method to assess the diffusion behavior of three different solutes (Cs, Sr and Pentafluoro Benzoic Acid or PFBA) in fractured, immobilized low activity waste (ILAW) glass forms. A fractured media time-lag diffusion experimental apparatus that allows the measurement of diffusion coefficients has been designed and built for this purpose. Use of time-lag diffusion method, a considerably easier experimental method than the other available methods, was not previously demonstrated for measuring diffusion in any fractured media. Hydraulic conductivity, porosity and diffusion coefficients of a solute were experimentally measured in fractured glass blocks using this method for the first time. Results agree with the range of properties reported for similar rock media earlier, indicating that the time-lag experimental method can effectively characterize the diffusion coefficients of fractured ILAW glass media

  13. 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

  14. Inhibition of nuclear waste solutions containing multiple aggressive anions

    International Nuclear Information System (INIS)

    Congdon, J.W.

    1987-01-01

    The inhibition of localized corrosion of carbon steel in caustic, high-level radioactive waste solutions was studied using cyclic potentiodynamic polarization scans, supplemented by partially immersed coupon tests. The electrochemical tests provided a rapid and accurate means of determining the relationship between the minimum inhibitor requirements and the concentration of the aggressive anions in this system. Nitrate, sulfate, chloride, and fluoride were identified as aggressive anions, however, no synergistic effects were observed between these anions. This observation may have important theoretical implications because it tends to contradict the behavior of aggressive anions as predicted by existing theories for localized corrosion. 10 refs., 5 figs., 2 tabs

  15. Cs separation from nitric acid solutions of radioactive waste

    International Nuclear Information System (INIS)

    Heckmann, K.; Pieronczyk, W.; Strnad, J.; Feldmaier, F.

    1989-01-01

    It was the objective of this study to selectively separate active caesium (Cs-134 and Cs-137) from acid radioactive waste solutions (especially MAW and HAWC). The following 'strategy' was designed for a separation process: synthesis of reagents which are acid-resistant and selective for caesium; precipitation of Cs + and separation of the precipitates by filtration or centrifugation or precipitation of Cs + and separation of the precipitates by flotation; caesium separation by liquid-liquid extraction. As precipitating agents, sodium tetraphenylborate (kalignost) and several of its fluorine derivatives were examined. (orig./RB) [de

  16. 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.

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

    International Nuclear Information System (INIS)

    Grimmel, E.

    1990-01-01

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

  18. Photometric estimation of plutonium in product solutions and acid waste solutions using flow injection analysis technique

    International Nuclear Information System (INIS)

    Dhas, A.J.A.; Dharmapurikar, G.R.; Kumaraguru, K.; Vijayan, K.; Kapoor, S.C.; Ramanujam, A.

    1995-01-01

    Flow injection analysis technique is employed for the measurement of plutonium concentrations in product nitrate solutions by measuring the absorbance of Pu(III) at 565 nm and of Pu(IV) at 470 nm, using a Metrohm 662 photometer, with a pyrex glass tube of 2 nm (ID) inserted in the light path of the detector serving as a flow cell. The photometer detector never comes in contact with radioactive solution. In the case of acid waste solutions Pu is first purified by extraction chromatography with 2-ethyl hexyl hydrogen 2 ethyl hexyl phosphonate (KSM 17)- chromosorb and the Pu in the eluate in complexed with Arsenazo III followed by the measured of absorbance at 665 nm. Absorbance of reference solutions in the desired concentration ranges are measured to calibrate the system. The results obtained agree with the reference values within ±2.0%. (author). 3 refs., 1 tab

  19. Community Solutions to Solid Waste Pollution. Operation Waste Watch: The New Three Rs for Elementary School. Grade 6. [Second Edition.

    Science.gov (United States)

    Virginia State Dept. of Waste Management, Richmond. Div. of Litter & Recycling.

    This publication, the last in a series of seven for elementary schools, is an environmental education curriculum guide with a focus on waste management issues. It contains a unit of exercises selected for sixth grade students focusing on community solutions to solid waste pollution. Waste management activities included in this unit seek to…

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

    International Nuclear Information System (INIS)

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

    1997-07-01

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

  1. Waste treatment

    International Nuclear Information System (INIS)

    Hutson, G.V.

    1996-01-01

    Numerous types of waste are produced by the nuclear industry ranging from high-level radioactive and heat-generating, HLW, to very low-level, LLW and usually very bulky wastes. These may be in solid, liquid or gaseous phases and require different treatments. Waste management practices have evolved within commercial and environmental constraints resulting in considerable reduction in discharges. (UK)

  2. Method for recovering palladium and technetium values from nuclear fuel reprocessing waste solutions

    Science.gov (United States)

    Horwitz, E. Philip; Delphin, Walter H.

    1979-07-24

    A method for recovering palladium and technetium values from nuclear fuel reprocessing waste solutions containing these and other values by contacting the waste solution with an extractant of tricaprylmethylammonium nitrate in an inert hydrocarbon diluent which extracts the palladium and technetium values from the waste solution. The palladium and technetium values are recovered from the extractant and from any other coextracted values with a strong nitric acid strip solution.

  3. 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

  4. 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.

  5. Underground radioactive waste disposal concept

    International Nuclear Information System (INIS)

    Frgic, L.; Tor, K.; Hudec, M.

    2002-01-01

    The paper presents some solutions for radioactive waste disposal. An underground disposal of radioactive waste is proposed in deep boreholes of greater diameter, fitted with containers. In northern part of Croatia, the geological data are available on numerous boreholes. The boreholes were drilled during investigations and prospecting of petroleum and gas fields. The available data may prove useful in defining safe deep layers suitable for waste repositories. The paper describes a Russian disposal design, execution and verification procedure. The aim of the paper is to discuss some earlier proposed solutions, and present a solution that has not yet been considered - lowering of containers with high level radioactive waste (HLW) to at least 500 m under the ground surface.(author)

  6. Radiolytic gas formation in high-level liquid waste solutions

    International Nuclear Information System (INIS)

    Brodda, B.-G.; Dix, Siegfried; Merz, E.R.

    1989-01-01

    High-level fission product waste solutions originating from the first-cycle raffinate stream of spent fast breeder reactor fuel reprocessing have been investigated gas chromatographically for their radiolytic and chemical gas production. The solutions showed considerable formation of hydrogen, carbon dioxide and dinitrogen oxide, whereas atmospheric oxygen was consumed completely within a short time. In particular, carbon dioxide resulted from the radiolytic degradation of entrained organic solvent. After nearly complete degradation of the organic solvent, the influence of hydrazine and nitrogen dioxide on hydrogen formation was investigated. Hydrazinium hydroxide led to the formation of dinitrogen oxide and nitrogen. After 60 d, the concentration of dinitrogen oxide had reduced to zero, whereas the amount of nitrogen formed had reached a maximum. This may be explained by simultaneous chemical and radiolytic reactions leading to the formation of dinitrogen oxide and nitrogen and photolytic fission of dinitrogen oxide. Addition of sodium nitrite resulted in the rapid formation of dinitrogen oxide. The rate of hydrogen production was not changed significantly after the addition of hydrazine or nitrite. The results indicate that under normal operating conditions no dangerous hydrogen radiolysis yields should develop in the course of reprocessing and high-level liquid waste tank storage. Organic entrainment may lead to enhanced radiolytic decomposition and thus to considerable hydrogen production rates and pressure build-up in closed systems. (author)

  7. Partitioning of actinide from simulated high level wastes arising from reprocessing of PHWR fuels: counter current extraction studies using CMPO

    International Nuclear Information System (INIS)

    Deshingkar, D.S.; Chitnis, R.R.; Wattal, P.K.; Theyyunni, T.K.; Nair, M.K.T.; Ramanujam, A.; Dhami, P.S.; Gopalakrishnan, V.; Rao, M.K.; Mathur, J.N.; Murali, M.S.; Iyer, R.H.; Badheka, L.P.; Banerji, A.

    1994-01-01

    High level wastes (HLW) arising from reprocessing of pressurised heavy water reactor (PHWR) fuels contain actinides like neptunium, americium and cerium which are not extracted in the Purex process. They also contain small quantities of uranium and plutonium in addition to fission products. Removal of these actinides prior to vitrification of HLW can effectively reduce the active surveillance period of final waste form. Counter current studies using indigenously synthesised octyl (phenyl)-N, N-diisobutylcarbamoylmethylphosphine oxide (CMPO) were taken up as a follow-up of successful runs with simulated sulphate bearing low acid HLW solutions. The simulated HLW arising from reprocessing of PHWR fuel was prepared based on presumed burnup of 6500 MWd/Te of uranium, 3 years cooling period and 800 litres of waste generation per tonne of fuel reprocessed. The alpha activity of the HLW raffinate after extraction with the CMPO-TBP mixture could be brought down to near background level. (author). 13 refs., 2 tabs., 12 figs

  8. Partitioning of actinide from simulated high level wastes arising from reprocessing of PHWR fuels: counter current extraction studies using CMPO

    Energy Technology Data Exchange (ETDEWEB)

    Deshingkar, D S; Chitnis, R R; Wattal, P K; Theyyunni, T K; Nair, M K.T. [Bhabha Atomic Research Centre, Bombay (India). Process Engineering and Systems Development Div.; Ramanujam, A; Dhami, P S; Gopalakrishnan, V; Rao, M K [Bhabha Atomic Research Centre, Bombay (India). Fuel Reprocessing Group; Mathur, J N; Murali, M S; Iyer, R H [Bhabha Atomic Research Centre, Bombay (India). Radiochemistry Div.; Badheka, L P; Banerji, A [Bhabha Atomic Research Centre, Bombay (India). Bio-organic Div.

    1994-12-31

    High level wastes (HLW) arising from reprocessing of pressurised heavy water reactor (PHWR) fuels contain actinides like neptunium, americium and cerium which are not extracted in the Purex process. They also contain small quantities of uranium and plutonium in addition to fission products. Removal of these actinides prior to vitrification of HLW can effectively reduce the active surveillance period of final waste form. Counter current studies using indigenously synthesised octyl (phenyl)-N, N-diisobutylcarbamoylmethylphosphine oxide (CMPO) were taken up as a follow-up of successful runs with simulated sulphate bearing low acid HLW solutions. The simulated HLW arising from reprocessing of PHWR fuel was prepared based on presumed burnup of 6500 MWd/Te of uranium, 3 years cooling period and 800 litres of waste generation per tonne of fuel reprocessed. The alpha activity of the HLW raffinate after extraction with the CMPO-TBP mixture could be brought down to near background level. (author). 13 refs., 2 tabs., 12 figs.

  9. Process for denitrating waste solutions containing nitric acid actinides simultaneously separating the actinides

    International Nuclear Information System (INIS)

    Gompper, K.

    1984-01-01

    The invention should reduce the acid and nitrate content of waste solutions containing nitric acid as much as possible, should reduce the total salt content of the waste solution, remove the actinides contained in it by precipitation and reduce the α radio-activity in the remaining solution, without having to worry about strong reactions or an increase in the volume of the waste solution. The invention achieves this by mixing the waste solution with diethyl oxalate at room temperature and heating the mixture to at least 80 0 C. (orig.) [de

  10. 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

  11. Effects of Fuel to Synthesis of CaTiO3 by Solution Combustion Synthesis for High-Level Nuclear Waste Ceramics.

    Science.gov (United States)

    Jung, Choong-Hwan; Kim, Yeon-Ku; Han, Young-Min; Lee, Sang-Jin

    2016-02-01

    A solution combustion process for the synthesis of perovskite (CaTiO3) powders is described. Perovskite is one of the crystalline host matrics for the disposal of high-level radioactive wastes (HLW) because it immobilizes Sr and Lns elements by forming solid solutions. Solution combustion synthesis, which is a self-sustaining oxi-reduction reaction between nitrate and organic fuel, the exothermic reaction, and the heat evolved convert the precursors into their corresponding oxide products above 1100 degrees C in air. To investigate the effects of amino acid on the combustion reaction, various types of fuels were used; a glycine, amine and carboxylic ligand mixture. Sr, La and Gd-nitrate with equivalent amounts of up to 20% of CaTiO3 were mixed with Ca and Ti nitrate and amino acid. X-ray diffraction analysis, SEM and TEM were conducted to confirm the formed phases and morphologies. While powders with an uncontrolled shape are obtained through a general oxide-route process, Ca(Sr, Lns)TiO3 powders with micro-sized soft agglomerates consisting of nano-sized primary particles can be prepared using this method.

  12. Innovative waste management solutions: An outlook for the future

    Energy Technology Data Exchange (ETDEWEB)

    1993-01-01

    At a conference on various aspects of waste management, papers were presented on the effects of landfills, plastic debris in the marine environment, skills development and training in the waste industry, composting, remediation of contaminated soils, disposal methods, sludge derived byproducts, recycling, waste management economics, municipal solid waste management, ship-generated waste, managing waste in national parks, and septic tank sludge treatment. Separate abstracts have been prepared for five papers from this conference.

  13. Recovery of fission products from waste solutions utilizing controlled cathodic potential electrolysis

    International Nuclear Information System (INIS)

    Carlin, W.W.; Darlington, W.B.

    1975-01-01

    Fission products, e.g., palladium, rhodium and technetium, are recovered from aqueous waste solutions thereof, e.g., aged Purex alkaline waste solutions. The metal values from the waste solutions are extracted by ion exchange techniques. The metals adsorbed by the ion exchange resin are eluted and selectively recovered by controlled cathodic potential electrolysis. The metal values deposited on the cathode are recovered and, if desired, further purified

  14. Synergistic extraction behaviour of americium from simulated acidic waste solutions

    International Nuclear Information System (INIS)

    Pathak, P.N.; Veeraraghavan, R.; Mohapatra, P.K.; Manchanda, V.K.

    1998-01-01

    The extraction behaviour of americium has been investigated with mixtures of 3-phenyl-4-benzoyl-5-isoxazolone (PBI) and oxodonors viz. tri-n-butyl phosphate (TBP), tri-n-octyl phosphine oxide (TOPO) and di-n-butyl octanamide (DBOA) using dodecane as the diluent from 1-6 M HNO 3 media. It is observed that D Am remains unaltered with PBI concentration (in the range 0.06-0.1 M) at 1.47 M TBP in the entire range of HNO 3 concentration. PBI and TBP in combination appears more promising compared to other synergistic systems. The possibility of using this mixture for americium removal from high level liquid waste solution has been explored. Extraction studies indicated that prior removal of uranium by 20% TBP in dodecane is helpful in the quantitative recovery of americium in three contacts. Effect of lanthanides on D Am is found to be marginal. (orig.)

  15. New sorbents and ion exchangers for nuclear waste solution remediation

    International Nuclear Information System (INIS)

    Clearfield, A.; Peng, G.Z.; Cahill, R.A.; Bellinghausen, P.; Aly, H.I.; Scott, K.; Wang, J.D.

    1993-01-01

    There is now a concerted effort underway to clean up the accumulated nuclear wastes as the major sites around the country. Because of the complexity of the mixtures in the holding tanks highly specific exchangers are required to fulfill a multitude of desired tasks. These include removal of Cs + , Sr 2+ , Tc, Actinides and possible recovery of rare and precious metals. No one exchanger or sequestrant can accomplish these tasks and a variety of exchangers in a multistep process will be required. The behavior of a number of inorganic ion exchangers in a multistep process will be required. The behavior of a number of inorganic ion exchangers and new organo-inorganic exchangers towards Cs + , Sr 2+ and rare-earth ions in acid and basic media will be described. Preliminary data on the effect of high levels of sodium nitrate on the uptake of these ions will also be presented, as well as the changes observed in selectivity in simulated waste solutions. A possible separation scheme based on these data will be described

  16. Ammonia nitrogen removal from aqueous solution by local agricultural wastes

    Science.gov (United States)

    Azreen, I.; Lija, Y.; Zahrim, A. Y.

    2017-06-01

    Excess ammonia nitrogen in the waterways causes serious distortion to environment such as eutrophication and toxicity to aquatic organisms. Ammonia nitrogen removal from synthetic solution was investigated by using 40 local agricultural wastes as potential low cost adsorbent. Some of the adsorbent were able to remove ammonia nitrogen with adsorption capacity ranging from 0.58 mg/g to 3.58 mg/g. The highest adsorption capacity was recorded by Langsat peels with 3.58 mg/g followed by Jackfruit seeds and Moringa peels with 3.37 mg/g and 2.64 mg/g respectively. This experimental results show that the agricultural wastes can be utilized as biosorbent for ammonia nitrogen removal. The effect of initial ammonia nitrogen concentration, pH and stirring rate on the adsorption process were studied in batch experiment. The adsorption capacity reached maximum value at pH 7 with initial concentration of 500 mg/L and the removal rate decreased as stirring rate was applied.

  17. Defense High Level Waste Disposal Container System Description Document

    International Nuclear Information System (INIS)

    Pettit, N. E.

    2001-01-01

    The Defense High Level Waste Disposal Container System supports the confinement and isolation of waste within the Engineered Barrier System of the Monitored Geologic Repository (MGR). Disposal containers are loaded and sealed in the surface waste handling facilities, transferred to the underground through the accesses using a rail mounted transporter, and emplaced in emplacement drifts. The defense high level waste (HLW) disposal container provides long-term confinement of the commercial HLW and defense HLW (including immobilized plutonium waste forms [IPWF]) placed within disposable canisters, and withstands the loading, transfer, emplacement, and retrieval loads and environments. US Department of Energy (DOE)-owned spent nuclear fuel (SNF) in disposable canisters may also be placed in a defense HLW disposal container along with commercial HLW waste forms, which is known as co-disposal. The Defense High Level Waste Disposal Container System provides containment of waste for a designated period of time, and limits radionuclide release. The disposal container/waste package maintains the waste in a designated configuration, withstands maximum handling and rockfall loads, limits the individual canister temperatures after emplacement, resists corrosion in the expected handling and repository environments, and provides containment of waste in the event of an accident. Defense HLW disposal containers for HLW disposal will hold up to five HLW canisters. Defense HLW disposal containers for co-disposal will hold up to five HLW canisters arranged in a ring and one DOE SNF canister inserted in the center and/or one or more DOE SNF canisters displacing a HLW canister in the ring. Defense HLW disposal containers also will hold two Multi-Canister Overpacks (MCOs) and two HLW canisters in one disposal container. The disposal container will include outer and inner cylinders, outer and inner cylinder lids, and may include a canister guide. An exterior label will provide a means by

  18. Separation of palladium from high-level waste using metal ferro cyanide loaded resins

    International Nuclear Information System (INIS)

    Valsala, T.P.; Joseph, Annie; Yeotikar, R.G.

    2005-01-01

    High-level waste (HLW) is generated during reprocessing of spent fuel. HLW contains corrosion products, unextracted actinides, process chemicals and fission products. A recent trend is there to consider waste as a source of wealth. Among the fission products separation and recovery of platinum group metals have gained great attention. HLW is a good source of palladium of the platinum group metal. The present study shows the feasibility of ion exchange separation of Pd from HLW. (author)

  19. 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

  20. Mixing of process heels, process solutions, and recycle streams: Results of the small-scale radioactive tests

    International Nuclear Information System (INIS)

    Lumetta, G.J.; Bramson, J.P.; Farmer III, O.T.; Greenwood, L.R.; Hoopes, F.V.; Mann, M.A.; Steele, M.J.; Steele, R.T.; Swoboda, R.G.; Urie, M.W.

    2000-01-01

    Various recycle streams will be combined with the low-activity waste (LAW) or the high-level waste (HLW) feed solutions during the processing of the Hanford tank wastes by BNFL, Inc. In addition, the LAW and HLW feed solutions will also be mixed with heels present in the processing equipment. This report describes the results of a test conducted by Battelle to assess the effects of mixing specific process streams. Observations were made regarding adverse reactions (mainly precipitation) and effects on the Tc oxidation state (as indicated by K d measurements with SuperLigreg s ign 639). The work was conducted according to test plan BNFL-TP-29953-023, Rev. 0, Small Scale Mixing of Process Heels, Solutions, and Recycle Streams. The test went according to plan, with only minor deviations from the test plan. The deviations from the test plan are discussed in the experimental section

  1. 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.

  2. 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)

  3. Chemistry of application of calcination/dissolution to the Hanford tank waste inventory

    International Nuclear Information System (INIS)

    Delegard, C.H.; Elcan, T.D.; Hey, B.E.

    1994-05-01

    Approximately 330,000 metric tons of sodium-rich radioactive waste originating from separation of plutonium from irradiated uranium fuel are stored in underground tanks at the Hanford Site in Washington State. Fractionation of the waste into low-level waste (LLW) and high-level waste (HLW) streams is envisioned via partial water dissolution and limited radionuclide extraction operations. Under optimum conditions, LLW would contain most of the chemical bulk while HLW would contain virtually all of the transuranic and fission product activity. Calcination at around 850 C, followed by water dissolution, has been proposed as an alternative initial treatment of Hanford Site waste to improve waste dissolution and the envisioned LLW/HLW split. Results of literature and laboratory studies are reported on the application of calcination/dissolution (C/D) to the fractionation of the Hanford Site tank waste inventory. Both simulated and genuine Hanford Site waste materials were used in the lab tests. To evaluation confirmed that C/D processing reduced the amount of several components from the waste. The C/D dissolutions of aluminum and chromium allow redistribution of these waste components from the HLW to the LLW fraction. Comparisons of simple water-washing with C/D processing of genuine Hanford Site waste are also reported based on material (radionuclide and chemical) distributions to solution and solid residue phases. The lab results show that C/D processing yielded superior dissolution of aluminum and chromium sludges compared to simple water dissolution. 57 refs., 26 figs., 18 tabs

  4. Chemistry of application of calcination/dissolution to the Hanford tank waste inventory

    Energy Technology Data Exchange (ETDEWEB)

    Delegard, C.H.; Elcan, T.D.; Hey, B.E.

    1994-05-01

    Approximately 330,000 metric tons of sodium-rich radioactive waste originating from separation of plutonium from irradiated uranium fuel are stored in underground tanks at the Hanford Site in Washington State. Fractionation of the waste into low-level waste (LLW) and high-level waste (HLW) streams is envisioned via partial water dissolution and limited radionuclide extraction operations. Under optimum conditions, LLW would contain most of the chemical bulk while HLW would contain virtually all of the transuranic and fission product activity. Calcination at around 850 C, followed by water dissolution, has been proposed as an alternative initial treatment of Hanford Site waste to improve waste dissolution and the envisioned LLW/HLW split. Results of literature and laboratory studies are reported on the application of calcination/dissolution (C/D) to the fractionation of the Hanford Site tank waste inventory. Both simulated and genuine Hanford Site waste materials were used in the lab tests. To evaluation confirmed that C/D processing reduced the amount of several components from the waste. The C/D dissolutions of aluminum and chromium allow redistribution of these waste components from the HLW to the LLW fraction. Comparisons of simple water-washing with C/D processing of genuine Hanford Site waste are also reported based on material (radionuclide and chemical) distributions to solution and solid residue phases. The lab results show that C/D processing yielded superior dissolution of aluminum and chromium sludges compared to simple water dissolution. 57 refs., 26 figs., 18 tabs.

  5. 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)

  6. Spray drying test of simulated borated waste solutions

    International Nuclear Information System (INIS)

    An Hongxiang; Zhou Lianquan; Fan Zhiwen; Sun Qi; Lin Xiaolong

    2007-01-01

    Performance and the effecting factors of spray drying of simulated borated waste solutions is studied for three contaeting methods between the atomized beads and the heated air, in which boron concentration is around 21000 ppm. The contacting modes are centrifugal atomizing co-current flow, pneumatic atomizing co-current flow and mixed flow. The results show that a free-flowing product in all these tests when the temperature of the solutions is between 62 degree C and 64 degree C, the inlet temperature of the spray drying chamber is between 210 degree C and 220 degree C, the temperature of the outlet of the spray drying chamber is between 110 and 120 degree C, the flow rate of the pressure air is 8.0 m 3 /h, the rotational speed of the centrifugal atomizer is 73.0 m/s. The diameters of the powder product which account for 95% of the feed range from 0.356 mm to 0.061 mm. The production capacity and water content in the powder increase in the order of pneumatic atomizing co-current flow, mixed flow and centrifugal atomizing co-current flow. The volume reduction coeffecient of spray drying is in the ranged of 0.22 and 0.27. (authors)

  7. 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)

  8. Immobilization of defense high-level waste: an assessment of technological strategies and potential regulatory goals. Volume II

    International Nuclear Information System (INIS)

    1979-06-01

    This volume contains the following appendices: selected immobilization processes, directory of selected European organizations involved in HLW management, U.S. high-level waste inventories, and selected European HLW program

  9. Policy and practice of radioactive waste management in India

    International Nuclear Information System (INIS)

    Sunder Radzhan, N.S.

    1986-01-01

    The Indian program on radioactive waste management comprising two main variants: engineering subsurface repositories for low- and intermediate-level wastes and deep geological formations for alpha-bearing and high-level wastes (HLW) is presented. One of the problems deals with the matrices with improved properties for HLW inclusion. The other aspect concerns development of management with alpha-emitting radionuclides in HLW. Special attention is paid to the problems of safety

  10. 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.

  11. 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.

  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. 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

  14. 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.

  15. 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

  16. Solute transport in fractured rock - applications to radionuclide waste repositories

    International Nuclear Information System (INIS)

    Neretnieks, I.

    1990-12-01

    Flow and solute transport in fractured rocks has been intensively studied in the last decade. The increased interest is mainly due to the plans in many countries to site repositories for high level nuclear waste in deep geologic formations. All investigated crystalline rocks have been found to be fractured and most of the water flows in the fractures and fracture zones. The water transports dissolved species and radionuclides. It is thus of interest to be able to understand and to do predictive modelling of the flowrate of water, the flowpaths and the residence times of the water and of the nuclides. The dissolved species including the nuclides will interact with the surrounding rock in different ways and will in many cases be strongly retarded relative to the water velocity. Ionic species may be ion exchanged or sorbed in the mineral surfaces. Charges and neutral species may diffuse into the stagnant waters in the rock matrix and thus be withdrawn from the mobile water. These effects will be strongly dependent on how much rock surface is in contact with the flowing water. It has been found in a set of field experiments and by other observations that not all fractures conduct water. Furthermore it is found that conductive fractures only conduct the water in a small part of the fracture in what is called channels or preferential flowpaths. This report summarizes the present concepts of water flow and solute transport in fractured rocks. The data needs for predictive modelling are discussed and both field and laboratory measurement which have been used to obtain data are described. Several large scale field experiments which have been specially designed to study flow and tracer transport in crystalline rocks are described. In many of the field experients new techniques have been developed and used. (81 refs.) (author)

  17. Removal of radioactive ions from nuclear waste solutions by electrodialysis

    Energy Technology Data Exchange (ETDEWEB)

    Sugimoto, S [Radia Industries Co. Ltd., Takasaki, Gunma (Japan)

    1978-10-01

    Removal of radioactive ions was studied from low and medium level radioactive waste solutions by electrodialysis using ion exchange membranes. The test solutions contained /sup 137/Cs/sup +/, /sup 106/Ru/sup 3 +/ or fission products (F.P.) as active ions and NaCl, Na/sub 2/SO/sub 4/ or Ca(NO/sub 3/)/sub 2/ as inactive coexisting salts. The decontamination factor of the active ions was in the order: /sup 137/Cs/sup +/ (greater than 99%) > /sup 90/Sr/sup 2 +/ > F.P. > /sup 106/Ru/sup 3 +/. The dialysis time required to attain the saturation was the shortest for monovalent cations K/sup +/, Cs/sup +/ and Na/sup +/, intermediate for divalent cation Sr/sup 2 +/, and the longest for trivalent cation Ru/sup 3 +/. The ratio of the decontamination factor of an active ion eta sub( a) to the desalination factor of an inactive ion eta sub( b) was nearly equal to unity for /sup 24/Na, /sup 42/K, /sup 137/Cs and /sup 90/Sr. On the other hand, the apparent selective permeability of an active ion (A/sup +/) against Na/sup +/ ion, T sub(Na/sup +/) sup( a) was higher than unity for all the active ions tested, and was in the order of /sup 137/Cs > /sup 90/Sr > /sup 42/K > /sup 24/Na, where T sub(Na/sup +/) sup( a) is defined by the ratio of ..gamma..sub( a) to ..gamma..sub(Na/sup +/) with ..gamma..sub( a) being the ratio of dilution of A in the diluate the ..gamma..sub(Na/sup +/) being that of Na/sup +/ in the same diluate. The decontamination factor of the active ions did not depend significantly on the species and concentration of the coexistent salts or on the concentration of the active ions.

  18. 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

  19. Development of ICP-AES based method for the characterization of high level waste

    International Nuclear Information System (INIS)

    Seshagiri, T.K.; Thulsidas, S.K.; Adya, V.C.; Kumar, Mithlesh; Radhakrishnan, K.; Mary, G.; Kulkarni, P.G.; Bhalerao, Bharti; Pant, D.K.

    2011-01-01

    An Inductively Coupled Plasma Atomic Emission Spectrometry (ICP-AES) method was developed for the trace metal characterization of high level waste solutions (HLW) of different origin and the method was validated by analysis of synthetic samples of simulated high level waste solutions (SHLW) from spent fuels of varying composition. In this context, an inter-laboratory comparison exercise (ILCE) was carried out with the simulated HLW of different spent fuel types, viz., research reactor (RR), pressurized heavy water reactor (PHWR) and fast breeder reactor (FBR). An over view of the ICP-AES determination of trace metallic constituents in such SHLW solutions is presented. The overall agreement between the various laboratories was good. (author)

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  2. 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)

  3. Institute for Nuclear Waste Disposal. Annual Report 2011

    International Nuclear Information System (INIS)

    Geckeis, H.; Stumpf, T.

    2012-01-01

    The R and D at the Institute for Nuclear Waste Disposal, INE, (Institut fuer Nukleare Entsorgung) of the Karlsruhe Institute of Technology (KIT) focuses on (i) long term safety research for nuclear waste disposal, (ii) immobilization of high level radioactive waste (HLW), (iii) separation of minor actinides from HLW and (iv) radiation protection.

  4. Chemical activation of tea waste and use for the removal of chromium (Vi) from aqueous solution

    International Nuclear Information System (INIS)

    Qureshi, K.; Bhatti, I.; Ansari, A.K.

    2009-01-01

    Tea waste is the residue left after the preparation of tea. At present the tea waste is regarded as a waste product having no use. In this study, tea waste is converted into an adsorbent. Tea waste is chemically activated with phosphoric acid at low temperature 450 degree C. This activated carbon is then utilized as an adsorbent for the removal of Chromium (VI) from aqueous solution. The various sorption parameters i.e pH, sorbent dose sorbate concentration, shaking time and shaking speed are first optimized. 75% of chromium from aqueous solution is effectively removed at pH 2. The best optimum conditions were obtained when 1 gm of sorbent was agitated at 100 rpm with 60 mg/l of sorbate for 50 minutes. Better results were obtained when low concentrations of sorbates were used. Hence tea waste could also be successfully used for the sorption of Chromium (VI), from industrial waste water. (author)

  5. 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

  6. Mining techniques and some aspects of high-level waste disposal

    International Nuclear Information System (INIS)

    Hoefnagels, J.A.R.

    1980-01-01

    The solutions to many problems of underground waste disposal involve mine engineering. This article attempts to highlight chosen issues and thereby create an overall impression, avoiding emphasis on single-aspect calculation. High level waste (H.L.W.) dominates current radioactive waste studies because of its specific characteristics and is therefore dealt with in this paper. However, depending on the method of disposal the other categories of radio active waste might become problems by themselves because of the relatively large quantities involved. (Auth.)

  7. Chemical and Nuclear Waste Disposal: Problems and Solutions

    OpenAIRE

    James P. Murray; Joseph J. Harrington; Richard Wilson

    1982-01-01

    The problems of waste disposal have always been with us. In biblical times, the residents of Jerusalem always burnt their wastes inthehideousValeofGehenna.Thisgavewaytoburialofwasteor sometimes dumping it in shallow oceans. All too often the sewage pipes of the seaside towns did not even take the waste to the low tide mark; and the use of the deep oceans as a disposal site has been almost unknown...

  8. Municipal solid waste management. Strategies and technologies for sustainable solutions

    Energy Technology Data Exchange (ETDEWEB)

    Ludwig, C.; Hellweg, S.; Stucki, S. (eds.)

    2002-10-01

    The way municipal solid waste is handled greatly determines its impact on the local as well as the global environment. New technologies habe emerged for the treatment of waste, for the recovery of raw materials and energy, and for safe final disposal. The environmental performance of technologies, their social acceptance and their economic viability are key issues to be considered in sustainable waste management. This book provides an overview of current practices in waste management and a synthesis of new developments achieved through interdisciplinary discussions of recent research results. (orig.)

  9. 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)

  10. 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

  11. 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

  12. Electrochemical processing of nitrate waste solutions. Phase 2, Final report

    Energy Technology Data Exchange (ETDEWEB)

    Genders, D.; Weinberg, N.; Hartsough, D. [Electrosynthesis Co., Inc., Cheektowaga, NY (US)

    1992-10-07

    The second phase of research performed at The Electrosynthesis Co., Inc. has demonstrated the successful removal of nitrite and nitrate from a synthetic effluent stream via a direct electrochemical reduction at a cathode. It was shown that direct reduction occurs at good current efficiencies in 1,000 hour studies. The membrane separation process is not readily achievable for the removal of nitrites and nitrates due to poor current efficiencies and membrane stability problems. A direct reduction process was studied at various cathode materials in a flow cell using the complete synthetic mix. Lead was found to be the cathode material of choice, displaying good current efficiencies and stability in short and long term tests under conditions of high temperature and high current density. Several anode materials were studied in both undivided and divided cell configurations. A divided cell configuration was preferable because it would prevent re-oxidation of nitrite by the anode. The technical objective of eliminating electrode fouling and solids formation was achieved although anode materials which had demonstrated good stability in short term divided cell tests corroded in 1,000 hour experiments. The cause for corrosion is thought to be F{sup {minus}} ions from the synthetic mix migrating across the cation exchange membrane and forming HF in the acid anolyte. Other possibilities for anode materials were explored. A membrane separation process was investigated which employs an anion and cation exchange membrane to remove nitrite and nitrate, recovering caustic and nitric acid. Present research has shown poor current efficiencies for nitrite and nitrate transport across the anion exchange membrane due to co-migration of hydroxide anions. Precipitates form within the anion exchange membranes which would eventually result in the failure of the membranes. Electrochemical processing offers a highly promising and viable method for the treatment of nitrate waste solutions.

  13. 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)

  14. Recent studies of uranium and plutonium chemistry in alkaline radioactive waste solutions

    International Nuclear Information System (INIS)

    King, William D.; Wilmarth, William R.; Hobbs, David T.; Edwards, Thomas B.

    2008-01-01

    Solubility studies of uranium and plutonium in a caustic, radioactive Savannah River Site tank waste solution revealed the existence of uranium supersaturation in the as-received sample. Comparison of the results to predictions generated from previously published models for solubility in these waste types revealed that the U model poorly predicts solubility while Pu model predictions are quite consistent with experimental observations. Separate studies using simulated Savannah River Site evaporator feed solution revealed that the known formation of sodium aluminosilicate solids in waste evaporators can promote rapid precipitation of uranium from supersaturated solutions

  15. Sorption Potentials of Waste Tyre for Some Heavy Metals (Pb Cd in Aqueous Solution

    Directory of Open Access Journals (Sweden)

    Austin Kanayo ASIAGWU

    2009-07-01

    Full Text Available An investigation into the adsorption potential of activated and inactivated waste tyre powders for some heavy metals (Pb2+ and Cd2+ in their aqueous solution has been studied. The result indicated that inactivated waste tyre is a good non-conventional adsorbent for the removal of Cd from aqueous solution. A total of 93.3% of Cadmium contents was removed. The inactivated waste type proved a good adsorbent for the removal of Pb2+ 5g of 500mm activated tyre removed over 86.66% of Pb2+ from solution.

  16. Alternative solutions for the disposal of radioactive wastes

    International Nuclear Information System (INIS)

    Ramsey, R.W. Jr.

    1976-01-01

    Besides outlining the possibility of dispatching concentrated highly radioactive waste by rockets into space, or of transmuting long-lived isotopes by nuclear reactions into short-lived ones, the author discusses further alternatives for the disposal of radioactive wastes, especially the storage in geologic formations. (HR/LN) [de

  17. Material resources, energy, and nutrient recovery from waste: are waste refineries the solution for the future?

    DEFF Research Database (Denmark)

    Tonini, Davide; Martinez-Sanchez, Veronica; Astrup, Thomas Fruergaard

    2013-01-01

    Waste refineries focusing on multiple outputs of material resources, energy carriers, and nutrients may potentially provide more sustainable utilization of waste resources than traditional waste technologies. This consequential life cycle assessment (LCA) evaluated the environmental performance....... Overall, the waste refinery provided global warming (GW) savings comparable with efficient incineration, MBT, and bioreactor landfilling technologies. The main environmental benefits from waste refining were a potential for improved phosphorus recovery (about 85%) and increased electricity production (by...

  18. Defense Waste Processing Facility Recycle Stream Evaporation

    International Nuclear Information System (INIS)

    STONE, MICHAEL

    2006-01-01

    The Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS) stabilizes high level radioactive waste (HLW) by vitrification of the waste slurries. DWPF currently produces approximately five gallons of dilute recycle for each gallon of waste vitrified. This recycle stream is currently sent to the HLW tank farm at SRS where it is processed through the HLW evaporators with the concentrate eventually sent back to the DWPF for stabilization. Limitations of the HLW evaporators and storage space constraints in the tank farm have the potential to impact the operation of the DWPF and could limit the rate that HLW is stabilized. After an evaluation of various alternatives, installation of a dedicated evaporator for the DWPF recycle stream was selected for further evaluation. The recycle stream consists primarily of process condensates from the pretreatment and vitrification processes. Other recycle streams consist of process samples, sample line flushes, sump flushes, and cleaning solutions from the decontamination and filter dissolution processes. The condensate from the vitrification process contains some species, such as sulfate, that are not appreciably volatile at low temperature and could accumulate in the system if 100% of the evaporator concentrate was returned to DWPF. These species are currently removed as required by solids washing in the tank farm. The cleaning solutions are much higher in solids content than the other streams and are generated 5-6 times per year. The proposed evaporator would be required to concentrate the recycle stream by a factor of 30 to allow the concentrate to be recycled directly to the DWPF process, with a purge stream sent to the tank farm as required to prevent buildup of sulfate and similar species in the process. The overheads are required to meet stringent constraints to allow the condensate to be sent directly to an effluent treatment plant. The proposed evaporator would nearly de-couple the DWPF process from the

  19. Commercial waste treatment program annual progress report for FY 1983

    Energy Technology Data Exchange (ETDEWEB)

    McElroy, J.L.; Burkholder, H.C. (comps.)

    1984-02-01

    This annual report describes progress during FY 1983 relating to technologies under development by the Commercial Waste Treatment Program, including: development of glass waste form and vitrification equipment for high-level wastes (HLW); waste form development and process selection for transuranic (TRU) wastes; pilot-scale operation of a radioactive liquid-fed ceramic melter (LFCM) system for verifying the reliability of the reference HLW treatment proces technology; evaluation of treatment requirements for spent fuel as a waste form; second-generation waste form development for HLW; and vitrification process control and product quality assurance technologies.

  20. 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

  1. 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

  2. Potential use of maize waste for the removal of Pb(II) from aqueous solution

    CSIR Research Space (South Africa)

    Okonkwo, J

    2006-09-01

    Full Text Available batch adsorption procedures. The utilization of tassels for the removal of toxic heavy metals from effluent solutions would, however, attach some economic value to this waste material. Tassel flowers were collected just prior to harvest, dried under...

  3. Determination of plutonium 241 in solutions of nuclear wastes

    International Nuclear Information System (INIS)

    Raymond, A.; Bilcot, J.B.; Poletiko, C.

    1990-09-01

    Determination of plutonium 241 in nuclear wastes is important because of long period and high energy of some daughter products. In this report are presented two quantitative analysis methods using both scintillation techniques: A complete method, in any case, by selective extraction of plutonium on an anionic resin allowing simultaneous determination of Pu 241 and the sum of other plutonium isotopes; a simplified method when alpha activity is higher than beta/gamma activity by liquid extraction with TTA. These methods are applied for analysis of 4 waste types: cement encapsulated wastes, bitumen encapsulated wastes, incineration ashes, leaching of encapsulated incineration ashes. In these 4 examples, Pu 241 activity is equal or higher than the sum of alpha plutonium isotope activity. Separation efficiency, measured from Pu 239 or with Pu 236 as tracer, is between 90 and 99% [fr

  4. Decomposition Technology Development of Organic Component in a Decontamination Waste Solution

    International Nuclear Information System (INIS)

    Jung, Chong Hun; Oh, W. Z.; Won, H. J.; Choi, W. K.; Kim, G. N.; Moon, J. K.

    2007-11-01

    Through the project of 'Decomposition Technology Development of Organic Component in a Decontamination Waste Solution', the followings were studied. 1. Investigation of decontamination characteristics of chemical decontamination process 2. Analysis of COD, ferrous ion concentration, hydrogen peroxide concentration 3. Decomposition tests of hardly decomposable organic compounds 4. Improvement of organic acid decomposition process by ultrasonic wave and UV light 5. Optimization of decomposition process using a surrogate decontamination waste solution

  5. Decomposition Technology Development of Organic Component in a Decontamination Waste Solution

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Chong Hun; Oh, W. Z.; Won, H. J.; Choi, W. K.; Kim, G. N.; Moon, J. K

    2007-11-15

    Through the project of 'Decomposition Technology Development of Organic Component in a Decontamination Waste Solution', the followings were studied. 1. Investigation of decontamination characteristics of chemical decontamination process 2. Analysis of COD, ferrous ion concentration, hydrogen peroxide concentration 3. Decomposition tests of hardly decomposable organic compounds 4. Improvement of organic acid decomposition process by ultrasonic wave and UV light 5. Optimization of decomposition process using a surrogate decontamination waste solution.

  6. Partitioning high-level waste from alkaline solution: A literature survey

    International Nuclear Information System (INIS)

    Marsh, S.F.

    1993-05-01

    Most chemical partitioning procedures are designed for acidic feed solutions. However, the high-level waste solutions in the underground storage tanks at US Department of Energy defense production sites are alkaline. Effective partitioning procedures for alkaline solutions could decrease the need to acidify these solutions and to dissolve the solids in acid, which would simplify subsequent processing and decrease the generation of secondary waste. The author compiles candidate technologies from his review of the chemical literature, experience, and personal contacts. Several of these are recommended for evaluation

  7. Selective separation of radionuclides from nuclear waste solutions with inorganic ion exchangers

    International Nuclear Information System (INIS)

    Lehto, J.; Harjula, R.

    1999-01-01

    Nuclear industry produces and stores large volumes of radioactive waste solutions. Removal of radionuclides from the solutions is an important and challenging task for two main reasons: reductions in the volumes of solidified waste, which have to be disposed of, and reductions in the radioactive discharges into the environment. Since the radioactive elements in most waste solutions are in trace concentrations and the waste solutions contain large excesses of inactive metal ions, highly selective separation methods are needed for the removal of radionuclides. A number of inorganic ion exchange materials are very selective to key radionuclides and they can play an important role in solving these problems. The spectrum of nuclear waste solutions is rather wide considering their radionuclide contents, concentrations of interfering salts and acidity/alkalinity. Therefore, several inorganic ions exchangers are needed for the removal of most harmful radionuclides from a variety of solutions. This paper discusses the use and requirements of inorganic ion exchange materials in nuclear waste management. Special attention is paid to the novel ion exchange materials developed in the Laboratory of Radiochemistry, University of Helsinki. (orig.)

  8. Development of integrated radioactive waste packaging and conditioning solutions in the UK

    Energy Technology Data Exchange (ETDEWEB)

    Sibley, Peter; Butter, Kevin; Zimmerman, Ian [EnergySolutions EU Ltd., Swindon, Wiltshire (United Kingdom); Viermann, Joerg [GNS Gesellschaft fur Nuklear-Service mbH, Essen (Germany); Messer, Matthias [GNS Gesellschaft fur Nuklear-Service mbH, Bristol (United Kingdom)

    2013-07-01

    In order to offer a more cost effective, safer and efficient Intermediate Level Waste (ILW) management service, EnergySolutions EU Ltd. and Gesellschaft fur Nuklear-Service mbH (GNS) have been engaged in the development of integrated radioactive waste retrieval, packaging and conditioning solutions in the UK. Recognising the challenges surrounding regulatory endorsement and on-site implementation in particular, this has resulted in an alternative approach to meeting customer, safety regulator and disposability requirements. By working closely with waste producers and the organisation(s) responsible for endorsing radioactive waste management operations in the UK, our proposed solutions are now being implemented. By combining GNS' off-the-shelf, proven Ductile Cast Iron Containers (DCICs) and water removal technologies, with EnergySolutions EU Ltd.'s experience and expertise in waste retrieval, safety case development and disposability submissions, a fully integrated service offering has been developed. This has involved significant effort to overcome technical challenges such as onsite equipment deployment, active commissioning, conditioning success criteria and disposability acceptance. Our experience in developing such integrated solutions has highlighted the importance of working in collaboration with all parties to achieve a successful and viable outcome. Ultimately, the goal is to ensure reliable, safe and effective delivery of waste management solutions. (authors)

  9. Partitioning and recovery of neptunium from high level waste streams of PUREX origin using 30% TBP

    International Nuclear Information System (INIS)

    Mathur, J.N.; Murali, M.S.; Balarama Krishna, M.V.; Iyer, R.H.; Chitnis, R.R.; Wattal, P.K.; Theyyunni, T.K.; Ramanujam, A.; Dhami, P.S.; Gopalakrishnan, V.

    1995-01-01

    237 Np is one of the longest-lived nuclides among the actinides present in the high level waste solutions of reprocessing origin. Its separation, recovery and transmutation can reduce the problem of long term storage of the vitrified waste to a great extent. With this objective, the present work was initiated to study the extraction of neptunium into TBP under the conditions relevant to high level waste, along with uranium and plutonium by oxidising it to hexavalent state using potassium dichromate and subsequently recovering it by selective stripping. Three types of simulated HLW solutions namely sulphate bearing (SB), with an acidity of ∼ 0.3 M and non-sulphate wastes originating from the reprocessing of fuels from pressurised heavy water reactor (PHWR) and fast breeder reactor (FBR) with acidities of 3.0 M HNO 3 were employed in these studies. The extraction of U(VI), Np(VI) and Pu(VI) was very high for PHWR- and FBR-HLW solutions, whereas for the SB-HLW solution, these values were less but reasonably high. Quantitative recovery of neptunium and plutonium was achieved using a stripping solution containing 0.1 M H 2 O 2 and 0.01 M ascorbic acid at an acidity of 2.0 M. Since, cerium present in the waste solutions is expected to undergo oxidation in presence of K 2 Cr 2 O 7 , its extraction behaviour was also studied under similar conditions. Based on the results, a scheme was formulated for the recovery of neptunium along with plutonium and was successfully applied to actual high level waste solution originating from the reprocessing of research reactor fuels. (author). 19 refs., 2 figs., 17 tabs

  10. 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)

  11. removal of hazardous pollutants from industrial waste solutions using membrane techniques

    International Nuclear Information System (INIS)

    Selim, Y.T.M.

    2001-01-01

    the removal of hazardous pollutants from industrial waste solutions is of essential demand field for both scientific and industrial work. the present work includes detailed studies on the possible use of membrane technology especially liquid emulsion membrane for the removal of hazardous pollutants such as; cadmium , cobalt , lead, copper and uranium from different industrial waste solution . this research can be applied for mixed waste problems. the work carried out in this thesis is presented in three main chapters, namely introduction, experimental and results and discussion

  12. 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)

  13. Removal of radionuclides from partitioning waste solutions by adsorption and catalytic oxidation methods

    Energy Technology Data Exchange (ETDEWEB)

    Yamagishi, Isao; Yamaguchi, Isoo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kubota, Masumitsu [Research Organization for Information Science and Technology (RIST), Tokai, Ibaraki (Japan)

    2000-09-01

    Adsorption of radionuclides with inorganic ion exchangers and catalytic oxidation of a complexant were studied for the decontamination of waste solutions generated in past partitioning tests with high-level liquid waste. Granulated ferrocyanide and titanic acid were used for adsorption of Cs and Sr, respectively, from an alkaline solution resulting from direct neutralization of an acidic waste solution. Both Na and Ba inhibited adsorption of Sr but Na did not that of Cs. These exchangers adsorbed Cs and Sr at low concentration with distribution coefficients of more than 10{sup 4}ml/g from 2M Na solution of pH11. Overall decontamination factors (DFs) of Cs and total {beta} nuclides exceeded 10{sup 5} and 10{sup 3}, respectively, at the neutralization-adsorption step of actual waste solutions free from a complexant. The DF of total {alpha} nuclides was less than 10{sup 3} for a waste solution containing diethylenetriaminepentaacetic acid (DTPA). DTPA was rapidly oxidized by nitric acid in the presence of a platinum catalyst, and radionuclides were removed as precipitates by neutralization of the resultant solution. The DF of {alpha} nuclides increased to 8x10{sup 4} by addition of the oxidation step. The DFs of Sb and Co were quite low through the adsorption step. A synthesized Ti-base exchanger (PTC) could remove Sb with the DF of more than 4x10{sup 3}. (author)

  14. Chemical hazards from decontamination solutions in low level waste

    International Nuclear Information System (INIS)

    Leventhal, L.; Miller, A.; Turney, J.; Naughton, M.; IMPELL Corp., Walnut Creek, CA; Electric Power Research Inst., Palo Alto, CA)

    1985-01-01

    Recent regulations are focussing more attention on the non-radioactive matrix materials associated with radioactive wastes. Decontamination of operating facilities is becoming a more significant source of low-level waste. This study reviewed the chemical and biological hazards of over 50 decontamination processes. Seventeen of the most prominent hard and soft decontamination processes were examined in detail. The chemical and biological hazards of these seventeen are presented in this paper. These hazards influence the choice of radwaste processing and packaging operations and methods. Federal, state and local regulations further impact on operations and waste disposal. Hazards to personnel, in plant and off-site, resulting from the decontamination cycle are evaluated. 1 fig., 5 tabs

  15. DEVELOPMENT OF A KINETIC MODEL OF BOEHMITE DISSOLUTION IN CAUSTIC SOLUTIONS APPLIED TO OPTIMIZE HANFORD WASTE PROCESSING

    International Nuclear Information System (INIS)

    Disselkamp, R.S.

    2011-01-01

    Boehmite (e.g., aluminum oxyhydroxide) is a major non-radioactive component in Hanford and Savannah River nuclear tank waste sludge. Boehmite dissolution from sludge using caustic at elevated temperatures is being planned at Hanford to minimize the mass of material disposed of as high-level waste (HLW) during operation of the Waste Treatment Plant (WTP). To more thoroughly understand the chemistry of this dissolution process, we have developed an empirical kinetic model for aluminate production due to boehmite dissolution. Application of this model to Hanford tank wastes would allow predictability and optimization of the caustic leaching of aluminum solids, potentially yielding significant improvements to overall processing time, disposal cost, and schedule. This report presents an empirical kinetic model that can be used to estimate the aluminate production from the leaching of boehmite in Hanford waste as a function of the following parameters: (1) hydroxide concentration; (2) temperature; (3) specific surface area of boehmite; (4) initial soluble aluminate plus gibbsite present in waste; (5) concentration of boehmite in the waste; and (6) (pre-fit) Arrhenius kinetic parameters. The model was fit to laboratory, non-radioactive (e.g. 'simulant boehmite') leaching results, providing best-fit values of the Arrhenius A-factor, A, and apparent activation energy, E A , of A = 5.0 x 10 12 hour -1 and E A = 90 kJ/mole. These parameters were then used to predict boehmite leaching behavior observed in previously reported actual waste leaching studies. Acceptable aluminate versus leaching time profiles were predicted for waste leaching data from both Hanford and Savannah River site studies.

  16. Waste-to-energy plants - a solution for a cleaner future

    International Nuclear Information System (INIS)

    Pfeiffer, J.

    2007-01-01

    Waste-to-energy plants reduce the municipal solid waste volume by about 80% and convert it into residue. The residue quality naturally depends on the burned waste quality and also on the combustion parameters. Hence, tighter control of the plant can improve the residue quality. The generated combustion energy is regarded as renewable energy and is typically used to feed a turbine to generate electricity. Waste-to-energy furnaces react slowly on changing waste charge, so they are not used for peak load generation. The generated electrical power is a plant by product and is sold as base load generation. Usually the waste is burned on a grate which limits the plant size to about 160,000 tons of waste per year or 20 tons of waste per hour or about 28 MW. More recent technology utilizes fluidized bed combustion, which allows larger plant sizes up to 50 MW. Due to the unknown waste composition and stringent environmental standards involved, waste-to-energy plants employ sophisticated flue gas cleaning devices for emission control. ABB's Performance Monitoring continuously compares actual plant and equipment performance to expected performance. This includes the on-line calculation of the waste calorific heat allowing operator decision support and automated control system responses. Dedicated reports offer detailed data on operations, maintenance and emissions to plant management staff. ABB combustion optimization solutions use model based predictive control techniques to reliably find the most suitable set-points for improving the heat rate and reducing emissions like NO x . (author)

  17. 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.)

  18. River Protection Project Mission Analysis Waste Blending Study

    International Nuclear Information System (INIS)

    Shuford, D.H.; Stegen, G.

    2010-01-01

    Preliminary evaluation for blending Hanford site waste with the objective of minimizing the amount of high-level waste (HLW) glass volumes without major changes to the overall waste retrieval and processing sequences currently planned. The evaluation utilizes simplified spreadsheet models developed to allow screening type comparisons of blending options without the need to use the Hanford Tank Waste Operations Simulator (HTWOS) model. The blending scenarios evaluated are expected to increase tank farm operation costs due to increased waste transfers. Benefit would be derived from shorter operating time period for tank waste processing facilities, reduced onsite storage of immobilized HLW, and reduced offsite transportation and disposal costs for the immobilized HLW.

  19. 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

  20. 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)

  1. 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)

  2. 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

  3. Household Hazardous Waste: Everyone's Problem--Everyone's Solution.

    Science.gov (United States)

    Evenson, Linda

    1985-01-01

    Examines the household hazardous waste problem, addressing several areas related to regulation, disposal, and control. Also gives a list of safer alternatives for household cleaners/disinfectants, paint products, and pesticides. Indicates that individuals can collectively make a difference in public exposure by changing purchases and practices.…

  4. Construction materials as a waste management solution for cellulose sludge.

    Science.gov (United States)

    Modolo, R; Ferreira, V M; Machado, L M; Rodrigues, M; Coelho, I

    2011-02-01

    Sustainable waste management system for effluents treatment sludge has been a pressing issue for pulp and paper sector. Recycling is always recommended in terms of environmental sustainability. Following an approach of waste valorisation, this work aims to demonstrate the technical viability of producing fiber-cement roof sheets incorporating cellulose primary sludge generated on paper and pulp mills. From the results obtained with preliminary studies it was possible to verify the possibility of producing fiber-cement sheets by replacing 25% of the conventional used virgin long fiber by primary effluent treatment cellulose sludge. This amount of incorporation was tested on an industrial scale. Environmental parameters related to water and waste, as well as tests for checking the quality of the final product was performed. These control parameters involved total solids in suspension, dissolved salts, chlorides, sulphates, COD, metals content. In the product, parameters like moisture, density and strength were controlled. The results showed that it is possible to replace the virgin long fibers pulp by primary sludge without impacts in final product characteristics and on the environment. This work ensures the elimination of significant waste amounts, which are nowadays sent to landfill, as well as reduces costs associated with the standard raw materials use in the fiber-cement industrial sector. Copyright © 2010 Elsevier Ltd. All rights reserved.

  5. Construction materials as a waste management solution for cellulose sludge

    International Nuclear Information System (INIS)

    Modolo, R.; Ferreira, V.M.; Machado, L.M.; Rodrigues, M.; Coelho, I.

    2011-01-01

    Sustainable waste management system for effluents treatment sludge has been a pressing issue for pulp and paper sector. Recycling is always recommended in terms of environmental sustainability. Following an approach of waste valorisation, this work aims to demonstrate the technical viability of producing fiber-cement roof sheets incorporating cellulose primary sludge generated on paper and pulp mills. From the results obtained with preliminary studies it was possible to verify the possibility of producing fiber-cement sheets by replacing 25% of the conventional used virgin long fiber by primary effluent treatment cellulose sludge. This amount of incorporation was tested on an industrial scale. Environmental parameters related to water and waste, as well as tests for checking the quality of the final product was performed. These control parameters involved total solids in suspension, dissolved salts, chlorides, sulphates, COD, metals content. In the product, parameters like moisture, density and strength were controlled. The results showed that it is possible to replace the virgin long fibers pulp by primary sludge without impacts in final product characteristics and on the environment. This work ensures the elimination of significant waste amounts, which are nowadays sent to landfill, as well as reduces costs associated with the standard raw materials use in the fiber-cement industrial sector.

  6. 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.

  7. Steel corrosion resistance in model solutions and reinforced mortar containing wastes

    NARCIS (Netherlands)

    Koleva, D.A.; Van Breugel, K.

    2012-01-01

    This work reports on the corrosion resistance of steel in alkaline model solutions and in cement-based materials (mortar). The model solutions and the mortar specimens were Ordinary Portland Cement (OPC) based. Further, hereby discussed is the implementation of an eco-friendly approach of waste

  8. Immobilization and Waste Form Product Acceptance for Low Level and TRU Waste Forms

    International Nuclear Information System (INIS)

    Holtzscheiter, E.W.; Harbour, J.R.

    1998-05-01

    The Tanks Focus Area is supporting technology development in immobilization of both High Level (HLW) and Low Level (LLW) radioactive wastes. The HLW process development at Hanford and Idaho is patterned closely after that of the Savannah River (Defense Waste Processing Facility) and West Valley Sites (West Valley Demonstration Project). However, the development and options open to addressing Low Level Waste are diverse and often site specific. To start, it is important to understand the breadth of Low Level Wastes categories

  9. The incineration of solid radioactive waste: a centralized solution

    International Nuclear Information System (INIS)

    Hernborg, G.; Broden, K.; Oehrn, G.

    1985-01-01

    Almost all the combustible low-level β- and γ-radioactive waste from Sweden, and even some waste from German nuclear power plants, is treated in an incineration plant at Studsvik. To date most of the ash has been put into 100-litre drums, which in turn have been put in 200-litre drums with concrete in between. Recently, methods have been developed and equipment installed for homogeneous solidification of the ash into concrete. Over the years since the start-up of the plant in 1976 the incinerator has worked with a high availability factor. Personnel doses and activity releases to the environment are well below limits set by regulatory authorities. (orig.)

  10. Reduction of waste solution volume generated on electrokinetic remediation

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Gye-Nam; Koo, Dae-Seo; Kim, Seung-Soo; Jeong, Jung-Whan; Han, Gyu-Seong; Moon, Jei-Kwon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    In this study, for the reduction of volume of metal oxides generated in cathode chamber, the optimum pH of waste electrolyte in cathode chamber were drawn out through several experiments with the manufactured electrokinetic decontamination equipment. Also, the required time to reach to below the clearance concentration level for self- disposal was estimated through experiments using the manufactured electrokinetic decontamination equipment. A diagram of soil decontamination process for the removal of uranium from contaminated soil was drawn out. The optimum pH of waste electrolyte in cathode chamber for the reduction of volume of metal oxides was below 2.35. Also, when the initial uranium concentration of the soils were 7-20 Bq/g, the required times to reach to below the clearance concentration level for self- disposal were 25-40 days. A diagram of soil decontamination process for the removal of uranium from contaminated soil was drawn out.

  11. Copper-Sulfate Pentahydrate as a Product of the Waste Sulfuric Acid Solution Treatment

    OpenAIRE

    Marković, Radmila; Stevanović, Jasmina; Avramović, Ljiljana; Nedeljković, Dragutin; Jugović, Branimir; Stajić Trošić, Jasna; Gvozdenović, Milica M.

    2012-01-01

    The aim of this study is synthesis of copper-sulfate pentahydrate from the waste sulfuric acid solution-mother liquor generated during the regeneration process of copper bleed solution. Copper is removed from the mother liquor solution in the process of the electrolytic treatment using the insoluble lead anodes alloyed with 6 mass pct of antimony on the industrial-scale equipment. As the result of the decopperization process, copper is removed in the form of the cathode sludge and is precipit...

  12. Community Solutions for Solid Waste Pollution, Level 6. Teacher Guide. Operation Waste Watch.

    Science.gov (United States)

    Virginia State Dept. of Waste Management, Richmond. Div. of Litter & Recycling.

    Operation Waste Watch is a series of seven sequential learning units which addresses the subject of litter control and solid waste management. Each unit may be used in a variety of ways, depending on the needs and schedules of individual schools, and may be incorporated into various social studies, science, language arts, health, mathematics, and…

  13. Cementation of the solid radioactive waste with polymer-cement solutions using the method of impregnation

    International Nuclear Information System (INIS)

    Gorbunova, O.

    2015-01-01

    Cementation of solid radioactive waste (SRW), i.e. inclusion of solid radioactive waste into cement matrix without cavities - is one of the main technological processes used for conditioning low and intermediate level radioactive waste. At FSUE 'Radon' the industrialized method of impregnation has been developed and since 2003 has been using for cementation of solid radioactive waste. The technology is that the polymer-cement solution, having high penetrating properties, is supplied under pressure through a tube to the bottom of the container in which solid radioactive waste has preliminarily been placed. The polymer-cement solution is evenly moving upwards through the channels between the particles of solid radioactive waste, fills the voids in the bulk volume of the waste and hardens, forming a cement compound, the amount of which is equal to the original volume. The aim of the investigation was a selection of a cement solution suitable for SRW impregnation (including fine particles) without solution depletion and bottom layers stuffing. It has been chosen a polymer: PHMG (polyhexamethylene-guanidine), which is a stabilizing and water-retaining component of the cement solution. The experiments confirm that the polymer increases the permeability of the cement solution by a 2-2.5 factor, the viscosity by a 1.2 factor, the stability of the consistency by a 1.5-1.7 factor, and extends the operating range of the W/C ratio to 0.5-1.1. So it is possible to penetrate a volume of SRW bigger by a 1.5-2.0 factor. It has been proved, that PHMG polymer increases strength and frost-resistance of the final compounds by a 1.8-2.7 factor, and contributes to fast strength development at the beginning of hardening and it decreases Cs-137 leashing rate by a 1.5-2 factor

  14. Laboratory plant for the separation of cesium from waste solutions of the PUREX process

    International Nuclear Information System (INIS)

    Richter, M.; Eckert, B.; Riemenschneider, J.; Mallon, C.; Mann, D.

    1983-01-01

    A laboratory plant for the separation of cesium from a fission product waste solution of the fuel reprocessing is described. The plant consists of two stages. In the first stage cesium is adsorbed on ammonium molybdatophosphate (AMP). Then the adsorbent is dissolved. From the solution cesium is adsorbed on a cationic ion exchanger in the second stage. Then AMP can be reproduced from this solution. For the elution of cesium in the second stage a NH 4 NO 3 solution (3 m) is used. Flow sheet, construction and the control device of the plant are described and the results of tests with a model solution are given. (author)

  15. A novel Canadian solution for processing and disposal of mixed liquid wastes

    International Nuclear Information System (INIS)

    Suryanarayan, S.; Husain, A.; Husain, S.; Grey, M.; Elwood, C.; White, T.; Wigle, K.

    2011-01-01

    In 2009, Bruce Power contracted with Kinectrics for the disposal of its accumulated mixed liquid waste (MLW) inventory. The waste consists of solvent, PCB (Poly Chlorinated Biphenyls) and non-PCB contaminated oils and aqueous waste drums. The radioactivity in the wastes is principally due to cobalt-60, cesium-137 and tritium. Historically, MLW drums originating from Canadian utilities were shipped to a licensed US facility for destruction via incineration. This option is relatively expensive considering the significant logistics and destruction costs involved. In addition, restrictions now apply on importation of PCB containing wastes in to the US. Because of this, Kinectrics developed a wholly Canadian solution for the disposal of the MLW. Disposal of Bruce Power's MLW was conceived to be carried out in three phases. Phase 1: Develop an overall plan for disposal of the accumulated wastes, Phase 2: Dispose the PCB oil waste drums (highest priority), and Phase 3: Dispose all other waste drums. Phases 1 & 2 have been completed and Phase 3 is currently underway with 17 drums having been disposed so far. A description of the key activities undertaken to date are described in this paper. This work sets the stage for the future management of MLW based exclusively or largely on disposal within Canada. All key technical, regulatory and logistical issues pertaining to the receipt, handling, processing and shipment of the wastes were addressed. Equipment was installed for basic processing of the incoming wastes. Based on Pathways methodology, it was shown that the wastes can be shipped to unlicensed facilities within Canada without exceeding the 10 μSv per annum exposure to the critical individual. Despite this and for compliance with ALARA, wastes exceeding self-imposed threshold levels of radioactivity will be solidified and shipped for storage as radioactive waste. (author)

  16. A novel Canadian solution for processing and disposal of mixed liquid wastes

    Energy Technology Data Exchange (ETDEWEB)

    Suryanarayan, S.; Husain, A. [Kinectrics Inc., Toronto, ON (Canada); Husain, S.; Grey, M. [Candesco, Toronto, ON (Canada); Elwood, C.; White, T.; Wigle, K. [Bruce Power, Tiverton, ON (Canada)

    2011-07-01

    In 2009, Bruce Power contracted with Kinectrics for the disposal of its accumulated mixed liquid waste (MLW) inventory. The waste consists of solvent, PCB (Poly Chlorinated Biphenyls) and non-PCB contaminated oils and aqueous waste drums. The radioactivity in the wastes is principally due to cobalt-60, cesium-137 and tritium. Historically, MLW drums originating from Canadian utilities were shipped to a licensed US facility for destruction via incineration. This option is relatively expensive considering the significant logistics and destruction costs involved. In addition, restrictions now apply on importation of PCB containing wastes in to the US. Because of this, Kinectrics developed a wholly Canadian solution for the disposal of the MLW. Disposal of Bruce Power's MLW was conceived to be carried out in three phases. Phase 1: Develop an overall plan for disposal of the accumulated wastes, Phase 2: Dispose the PCB oil waste drums (highest priority), and Phase 3: Dispose all other waste drums. Phases 1 & 2 have been completed and Phase 3 is currently underway with 17 drums having been disposed so far. A description of the key activities undertaken to date are described in this paper. This work sets the stage for the future management of MLW based exclusively or largely on disposal within Canada. All key technical, regulatory and logistical issues pertaining to the receipt, handling, processing and shipment of the wastes were addressed. Equipment was installed for basic processing of the incoming wastes. Based on Pathways methodology, it was shown that the wastes can be shipped to unlicensed facilities within Canada without exceeding the 10 μSv per annum exposure to the critical individual. Despite this and for compliance with ALARA, wastes exceeding self-imposed threshold levels of radioactivity will be solidified and shipped for storage as radioactive waste. (author)

  17. Expected behavior of HLW glass in storage

    International Nuclear Information System (INIS)

    McElroy, J.L.

    1975-01-01

    Glass produced by solidification of high-level radioactive liquid waste is studied. Conditions to which the waste form will be exposed in a typical handling sequence representative of current U. S. planning are tabulated. The reference matrix for waste form characterization is discussed, and some of the properties of high-level waste glass are described: physical properties, leachability, fracturing, vaporization, and containment in canister. 12 fig, 5 tables

  18. 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

  19. Engineering solutions of environmental problems in organic waste handling

    Science.gov (United States)

    Briukhanov, A. Y.; Vasilev, E. V.; Shalavina, E. V.; Kucheruk, O. N.

    2017-10-01

    This study shows the urgent need to consider modernization of agricultural production in terms of sustainable development, which takes into account environmental implications of intensive technologies in livestock farming. Some science-based approaches are offered to address related environmental challenges. High-end technologies of organic livestock waste processing were substantiated by the feasibility study and nutrient balance calculation. The technologies were assessed on the basis of best available techniques criteria, including measures such as specific capital and operational costs associated with nutrient conservation and their delivery to the plants.

  20. Vitrification: a solution for the wastes of wastes; La vitrification: ca chauffe pour les ultimes

    Energy Technology Data Exchange (ETDEWEB)

    Guihard, B. [Europlasma, 33 - Saint Medard en Jalles (France)

    1997-07-01

    The incineration of wastes generates other wastes (fly ashes) that concentrate a large amount of polluting substances (heavy metals, salts..). French law requires a stabilization of this kind of wastes before their storage. Today vitrification can be considered as an alternative to the stabilization and storage way, the vitrified products could be seen as an interesting material in the building industry or in road works. A few years ago the municipality of Bordeaux decided to launch a demonstration program and a REFIOM (fly ashes) vitrification unit has been operating since 1997. (A.C.)

  1. 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

  2. Development of a freeze-drying process of waste-solution, 2

    International Nuclear Information System (INIS)

    Kondo, Isao; Kawasaki, Takeshi

    1988-01-01

    The waste solution treatment process in Plutonium Conversion Development Facility (PCDF) consists of Evaporation-Condensation and Neutrazation-Agglometation-Precipitation process, which produces the distillate as recovered acid at first step and separates Pu-U element from condenced solution at second step. This process needs many stages to get high decontamination efficiency and then the Evaporator is in very corrosive state because the nitric acid solution is heated over 100 degrees C to be evaporated. So, in PCDF, it was started the development of Freeze-Drying process to waste solution treatment. This process is suitable for a little quantity of the solution including nitric acid as produced in the Microwave Heating method. Moreover the process has high decontamination efficiency and has good performance of equipment. The result of the cold test of Freeze-Drying process with nitric acid is discribed in this paper. (author)

  3. Liquid waste treatment system. Final report

    International Nuclear Information System (INIS)

    Baker, M.N.; Houston, H.M.

    1999-01-01

    Pretreatment of high-level liquid radioactive waste (HLW) at the West Valley Demonstration Project (WVDP) involved three distinct processing operations: decontamination of liquid HLW in the Supernatant Treatment System (STS); volume reduction of decontaminated liquid in the Liquid Waste Treatment System (LWTS); and encapsulation of resulting concentrates into an approved cement waste form in the Cement Solidification System (CSS). Together, these systems and operations made up the Integrated Radwaste Treatment System (IRTS)

  4. Determination of microamounts of uranium in waste solutions

    International Nuclear Information System (INIS)

    Birringer, K.J.; Netzer, S.; Kuhn, E.; Groll, P.

    1975-07-01

    A method for the determination of microamounts of uranium in presence of high amounts of fission and corrosion products is described. Uranium is separated by reversed-phase chromatography on a small column, packed with Voltalef micro and impregnated with TOPO. For the direct photometric determination uranium is eluted by TAM dissolved in ethanol/pyridine. The efficiency of the separation, using a suitable scrub-solution, was tested with solutions of simulated inactive fission and corrosion products. The reproducibility of the method, with 24 μg of uranium, is +- 2,5%. (orig.) [de

  5. Idaho National Engineering Laboratory High-Level Waste Roadmap

    International Nuclear Information System (INIS)

    1993-08-01

    The Idaho National Engineering Laboratory (INEL) High-Level Waste (HLW) Roadmap takes a strategic look at the entire HLW life-cycle starting with generation, through interim storage, treatment and processing, transportation, and on to final disposal. The roadmap is an issue-based planning approach that compares ''where we are now'' to ''where we want and need to be.'' The INEL has been effectively managing HLW for the last 30 years. Calcining operations are continuing to turn liquid HLW into a more manageable form. Although this document recognizes problems concerning HLW at the INEL, there is no imminent risk to the public or environment. By analyzing the INEL current business operations, pertinent laws and regulations, and committed milestones, the INEL HLW Roadmap has identified eight key issues existing at the INEL that must be resolved in order to reach long-term objectives. These issues are as follows: A. The US Department of Energy (DOE) needs a consistent policy for HLW generation, handling, treatment, storage, and disposal. B. The capability for final disposal of HLW does not exist. C. Adequate processes have not been developed or implemented for immobilization and disposal of INEL HLW. D. HLW storage at the INEL is not adequate in terms of capacity and regulatory requirements. E. Waste streams are generated with limited consideration for waste minimization. F. HLW is not adequately characterized for disposal nor, in some cases, for storage. G. Research and development of all process options for INEL HLW treatment and disposal are not being adequately pursued due to resource limitations. H. HLW transportation methods are not selected or implemented. A root-cause analysis uncovered the underlying causes of each of these issues

  6. Depleted Hydrocarbon Reservoirs Present a Safe and Practical Burial Solution for Graphite Waste

    International Nuclear Information System (INIS)

    Rahmani, L.

    2016-01-01

    A solution for graphite waste is proposed that combines reliance on thick impermeable host rock that is needed to confine the long-life radioactivity content of most irradiated graphite with low capitalistic and operational unit volume costs that are required to render this bulky waste form manageable. The solution, uniquely applicable to irradiated graphite due to its low dose rates, moderate mechanical strength and light density, consists in three steps: first, graphite is fine-crushed under water; second, it is made in an aqueous suspension; third, the suspension is injected into a deep, disused hydrocarbon reservoir. Each of these steps only involves well mastered techniques. Regulatory changes that may allow this solution to be added to the gamut of available waste routes, geochemical issues, availability of depleted reservoirs and cost projections are presented. (author)

  7. 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

  8. Processing of radioactive waste solutions in a vacuum evaporator-crystallizer

    International Nuclear Information System (INIS)

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

    1975-01-01

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

  9. Precipitation-filtering technology for uranium waste solution generated on washing-electrokinetic decontamination

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Gye-Nam, E-mail: kimsum@kaeri.re.kr; Park, Uk-Ryang; Kim, Seung-Soo; Moon, Jei-Kwon

    2015-05-15

    Graphical abstract: A recycling process diagram for the volume reduction of waste solution generated from washing-electrokinetic decontamination. - Highlights: • A process for recycling a waste solution generated was developed. • The total metal precipitation rate by NaOH in a supernatant after precipitation was the highest at pH 9. • The uranium radioactivity in the treated solution upon injection of 0.2 g of alum was lower. • After drying, the volume of sludge was reduced to 35% of the initial sludge volume. - Abstract: Large volumes of uranium waste solution are generated during the operation of washing-electrokinetic decontamination equipment used to remove uranium from radioactive soil. A treatment technology for uranium waste solution generated upon washing-electrokinetic decontamination for soil contaminated with uranium has been developed. The results of laboratory-size precipitation experiments were as follows. The total amount of metal precipitation by NaOH for waste solution was highest at pH 11. Ca(II), K(I), and Al(III) ions in the supernatant partially remained after precipitation, whereas the concentration of uranium in the supernatant was below 0.2 ppm. Also, when NaOH was used as a precipitant, the majority of the K(I) ions in the treated solution remained. The problem of CaO is to need a long dissolution time in the precipitation tank, while Ca(OH){sub 2} can save a dissolution time. However, the volume of the waste solution generated when using Ca(OH){sub 2} increased by 8 mL/100 mL (waste solution) compared to that generated when using CaO. NaOH precipitant required lower an injection volume lower than that required for Ca(OH){sub 2} or CaO. When CaO was used as a precipitant, the uranium radioactivity in the treated solution at pH 11 reached its lowest value, compared to values of uranium radioactivity at pH 9 and pH 5. Also, the uranium radioactivity in the treated solution upon injection of 0.2 g of alum with CaO or Ca(OH){sub 2} was

  10. Economic analysis of waste management alternatives for reprocessing wastes

    International Nuclear Information System (INIS)

    McKee, R.W.; Clark, L.L.; Daling, P.M.; Nesbitt, J.F.; Swanson, J.L.

    1984-02-01

    This study describes the results of a cost analysis of a broad range of alternatives for management of reprocessing wastes that would require geologic repository disposal. The intent was to identify cost-effective alternatives and the costs of potential repository performance requirements. Four integrated treatment facility alternatives for transuranic (TRU) wastes are described and compared. These include no treatment, compaction, incineration, and hulls melting. The advantages of reducing high-level wastes (HLW) volume are also evaluated as are waste transportation alternatives and several performance-related alternatives for emplacing waste in a basalt repository. Results show (1) that system costs for disposal of reprocessing waste are likely to be higher than those for disposal of spent fuel; (2) that volume reduction is cost-effective for both remote-handled (RH) TRU wastes and HLW, and that rail transport for HLW is more cost-effective than truck transport; (3) that coemplacement of RH-TRU wastes with HLW does not have a large cost advantage in a basalt repository; and (4) that, relative to performance requirements, the cost impact for elimination of combustibles is about 5%, long-lived containers for RH-TRU wastes can increase repository costs 10% to 20%, and immediate backfill compared to delayed backfill (bentonite/basalt) around the HLW canisters would increase repository costs up to 10% or overall system costs up to about 5%. 13 references, 4 figures, 12 tables

  11. Processing results of 1,800 gallons of mercury and radioactively contaminated mixed waste rinse solution

    International Nuclear Information System (INIS)

    Thiesen, B.P.

    1993-01-01

    The mercury-contaminated rinse solution (INEL waste ID number-sign 123; File 8 waste) was successfully treated at the Idaho National Engineering Laboratory (INEL). This waste was generated during the decontamination of the Heat Transfer Reactor Experiment 3 (HTRE-3) reactor shield tank. Approximately 1,800 gal of waste was generated and was placed into 33 drums. Each drum contained precipitated sludge material ranging from 1--10 in. in depth, with the average depth of about 2.5 in. The pH of each drum varied from 3--11. The bulk liquid waste had a mercury level of 7.0 mg/l, which exceeded the Resource Conservation and Recovery Act (RCRA) limit of 0.2 mg/l. The average liquid bulk radioactivity was about 2.1 pCi/ml, while the average sludge contamination was about 13,800 pci/g. Treatment of the waste required separation of the liquid from the sludge, filtration, pH adjustment, and ion exchange. Because of difficulties in processing, three trials were required to reduce the mercury levels to below the RCRA limit. In the first trial, insufficient filtration of the waste allowed solid particulate produced during pH adjustment to enter into the ion exchange columns and ultimately the waste storage tank. In the second trial, the waste was filtered down to 0.1 μ to remove all solid mercury compounds. However, before filtration could take place, a solid mercury complex dissolved and mercury levels exceeded the RCRA limit after filtration. In the third trial, the waste was filtered through 0.3-A filters and then passed through the S-920 resin to remove the dissolved mercury. The resulting solut

  12. 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

  13. 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

  14. Cognition of high-level radioactive waste disposal in the Tokyo metropolitan area

    International Nuclear Information System (INIS)

    Kimura, Hiroshi

    2010-01-01

    In Japan, the disposal of high-level radioactive waste (HLW) produced by nuclear power generation is an urgent issue. Recently, some questionnaire surveys were conducted. Especially the surveys in the Tokyo metropolitan area which were conducted by AESJ include the fulfilling questions concerning HLW relatively. In this paper, the author shows the results of surveys by AESJ. These results show that the issue concerning HLW is not so much concern for the respondents by comparison with many kinds of issues in the society. They also show that female respondents have less understanding about HLW disposal and have more degree of anxiety against HLW and disposal than male respondents. (author)

  15. Safety assessment driving radioactive waste management solutions (SADRWMS Methodology) implemented in a software tool (SAFRAN)

    Energy Technology Data Exchange (ETDEWEB)

    Kinker, M., E-mail: M.Kinker@iaea.org [International Atomic Energy Agency (IAEA), Vienna (Austria); Avila, R.; Hofman, D., E-mail: rodolfo@facilia.se [FACILIA AB, Stockholm (Sweden); Jova Sed, L., E-mail: jovaluis@gmail.com [Centro Nacional de Seguridad Nuclear (CNSN), La Habana (Cuba); Ledroit, F., E-mail: frederic.ledroit@irsn.fr [IRSN PSN-EXP/SSRD/BTE, (France)

    2013-07-01

    In 2004, the International Atomic Energy Agency (IAEA) organized the International Project on Safety Assessment Driving Radioactive Waste Management Solutions (SADRWMS) to examine international approaches to safety assessment for predisposal management of radioactive waste. The initial outcome of the SADRWMS Project was achieved through the development of flowcharts which could be used to improve the mechanisms for applying safety assessment methodologies to predisposal management of radioactive waste. These flowcharts have since been incorporated into DS284 (General Safety Guide on the Safety Case and Safety Assessment for Predisposal Management of Radioactive Waste), and were also considered during the early development stages of the Safety Assessment Framework (SAFRAN) Tool. In 2009 the IAEA presented DS284 to the IAEA Waste Safety Standards Committee, during which it was proposed that the graded approach to safety case and safety assessment be illustrated through the development of Safety Reports for representative predisposal radioactive waste management facilities and activities. To oversee the development of these reports, it was agreed to establish the International Project on Complementary Safety Reports: Development and Application to Waste Management Facilities (CRAFT). The goal of the CRAFT project is to develop complementary reports by 2014, which the IAEA could then publish as IAEA Safety Reports. The present work describes how the DS284 methodology and SAFRAN Tool can be applied in the development and review of the safety case and safety assessment to a range of predisposal waste management facilities or activities within the Region. (author)

  16. Management of radioactive waste in France-policy, issues, and solutions

    International Nuclear Information System (INIS)

    Tamborini, J.

    1996-01-01

    The French nuclear industry has conducted a study to define a policy and an organization to deal with the waste generated from nuclear power plants, the fuel cycle industries, and medicine, research, and other industrial nuclear applications. This has resulted in the introduction of an organization which, by appropriate and responsible management, can guarantee to protect people and the environment while ensuring industrial effectiveness. The body in charge of waste management in France is the National Radioactive Waste Management Agency (ANDRA) created in 1979. The French policy is based on waste classification and the related solutions for the evacuation of these wastes. High-level and long-lived waste management is regulated by a law passed Dec 30, 1991. The law outlines the research program to be conducted. Three main research objectives are prescribed: 1. reduction of the waste volumes and toxicity (partitioning and transmutation); 2. assessment of the waste isolation properties of deep geologic formations by underground research laboratories; 3. development of solidification processes and storage techniques for long-term interim storage in near-surface facilities. This research will be implemented within a 15 yr period. At present, applications are submitted to the authorities for the construction of underground research laboratories. At the end of this period, reports will be submitted to parliament. It will have to choose among various options. The construction of a deep geologic repository, if this option is chosen, will need the passage of a new law

  17. Safety assessment driving radioactive waste management solutions (SADRWMS Methodology) implemented in a software tool (SAFRAN)

    International Nuclear Information System (INIS)

    Kinker, M.; Avila, R.; Hofman, D.; Jova Sed, L.; Ledroit, F.

    2013-01-01

    In 2004, the International Atomic Energy Agency (IAEA) organized the International Project on Safety Assessment Driving Radioactive Waste Management Solutions (SADRWMS) to examine international approaches to safety assessment for predisposal management of radioactive waste. The initial outcome of the SADRWMS Project was achieved through the development of flowcharts which could be used to improve the mechanisms for applying safety assessment methodologies to predisposal management of radioactive waste. These flowcharts have since been incorporated into DS284 (General Safety Guide on the Safety Case and Safety Assessment for Predisposal Management of Radioactive Waste), and were also considered during the early development stages of the Safety Assessment Framework (SAFRAN) Tool. In 2009 the IAEA presented DS284 to the IAEA Waste Safety Standards Committee, during which it was proposed that the graded approach to safety case and safety assessment be illustrated through the development of Safety Reports for representative predisposal radioactive waste management facilities and activities. To oversee the development of these reports, it was agreed to establish the International Project on Complementary Safety Reports: Development and Application to Waste Management Facilities (CRAFT). The goal of the CRAFT project is to develop complementary reports by 2014, which the IAEA could then publish as IAEA Safety Reports. The present work describes how the DS284 methodology and SAFRAN Tool can be applied in the development and review of the safety case and safety assessment to a range of predisposal waste management facilities or activities within the Region. (author)

  18. 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

  19. 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)

  20. Analysis for silicon in solution in high level waste glass durability studies

    International Nuclear Information System (INIS)

    Lewis, R.A.; Smart, R.St.C.; Dale, L.S.; Levins, D.M.

    1982-01-01

    In comparative studies of the durability of HLW glasses, the measurement of the dissolution of the silicate network, in terms of both rate and extent, is of prime importance. To achieve this, analytical techniques such as colorimetry, flame atomic absorption spectrometry and inductively-coupled plasma emission spectrometry are used. The reliability of these analytical techniques for determination of silicon concentration in dissolution of HLW glasses, is examined. At high concentrations both FAA and ICP are accurate but colorimetry, even with HF pretreatment or NaOH digestion, does not give agreement with ICP. At concentrations below 40 mg l -1 all three methods are reliable. (Auth.)

  1. Progress report for 1985/86 from the Waste Treatment and Disposal Working Party covering joint funded work

    International Nuclear Information System (INIS)

    Claxton, D.G.S.A.

    1986-01-01

    The Waste Treatment and Disposal Working Party (WTDWP) covered the areas of: ILW Product Evaluation, ILW and HLW Disposal Studies and ILW and HLW Quality Checking. The objectives of the programme were to evaluate potential waste products arising from the treatment of ILW, and to develop appropriate techniques which could be used to check the quality of the finished waste product. (author)

  2. Potential agents for removal of actinides from waste solutions

    International Nuclear Information System (INIS)

    Romanovski, V.V.; Whisenhunt, D.W.; Veeck, A.C.; Andersen, W.A.; Hoffman, D.C.; Jide, X.; White, D.; Raymond, K.N.

    1996-01-01

    The uptake of Th(IV) from nitric acid and hydrochloric acid solutions by chelating ion exchange resins containing catechol, 1,2- hydroxypyridinone (1,2-HOPO) and 3,4-hydroxypyridinone (3,4-HOPO) functional groups, has been investigated. These polystyrene based materials show excellent kinetics for uptake of Th(IV) and have a high loading capacity. Liquid/liquid extractants have also been synthesized by addition of lipophilic side chains to the chelating groups (1,2-HOPO; 3,4-HOPO; 3,2-HOPO; catecholamide; terephthalamide). The initial evaluation of the extraction properties has been carried out

  3. Application of Epoxy Based Coating Instacote on Waste Tank Tops

    International Nuclear Information System (INIS)

    Pike, J.A.

    1998-01-01

    This evaluation examines the compatibility of coating Instacote with existing High-Level Waste facilities and safety practices. No significant incompatibilities are identified. The following actions need to be completed as indicated when applying Instacote on waste tank tops:(1) Prior to application in ITP facilities, the final product should be tested for chemical resistance to sodium tetraphenylborate solutions or sodium titanate slurries.(2) Any waste contaminated with Part A or B that can not be removed by the vendor such as for radiological contamination, HLW must hold the waste until HLW completes a formal assessment of the waste, disposal criteria, and impact.(3) Prior to the start of any application of the coating, each riser needs to be evaluated for masking and masking applied if needed.(4) At the conclusion of an application actual total weight of material applied to a waste tank needs to documented and sent to the tank top loading files for reference purposes.(5) Verify that the final product contains less than 250 ppm chloride

  4. Porous materials based on foaming solutions obtained from industrial waste

    Science.gov (United States)

    Starostina, I. V.; Antipova, A. N.; Ovcharova, I. V.; Starostina, Yu L.

    2018-03-01

    This study analyzes foam concrete production efficiency. Research has shown the possibility of using a newly-designed protein-based foaming agent to produce porous materials using gypsum and cement binders. The protein foaming agent is obtained by alkaline hydrolysis of a raw mixture consisting of industrial waste in an electromagnetic field. The mixture consists of spent biomass of the Aspergillus niger fungus and dust from burning furnaces used in cement production. Varying the content of the foaming agent allows obtaining gypsum binder-based foam concretes with the density of 200-500 kg/m3 and compressive strength of 0.1-1.0 MPa, which can be used for thermal and sound insulation of building interiors. Cement binders were used to obtain structural and thermal insulation materials with the density of 300-950 kg/m3 and compressive strength of 0.9-9.0 MPa. The maximum operating temperature of cement-based foam concretes is 500°C because it provides the shrinkage of less than 2%.

  5. Volumetric determination of hydroxide, aluminate, and carbonate in alkaline solutions of nuclear waste

    International Nuclear Information System (INIS)

    Baumann, E.W.

    1975-06-01

    An integrated procedure was developed for determining OH - , Al(OH) 4 - , and CO 3 2- in alkaline nuclear waste. The free alkali, the hydroxide released when Al(OH) 3 is complexed with oxalate, and the precipitated BaCO 3 were determined by acidimetric titration. With a 50-μl sample, the relative standard deviations were 1 to 2 percent for nonradioactive test solutions and 2 to 5 percent for radioactive process solutions. (U.S.)

  6. Intercomparison of Cement Solid-Solution Models. Issues Affecting the Geochemical Evolution of Repositories for Radioactive Waste

    International Nuclear Information System (INIS)

    Benbow, Steven; Savage, David; Walker, Colin

    2007-05-01

    Many concepts for the geological storage of radioactive waste incorporate cement based materials, which act to provide a chemical barrier, impede groundwater flow or provide structural integrity of the underground structures. Thus, it is important to understand the long-term behaviour of these materials when modelling scenarios for the potential release and migration of radionuclides. In the presence of invasive groundwater, the chemical and physical properties of cement, such as its pH buffering capacity, resistance to flow, and its mechanical properties, are expected to evolve with time. Modelling the degradation of cement is complicated by the fact that the long term pH buffer is controlled by the incongruent dissolution behaviour of calcium-silicate-hydrate (C-S-H) gel. It has been previously shown (SKI Report 2005:64) that it is possible to simulate the long term evolution of both the physical and chemical properties of cement based materials in an invasive groundwater using a fully coupled geochemical transport model. The description of the incongruent dissolution of C-S-H gel was based on a binary solid solution aqueous solution (SSAS) between end-member components portlandite (Ca(OH) 2 ) and a C-S-H gel composition expressed by its component oxides (CaH 2 SiO 4 ). The models considered a range of uncertainties including different groundwater compositions, parameterised couplings between the evolution of porosity with permeability and diffusivity and alternative secondary mineral assemblages. The results of the modelling suggested that alternative evolutions were possible under these different conditions. The focus of this report is to address the uncertainty regarding the choice of model for the C-S-H gel dissolution. We compare two alternative C-S-H SSAS models with the one that was used in the previous report, with an emphasis on a direct comparison of the model predictions. Thus we have chosen one simple simulated experimental model based on those in the

  7. Numerical Model of Fluid Flow through Heterogeneous Rock for High Level Radioactive Waste Disposal

    International Nuclear Information System (INIS)

    Shirai, M.; Chiba, R.; Takahashi, T.; Hashida, T.; Fomin, S.; Chugunov, V.; Niibori, Y.

    2007-01-01

    An international consensus has emerged that deep geological disposal on land is one of the most appropriate means for high level radioactive wastes (HLW). The fluid transport is slow and radioactive elements are dangerous, so it's impossible to experiment over thousands of years. Instead, numerical model in such natural barrier as fractured underground needs to be considered. Field observations reveal that the equation with fractional derivative is more appropriate for describing physical phenomena than the equation which is based on the Fick's law. Thus, non-Fickian diffusion into inhomogeneous underground appears to be important in the assessment of HLW disposal. A solute transport equation with fractional derivative has been suggested and discussed in literature. However, no attempts were made to apply this equation for modeling of HLW disposal with account for the radioactive decay. In this study, we suggest the use of a novel fractional advection-diffusion equation which accounts for the effect of radioactive disintegration and for interactions between major, macro pores and fractal micro pores. This model is fundamentally different from previous proposed model of HLW, particularly in utilizing fractional derivative. Breakthrough curves numerically obtained by the present model are presented for a variety of rock types with respect to some important nuclides. Results of the calculation showed that for longer distance our model tends to be more conservative than the conventional Fickian model, therefore our model can be said to be safer

  8. Denitration of simulated high-level liquid wastes and selective removal of cesium with zeolites

    Energy Technology Data Exchange (ETDEWEB)

    Mimura, Hitoshi; Kanno, Takuji [Tohoku Univ., Sendai (Japan). Research Inst. of Mineral Dressing and Metallurgy; Kimura, Toshiya

    1982-03-01

    Denitration of high-level liquid wastes (HLW) from nuclear fuel reprocessing has been studied. Selective removal of Cs has been also examined with various types of zeolites. The following zeolites were used in this study; Na-synthetic mordenite (NaSM), Na-natural mordenite (NaNM), Na-natural clinoptilolite (NaCP) and H-synthetic mordenites (HSM). The effective denitration is found in the simulated HLW (15 components, 2N HNO/sub 3/ soln.) containing platinum group elements in the case of the addition of formic acid, and the pH of the solution shows the value of 5.4 when the excess formic acid ((HCOOH)/(HNO/sub 3/) = 2.0) was added. Platinum group elements may react as a catalyst for the decomposition of nitric acid and the excess formic acid. The break-through properties of NaSM column are poor for the simulated HLW, and the selective removal of Cs appears to be difficult. On the other hand, good results are obtained in the denitrated HLW, i.e., break-through capacity, total capacity and column utilization are 59.4 (meq./100 g zeolite), 147 (meq./100 g zeolite) and 40.4 (%), respectively. The break-through properties of NaSM and NaNM are superior to those of HSM. The break-through capacity and column utilization increase with an increase in column temperature.

  9. Denitration of simulated high-level liquid wastes and selective removal of cesium with zeolites

    International Nuclear Information System (INIS)

    Mimura, Hitoshi; Kanno, Takuji; Kimura, Toshiya.

    1982-01-01

    Denitration of high-level liquid wastes (HLW) from nuclear fuel reprocessing has been studied. Selective removal of Cs has been also examined with various types of zeolites. The following zeolites were used in this study; Na-synthetic mordenite (NaSM), Na-natural mordenite (NaNM), Na-natural clinoptilolite (NaCP) and H-synthetic mordenites (HSM). The effective denitration is found in the simulated HLW (15 components, 2N HNO 3 soln.) containing platinum group elements in the case of the addition of formic acid, and the pH of the solution shows the value of 5.4 when the excess formic acid ([HCOOH]/[HNO 3 ] = 2.0) was added. Platinum group elements may react as a catalyst for the decomposition of nitric acid and the excess formic acid. The break-through properties of NaSM column are poor for the simulated HLW, and the selective removal of Cs appears to be difficult. On the other hand, good results are obtained in the denitrated HLW, i.e., break-through capacity, total capacity and column utilization are 59.4 (meq./100 g zeolite), 147 (meq./100 g zeolite) and 40.4 (%), respectively. The break-through properties of NaSM and NaNM are superior to those of HSM. The break-through capacity and column utilization increase with an increase in column temperature. (author)

  10. 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.

  11. The Waste Isolation Pilot Plant: a potential solution for the disposal of transuranic waste

    National Research Council Canada - National Science Library

    National Research Council Staff; Commission on Geosciences, Environment and Resources; Division on Earth and Life Studies; National Research Council; National Academy of Sciences

    ... Isolation Pilot Plant Board on Radioactive Waste Management Commission on Geosciences, Environment, and Resources National Research Council NATIONAL ACADEMY PRESS Washington, D.C. 1996 i Copyrighttrue Please breaks inserted. are Page files. accidentally typesetting been have may original from the errors not typographic original retained, and from the c...

  12. MODELING AN ION EXCHANGE PROCESS FOR CESIUM REMOVAL FROM ALKALINE RADIOACTIVE WASTE SOLUTIONS

    International Nuclear Information System (INIS)

    Smith, F.; Hamm, Luther; Aleman, Sebastian; Michael, Johnston

    2008-01-01

    The performance of spherical Resorcinol-Formaldehyde ion-exchange resin for the removal of cesium from alkaline radioactive waste solutions has been investigated through computer modeling. Cesium adsorption isotherms were obtained by fitting experimental data using a thermodynamic framework. Results show that ion-exchange is an efficient method for cesium removal from highly alkaline radioactive waste solutions. On average, two 1300 liter columns operating in series are able to treat 690,000 liters of waste with an initial cesium concentration of 0.09 mM in 11 days achieving a decontamination factor of over 50,000. The study also tested the sensitivity of ion-exchange column performance to variations in flow rate, temperature and column dimensions. Modeling results can be used to optimize design of the ion exchange system

  13. Selective Recovery Of Copper From Solutions After Bioleaching Electronic Waste

    Directory of Open Access Journals (Sweden)

    Willner Joanna

    2015-06-01

    Full Text Available Research on selective extraction of copper from solution after bioleaching grounded printed circuit boards (PCBs using LIX 860N-IC were conducted. The effect of LIX 860N-IC concentration, phase ratio and influence of initial pH value of aqueous phase on the extraction of copper and iron was examined. It was found that the extraction rate of copper increases with the LIX 860N-IC concentration. Best results of Cu extraction (98 % were achieved with extractant concentration of 5 % and pH 1.9. Higher pH value of aqueous phase (pH=2.4 is conducive to the simultaneous effect of Fe co-extraction.

  14. Malonamides as new extractants for nuclear waste solutions

    International Nuclear Information System (INIS)

    Cuillerdier, C.; Musikas, C.; Hoel, P.; Nigond, L.; Vitart, X.

    1989-01-01

    Substituted malonamides are able to extract α emitters from radioactive solutions in nitric acid, all the actinides (III, IV, VI) are well extracted and can be easily back extracted. Some problems remain with neptunium and technetium. These solvents are not expensive. For an industrial purpose, synthesis has been optimized, and a proper choice of commercial basic products can decrease the cost. The solvent obtained on a pilot scale (1 kg) was found to be pure enough, it didn't need any additional treatment. Degradation under hydrolysis or radiolysis is not important in the conditions of practical experiment (t 0 C). Degradation products can be washed with NaOH (carboxylic acids) they don't give precipitates or emulsions. Efficiency of the solvent is good compared to CMPO, taking into account the lack of extensive industrial development. Further researches are undertaken in two main directions: optimizing the synthesis and use of aliphatic diluents

  15. Treatment of nuclear waste solutions using a new class of extractants: pentaalkyl propane diamides

    International Nuclear Information System (INIS)

    Cuillerdier, C.; Musikas, C.; Hoel, P.

    1990-01-01

    A new class of bifunctional extractants pentaalkyl propane diamides is studied in order to extract trivalent cations (Am 3+ , Cm 3+ ...) and other actinides contained in waste solutions of nuclear industry. These solvents are completely incinerable and don't produce harmfull degradation products. Their main chemicals properties are reviewed. The results of a mixer-settler battery experiment are given

  16. Secondary Waste Form Down-Selection Data Package—Fluidized Bed Steam Reforming Waste Form

    Energy Technology Data Exchange (ETDEWEB)

    Qafoku, Nikolla; Westsik, Joseph H.; Strachan, Denis M.; Valenta, Michelle M.; Pires, Richard P.

    2011-09-12

    The Hanford Site in southeast Washington State has 56 million gallons of radioactive and chemically hazardous wastes stored in 177 underground tanks (ORP 2010). The U.S. Department of Energy (DOE), Office of River Protection (ORP), through its contractors, is constructing the Hanford Tank Waste Treatment and Immobilization Plant (WTP) to convert the radioactive and hazardous wastes into stable glass waste forms for disposal. Within the WTP, the pretreatment facility will receive the retrieved waste from the tank farms and separate it into two treated process streams. These waste streams will be vitrified, and the resulting waste canisters will be sent to offsite (high-level waste [HLW]) and onsite (immobilized low-activity waste [ILAW]) repositories. As part of the pretreatment and ILAW processing, liquid secondary wastes will be generated that will be transferred to the Effluent Treatment Facility (ETF) on the Hanford Site for further treatment. These liquid secondary wastes will be converted to stable solid waste forms that will be disposed of in the Integrated Disposal Facility (IDF). To support the selection of a waste form for the liquid secondary wastes from WTP, Washington River Protection Solutions (WRPS) has initiated secondary waste form testing work at Pacific Northwest National Laboratory (PNNL). In anticipation of a down-selection process for a waste form for the Solidification Treatment Unit to be added to the ETF, PNNL is developing data packages to support that down-selection. The objective of the data packages is to identify, evaluate, and summarize the existing information on the four waste forms being considered for stabilizing and solidifying the liquid secondary wastes. At the Hanford Site, the FBSR process is being evaluated as a supplemental technology for treating and immobilizing Hanford LAW radioactive tank waste and for treating secondary wastes from the WTP pretreatment and LAW vitrification processes.

  17. 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.

  18. 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

  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

    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.

  20. 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.

  1. Bioleaching of fly ash from municipal solid waste incineration using kitchen waste saccharified solution as culture medium

    International Nuclear Information System (INIS)

    Wei, S.; Juan, W.; Qunhui, W.

    2013-01-01

    Summary: Reduced sugar in saccharified solution from kitchen waste was used as the carbon source. Domesticated A. niger AS 3.879C , which can withstand 20% of kitchen waste, was used as the inoculum in the bioleaching process of municipal solid waste incineration fly ash. The effect of reduced sugar concentration, fly ash concentration, and medium volume on the heavy metal extraction and yield of fly ash as well as the optimum bioleaching conditions; the inoculation amount of AS 3 .879C 1% (v/v), reduced sugar concentration of 80 g/l, fly ash concentration of 20 g/l, medium volume of 200 ml, and the addition of fly ash (20 g/l) after culturing for 4 days at 30 degree C and 140 r/min were obtained. Under the optimum condition, the extraction yield of the seven tested heavy metals are in the order of Cd > Zn > Cu > Mn > Pb > Cr > Fe; the extraction yield of Cd and Zn reached 88.7% and 73.1% respectively. Fly ash satisfied the Standard for Pollution Control on the Security Landfill Site for Hazardous Wastes (GB 18598-2001) after heavy metal extraction. (author)

  2. 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

  3. 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.

  4. 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).

  5. 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.

  6. 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)

  7. Vitrification of HLW in cold crucible melter

    International Nuclear Information System (INIS)

    Bordier, G.

    2005-01-01

    The performance of the vitrification process currently used in the La Hague commercial reprocessing plants has been continuously improved during more than ten years of operation. In parallel the CEA (French Atomic Energy Commission), COGEMA (Industrial Operator), and SGN (COGEMA's Engineering) have developed the cold crucible melter vitrification technology to obtain greater operating flexibility, increased plant availability and further reduction of secondary waste generated during operations. The cold crucible is a compact water-cooled melter in which the radioactive waste and the glass additives are melted by direct high frequency induction. The cooling of the melter produces a solidified glass layer that protects the melter's inner wall from corrosion. Because the heat is transferred directly to the melt, high operating temperatures can be achieved with no impact on the melter itself. COGEMA plans to implement the cold crucible technology to vitrify high level liquid waste from reprocessed spent U-Mo-Sn-Al fuel (used in gas cooled reactor). The cold crucible was selected for the vitrification of this particularly hard-to-process waste stream because it could not be reasonably processed in the standard hot induction melters currently used at the La Hague vitrification facilities: the waste has a high molybdenum content which makes it very corrosive and also requires a special high temperature glass formulation to obtain sufficiently high waste loading factors (12 % in molybdenum). A special glass formulation has been developed by the CEA and has been qualified through lab and pilot testing to meet standard waste acceptance criteria for final disposal of the U-Mo waste. The process and the associated technologies have been also qualified on a full-scale prototype at the CEA pilot facility in Marcoule. Engineering study has been integrated in parallel in order to take into account that the Cold Crucible should be installed remotely in one of the R7 vitrification

  8. Recovery of cyanide in gold leach waste solution by volatilization and absorption.

    Science.gov (United States)

    Gönen, N; Kabasakal, O S; Ozdil, G

    2004-09-10

    In this study, the effects of pH, time and temperature in regeneration of cyanide in the leaching waste solution of gold production from disseminated gold ore by cyanidation process were investigated and the optimum conditions, consumptions and cyanide recovery values were determined. The sample of waste solution containing 156 mg/l free CN- and 358 mg/l total CN-, that was obtained from Gümüşhane-Mastra/Turkey disseminated gold ores by cyanidation and carbon-in-pulp (CIP) process under laboratory conditions was used in the experiments. Acidification with H2SO4, volatilization of hydrogen cyanide (HCN) with air stripping and absorption of HCN in a basic solution stages were applied and under optimum conditions, 100% of free cyanide and 48% of complex cyanide and consequently 70% of the total cyanide in the liquid phase of gold leach effluent are recovered.

  9. 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

  10. 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

  11. R and D Activities on high-level nuclear waste management

    International Nuclear Information System (INIS)

    Watanabe, Shosuke

    1985-01-01

    High-level liquid waste (HLLW) at Tokai Reprocessing Plant has been generated from reprocessing of spent fuels from the light water reactors, and successfully managed since 1977. At the time of 1984, about 154m 3 of HLLW from 170 tons of spent fuels were stored in three high-integrity stainless steel tanks (90m 3 for each) as a nitric acid aqueous solution. The HLLW arises mainly from the first cycle solvent extraction phase. Alkaline solution to scrub the extraction solvent is another source of HLLW. The Advisory Committee on Radioactive Waste Management reported the concept on disposal of high-level waste (HLW) in Japan in 1980 report, that the waste be solidified into borosilicate glass and then be disposed in deep geologic formation so as to minimize the influence of the waste on human environment, with the aid of multibarrier system which is the combination of natural barrier and engineered barrier

  12. Radioactive waste disposal: an international law perspective

    International Nuclear Information System (INIS)

    Barrie, G.N.

    1989-01-01

    The question of radioactive waste disposal is the most intractable technical and political problem facing nuclear industry. Environmentalists world-wide demand a nuclear waste policy that must be ecologically acceptable internationally. Radioactive wastes and oil pollution were the first two types of marine pollution to receive international attention and various marine pollution controls were established. Ocean disposal was co-ordinated by the Nuclear Energy Agency and the Organization of Economic Co-operation and Development in 1967. The first treaty was the 1958 Convention on the High Seas (High Seas Convention). In response to its call for national co-operation the International Atomic Energy Agency (IAEA) established its Brynielson panel. The IAEA first issued guidelines on sea dumping in 1961. The London Dumping Convention, written in 1972, is the only global agreement concerned solely with the disposal of wastes in the marine environment by dumping. None of the global agreements make specific reference to sea-bed disposal of high-level radioactive wastes. Negotiations began at the Third UN Conference on the Law of the Sea (UNCLOS III) for the codification of a comprehensive treaty concerned with the protection, conservation, sustainable use and development of the marine environment. Burial in deep geological formations is a method of HLW disposal which decreases the chances of accidental intrusion by mankind and has little likelihood of malicious intrusion. National waste management programmes of different countries differ but there is agreement on the acceptable technical solutions to issues of waste management. The final disposition of HLW - storage or disposal - has not been decisively determined, but there is growing consensus that geological land-based disposal is the most viable alternative. Expanded international technical co-operation could well reduce the time needed to develop effective waste disposal mechanisms

  13. Performance of surrogate high-level waste glass in the presence of iron corrosion products

    International Nuclear Information System (INIS)

    Jain, V.; Pan, Y.M.

    2004-01-01

    Radionuclide release from a waste package (WP) is a series of processes that depend upon the composition and flux of groundwater contacting the waste-forms (WF); the corrosion rate of WP containers and internal components made of Alloy 22, 316L SS, 304L SS and carbon steel; the dissolution rate of high-level radioactive waste (HLW) glass and spent nuclear fuel (SNF); the solubility of radionuclides; and the retention of radionuclides in secondary mineral phases. In this study, forward reaction rate measurements were made on a surrogate HLW glass in the presence of FeCl 3 species. Results indicate that the forward reaction rate increases with an increase in the FeCl 3 concentration. The addition of FeCl 3 causes the drop in the pH due to hydrolysis of Fe 3+ ions in the solution. Results based on the radionuclide concentrations and dissolution rates for HLW glass and SNF indicate that the contribution from glass is similar to SNF at 75 deg C. (authors)

  14. External Criticality Risk of Immobilized Plutonium Waste Form in a Geologic Repository

    International Nuclear Information System (INIS)

    McClure, J.

    2001-01-01

    This purpose of this technical report is to provide a comprehensive summary of the waste package (WP) external criticality-related risk of the Plutonium Disposition ceramic waste form, which is being developed and evaluated by the Office of Fissile Materials Disposition of the United States Department of Energy (DOE). Potential accumulation of the fissile materials, 239 Pu and 235 U, in rock formations having a favorable chemical environment for such actions, requires analysis because autocatalytic configurations, while unlikely to form, never-the-less have consequences which are undesirable and require evaluation. Secondly, the WP design has evolved necessitating a re-evaluation of the internal WP degradation scenarios that contribute to the external source terms. The scope of this study includes a summary of the revised WP degradation calculations, a summary of the accumulation mechanisms in fractures and lithophysae in the tuff beneath the WP footprint, and a summary of the criticality risk calculations from any accumulated fissile material. Accumulations of fissile material external to the WP sufficient to pose a potential criticality risk require a deposition mechanism operating over sufficient time to reach required levels. The transporting solution concentrations themselves are well below critical levels (CRWMS 2001e). The ceramic waste form consists of Pu immobilized in ceramic disks, which would be embedded in High-Level Waste (HLW) glass in the standard HLW glass disposal canister. The ceramic disks would occupy approximately 12% of the HLW canister volume, while most of the remaining 88% of the volume would be occupied by HLW glass

  15. Examining the role of canister cooling conditions on the formation of nepheline from nuclear waste glasses

    Energy Technology Data Exchange (ETDEWEB)

    Christian, J. H. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-09-01

    Nepheline (NaAlSiO₄) crystals can form during slow cooling of high-level waste (HLW) glass after it has been poured into a waste canister. Formation of these crystals can adversely affect the chemical durability of the glass. The tendency for nepheline crystallization to form in a HLW glass increases with increasing concentrations of Al₂O₃ and Na₂O.

  16. 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

  17. Investigating the effect of compression on solute transport through degrading municipal solid waste

    Energy Technology Data Exchange (ETDEWEB)

    Woodman, N.D., E-mail: n.d.woodman@soton.ac.uk; Rees-White, T.C.; Stringfellow, A.M.; Beaven, R.P.; Hudson, A.P.

    2014-11-15

    Highlights: • The influence of compression on MSW flushing was evaluated using 13 tracer tests. • Compression has little effect on solute diffusion times in MSW. • Lithium tracer was conservative in non-degrading waste but not in degrading waste. • Bromide tracer was conservative, but deuterium was not. - Abstract: The effect of applied compression on the nature of liquid flow and hence the movement of contaminants within municipal solid waste was examined by means of thirteen tracer tests conducted on five separate waste samples. The conservative nature of bromide, lithium and deuterium tracers was evaluated and linked to the presence of degradation in the sample. Lithium and deuterium tracers were non-conservative in the presence of degradation, whereas the bromide remained effectively conservative under all conditions. Solute diffusion times into and out of less mobile blocks of waste were compared for each test under the assumption of dominantly dual-porosity flow. Despite the fact that hydraulic conductivity changed strongly with applied stress, the block diffusion times were found to be much less sensitive to compression. A simple conceptual model, whereby flow is dominated by sub-parallel low permeability obstructions which define predominantly horizontally aligned less mobile zones, is able to explain this result. Compression tends to narrow the gap between the obstructions, but not significantly alter the horizontal length scale. Irrespective of knowledge of the true flow pattern, these results show that simple models of solute flushing from landfill which do not include depth dependent changes in solute transport parameters are justified.

  18. 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

  19. 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

  20. 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

  1. Time to bypass the UK's stagnant waste programme

    International Nuclear Information System (INIS)

    Burton, W.R.; Haslam, C.J.

    1995-01-01

    It is envisaged that a big expansion of nuclear power will be required in the United Kingdom to meet the demand for electric power after gas supplies run out. However, an acceptable scheme for the disposal of all kinds of radioactive waste must be demonstrated before such an expansion is contemplated. Alternatives to the plans being developed by UK Nirex for the burial of low and intermediate level wastes (LLW and ILW) are advanced. The movement of groundwater which could carry radioactivity from an underground repository back to the land surface or into the sea is the main safety issue associated with burying nuclear waste. The water movement would be induced by the head of water from surrounding hillsides or by convection in water warmed by heat-emitting high level waste (HLW). By taking advantage of the coastal situation of both Sellafield and Dounreay, the two UK sites where waste is likely to be buried, these effects can be countered. Drained trench burial with a saline groundwater underpass created by drawing in seawater, is suggested for LLW and some short-lived ILW. A stagnant saline zone, again created from drawn in sea water, is proposed for deep disposal of ILW with a ''flyover'' to drain down surrounding hills. The disposal of HLW in liquid form in nitric acid solution at even deeper levels also making use of a stagnant saline zone is also discussed. (UK)

  2. Sampling plan to support HLW tank 16

    International Nuclear Information System (INIS)

    Rodwell, P.O.; Martin, B.

    1997-01-01

    Plans are to remove the residual waste from the annulus of High-Level Waste Tank 16, located in the H-Area Tank Farm, in 1998. The interior of the tank is virtually clean. In the late 1970's, the waste was removed from the interior of the tank by several campaigns of waste removal with slurry pumps, spray washing, and oxalic acid cleaning. The annulus of the tank at one time had several thousand gallons of waste salt, which had leaked from the tank interior. Some of this salt was removed by adding water to the annulus and circulating, but much of the salt remains in the annulus. In order to confirm the source term used for fate and transport modeling, samples of the tank interior and annulus will be obtained and analyzed. If the results of the analyses indicate that the data used for the initial modeling is bounding then no changes will be made to the model. However, if the results indicate that the source term is higher than that assumed in the initial modeling, thus not bounding, additional modeling will be performed. The purpose of this Plan is to outline the approach to sampling the annulus and interior of Tank 16 as a prerequisite to salt removal in the annulus and closure of the entire tank system. The sampling and analysis of this tank system must be robust to reasonably ensure the actual tank residual is within the bounds of analysis error

  3. Preliminary waste form characteristics report Version 1.0. Revision 1

    International Nuclear Information System (INIS)

    Stout, R.B.; Leider, H.R.

    1991-01-01

    This report focuses on radioactive waste form characteristics that will be used to design a waste package and an engineered barrier system (EBS) for a suitable repository as part of the Yucca Mountain Project. The term waste form refers to irradiated reactor fuel, other high-level waste (HLW) in various physical forms, and other radioactive materials (other than HLW) which are received for emplacement in a geologic repository. Any encapsulating of stabilizing matrix is also referred to as a waste form

  4. Analytic solution for one-dimensional diffusion of radionuclides from a waste package

    International Nuclear Information System (INIS)

    Oliver, D.L.

    1985-01-01

    This work implements an analytical solution for diffusion of radionuclides from a cylindrical waste form through the packing material into the surrounding host rock. Recent interest in predicting the performance of a proposed geological repository for nuclear waste has led to the development of several computer programs to predict the performance of such a repository for the next several millenia. These numerical codes are generally designed to accommodate a broad spectrum of geometrical configurations and repository conditions in order to accurately predict the behavior of the radionuclides in the repository environment. Confidence in such general purpose codes is gained by verifying the numerical modeling and the software through comparison of the numerical predictions generated by these computer codes with analytical solutions to reasonably complex problems. The analysis discussed herein implements the analytic solution, proposed by J.C. Jaeger in 1941 for radial diffusion through two concentric circular cylinders. Jaeger's solution was applied to the problem of diffusional mass transfer from a long cylindrical waste form and subsequently into the surrounding geological formation. Analytic predictions of fractional release rates, including the effects of sorption, were generated

  5. Corrosivity of solutions from evaporation of radioactive liquid wastes. Final report

    International Nuclear Information System (INIS)

    Payer, H.; Kolic, E.S.; Boyd, W.K.

    1977-01-01

    New double-shell storage tanks are constructed with ASTM A-516 Grade 65 steel. This study had two main objectives: To characterize the corrosivity of synthetic nonradioactive terminal waste solutions to ASTM A-516 Grade 65 steel and to determine the severity of stress-corrosion cracking of carbon steel in terminal waste solutions. The information developed provides guidance in the characterization of the aggressiveness of actual terminal liquors and in the design and operation of fail-safe tanks. Corrosion behavior was measured over a range of oxidizing conditions by the potentiodynamic polarization technique. Oxidizing conditions in a solution likely to promote general corrosion, pitting or stress-corrosion cracking (SCC) were identified. Absolute stress-corrosion cracking susceptibility was determined by constant strain rate procedure for ASTM A-516 Grade 65 steel for conditions identified by polarization experiments as likely to promote SCC. Based on the results of this study, terminal waste storage tanks are safe from stress-corrosion cracking under freely corroding conditions. Corrosion potential of steel in solutions within anticipated compositions is at the positive end of the critical range for stress-corrosion cracking, and no conditions were observed which would lower the potential to more negative values within the cracking range under freely corroding conditions. Measurement of corrosion potential and hydroxide concentration provides a means to extend these results to compositions outside of the composition range studied

  6. An analytic solution for one-dimensional diffusion of radionuclides from a waste package

    International Nuclear Information System (INIS)

    1985-01-01

    This work implements an analytical solution for diffusion of radionuclides from a cylindrical waste form through the packing material into the surrounding host rock. Recent interest in predicting the performance of a proposed geological repository for nuclear waste has led to the development of several computer programs to predict the performance of such a repository for the next several millenia. These numerical codes are generally designed to accommodate a broad spectrum of geometrical configurations and repository conditions in order to accurately predict the behavior of the radionuclides in the repository environment. Confidence in such general purpose codes is gained by verifying the numerical modeling and the software through comparison of the numerical predictions generated by these computer codes with analytical solutions to reasonably complex problems. The analysis discussed herein implements the analytic solution, proposed by J.C. Jaeger in 1941 for radial diffusion through two concentric circular cylinders. Jaeger's solution was applied to the problem of diffusional mass transfer from a long cylindrical waste form and subsequently into the surrounding geological formation. Analytic predictions of fractional release rates, including the effects of sorption, were generated. 6 refs., 2 figs., 2 tabs

  7. Regulatory dilemmas of a trans-solutional problem: Spatial and temporal isolation of radioactive wastes

    Energy Technology Data Exchange (ETDEWEB)

    Johnsrud, J.H. [State College, PA (United States). Sierra Club National Nuclear Waste Task Force

    1999-04-01

    This paper explores the `trans-solutional` nature of nuclear waste control - that it is in essence beyond human solution. To protect present and future human health, radioactive wastes require effective isolation from the biosphere for the full hazardous life of the wastes. Waste sequestration is essential to protect human beings, other forms of life in the bio-system, and the environment from adverse mutational impacts of exposures to ionizing radiation experienced in excess of those received from naturally-occurring background sources. The linear hypothesis of dose-response describes the relationship of radiation exposures to health at least with respect to cancer induction and hereditary genetic effects. The issue of risks of low-level radiation effects remains in controversy, with pressures exerted on regulators to ignore or minimize those impacts. However, recent research indicates that chronic low-dose exposures via inhalation and ingestion pathways may also give rise to non-fatal non-cancer deleterious health effects. Fatal cancers, now the primary measure of radiation injury in setting standards, may be less significant to a population in the long run than more subtle low-level impacts affecting genetic material. The latter, hard to identify or measure, may reduce developmental and reproductive capability. Given the hazardous longevity of high-level wastes, it is imperative that both protective standards and waste regulation be framed within an ethic of species responsibility. In our half century we have generated vast amounts of long-lived waste, with more promised in the coming millennium. The regulatory obligation is to isolate all nuclear wastes to best prevent any releases to the biosphere now but also to assure future generations an equal opportunity when our `disposal` methods inevitably fail over future time, to be able to retrieve and continue to isolate the wastes that we have caused to be produced. lt follows that the standards must not calculate

  8. Regulatory dilemmas of a trans-solutional problem: Spatial and temporal isolation of radioactive wastes

    International Nuclear Information System (INIS)

    Johnsrud, J.H.

    1999-01-01

    This paper explores the 'trans-solutional' nature of nuclear waste control - that it is in essence beyond human solution. To protect present and future human health, radioactive wastes require effective isolation from the biosphere for the full hazardous life of the wastes. Waste sequestration is essential to protect human beings, other forms of life in the bio-system, and the environment from adverse mutational impacts of exposures to ionizing radiation experienced in excess of those received from naturally-occurring background sources. The linear hypothesis of dose-response describes the relationship of radiation exposures to health at least with respect to cancer induction and hereditary genetic effects. The issue of risks of low-level radiation effects remains in controversy, with pressures exerted on regulators to ignore or minimize those impacts. However, recent research indicates that chronic low-dose exposures via inhalation and ingestion pathways may also give rise to non-fatal non-cancer deleterious health effects. Fatal cancers, now the primary measure of radiation injury in setting standards, may be less significant to a population in the long run than more subtle low-level impacts affecting genetic material. The latter, hard to identify or measure, may reduce developmental and reproductive capability. Given the hazardous longevity of high-level wastes, it is imperative that both protective standards and waste regulation be framed within an ethic of species responsibility. In our half century we have generated vast amounts of long-lived waste, with more promised in the coming millennium. The regulatory obligation is to isolate all nuclear wastes to best prevent any releases to the biosphere now but also to assure future generations an equal opportunity when our 'disposal' methods inevitably fail over future time, to be able to retrieve and continue to isolate the wastes that we have caused to be produced. lt follows that the standards must not calculate

  9. Nuclear waste disposal: perspective of a geochemist

    International Nuclear Information System (INIS)

    Sengupta, Pranesh; Dey, G.K.

    2011-01-01

    Satisfying the growing requirement in an environment friendly way is one of the most important tasks we need to accomplish these days. Considering the restricted non-renewable energy resources and limited technological progresses achieved in the renewable energy sectors in India, nuclear energy appears to be one of the most lucrative solutions towards the forthcoming energy crisis. Successful implementation of nuclear energy program however requires careful execution of high level nuclear waste management activities. One very important aspect of this process is to identify and develop suitable inert matrix(ces) for conditioning of nuclear waste(s) using natural analogue studies. And this establishes the very vital linkage between geochemical studies and nuclear waste immobilization. One good example of such an interdisciplinary approach can be seen in the methodologies adopted for immobilization of sulfate bearing high level nuclear wastes (SO 4 -HLW). It has been reported on several occasions that sulfur-rich melt get separated from silicate melt within magma chamber. Similar process has also been witnessed within vitrification furnaces whenever an attempt has been made to condition SO 4 -HLW within borosilicate glass matrices. Since such liquid-liquid phase separation leads to multiple difficulties in connection to radionuclide immobilization and plant scale vitrification processes, solutions were sought from natural analogue studies. Such as integrated approach ultimately resulted in establishing two different methodologies e.g. (i) modifying the borosilicate network through introduction of Ba 2+ cation; a process being followed in India and (ii) using phosphatic melt as a host instead of borosilicate melt; a process being followed in Russia. Detail of these two routes and the geochemical linkage in nuclear waste immobilization will be discussed.(author)

  10. Removal of fission products from waste solutions using 16 different soil samples

    International Nuclear Information System (INIS)

    Bangash, M.A.; Hanif, J.

    1997-01-01

    Most of the nuclear sites use pits in the surrounding soils for the storage/disposal of low active waste (LAW) solutions. The characteristics of the soil if not suitable for the fixation or adsorption of the radioactive nuclides, may cause migration of these nuclides to hydrosphere. The phenomenon has the risk of radio toxic pollution for the living bodies therefore minerals composing the soil and their adsorption properties need to be investigated. For this purpose 16 different soil samples were collected from all over Pakistan. Mineralogical composition of the soils was determined by X-ray diffraction analysis. It was found that most of the samples contained clay minerals, illite, kaolinite and montmorillonite. Studies for the removal of fission products like, /sup 137/Cs. /sup 60/Sr and activation product /sup 60/CO from solution were carried out on these samples. The sorption experiments were performed by batch technique using radioactive as tracers. Distribution co-efficient were determined by mixing he element solution at pH 3 with the soil at soil solution ratios of 1 to 20. It is revealed from the experimental data that efficient removal of fission products from solutions is achieved by soil samples containing clay mineral montmorillonite, followed by little and kaolinite. These soils thus can be effectively used for the disposal of low level radioactive waste solutions without causing any environmental hazard. (author)

  11. Engineering-scale vitrification of commercial high-level waste

    International Nuclear Information System (INIS)

    Bonner, W.F.; Bjorklund, W.J.; Hanson, M.S.; Knowlton, D.E.

    1980-04-01

    To date, technology for immobilizing commercial high-level waste (HLW) has been extensively developed, and two major demonstration projects have been completed, the Waste Solidification Engineering Prototypes (WSEP) Program and the Nuclear Waste Vitrification Project (NWVP). The feasibility of radioactive waste solidification was demonstrated in the WSEP program between 1966 and 1970 (McElroy et al. 1972) using simulated power-reactor waste composed of nonradioactive chemicals and HLW from spent, Hanford reactor fuel. Thirty-three engineering-scale canisters of solidified HLW were produced during the operations. In early 79, the NWVP demonstrated the vitrification of HLW from the processing of actual commercial nuclear fuel. This program consisted of two parts, (1) waste preparation and (2) vitrification by spray calcination and in-can melting. This report presents results from the NWVP

  12. Strategies and solutions in the temporary management of spent fuel in Spain

    International Nuclear Information System (INIS)

    Martinez Abad, J. E.; Rivera, M. I.

    2009-01-01

    The basic strategy for the spent fuel and HLW management contemplated in the Sixth General Radioactive Waste Plan focused on the centralised interim storage of spent fuel, based on proved dry storage system technologies, over the time periods required until their definitive or very long term management. Specially, the solution proposed as the most suitable for the Spanish case is the construction of a centralised interim spent fuel and HLW storage facility (ATC) for which as site is being searched. Until it becomes in operation, the interim spent fuel storage will be safety performed in the NPP reracked spent fuel pools or individual ISFSI constructed in the NPP site, in those cases additional storage capacity is required. (Author) 22 refs

  13. 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

  14. Brief historical perspective on the definition of high-level nuclear wastes

    International Nuclear Information System (INIS)

    Jacobs, D.G.; Szluha, A.T.; Gablin, K.A.; Croff, A.G.

    1985-03-01

    This report constitutes a historical perspective on the definition of HLW with emphasis on the US situation. The major HLW definitions are summarized chronologically, including a categorization of the considerations (e.g., waste source, heat generation rate, radiological effects) forming the bases of the definitions. High-level waste (HLW) definitions are then discussed in terms of these considerations. A brief discussion of the institutional aspects of HLW regulation and management are presented. An appendix to the report constitutes an annotated, chronological bibliography that formed the basis of the perspective

  15. MANAGEMENT OF SOLID WASTE GENERATED BY THE INTEGRATED STEELWORKS ACTIVITY AND SOLUTIONS TO REDUCE THE ENVIRONMENTAL IMPACT

    Directory of Open Access Journals (Sweden)

    Anişoara CIOCAN

    2010-05-01

    Full Text Available The development of steel industry is subject to solve major problems arising from industry-nature relationship, strictly targeted on pollution control and protection of natural resources and energy. In this paper we discussed about the management of solid waste generated by an integrated steelwork located near a major urban area and the adopted solutions for the reduction of environmental impact. There are summarized technical solutions that are currently applied and were proposed some solutions that can be applied in accordance with the environmental legislations. The new solutions are proposed for integrated management of solid wastes in accordance with: the exact quantification (quantitative, qualitative and the generation sources of emissions and solid wastes; controlled storage; minimization of the wastes and its harmfulness; transformation of the wastes into valuable by-products used directly by the company in a subsequent process, or by external down-stream user.

  16. Adsorption of Ag (I) from aqueous solution by waste yeast: kinetic, equilibrium and mechanism studies.

    Science.gov (United States)

    Zhao, Yufeng; Wang, Dongfang; Xie, Hezhen; Won, Sung Wook; Cui, Longzhe; Wu, Guiping

    2015-01-01

    One type of biosorbents, brewer fermentation industry waste yeast, was developed to adsorb the Ag (I) in aqueous solution. The result of FTIR analysis of waste yeast indicated that the ion exchange, chelating and reduction were the main binding mechanisms between the silver ions and the binding sites on the surface of the biomass. Furthermore, TEM, XRD and XPS results suggested that Ag(0) nanoparticles were deposited on the surface of yeast. The kinetic experiments revealed that sorption equilibrium could reach within 60 min, and the removal efficiency of Ag (I) could be still over 93 % when the initial concentration of Ag (I) was below 100 mg/L. Thermodynamic parameters of the adsorption process (ΔG, ΔH and ΔS) identified that the adsorption was a spontaneous and exothermic process. The waste yeast, playing a significant role in the adsorption of the silver ions, is useful to fast adsorb Ag (I) from low concentration.

  17. A nuclear waste deposit in space - the ultimate solution for low-cost and safe disposal

    International Nuclear Information System (INIS)

    Ruppe, H.O.; Hayn, D.; Braitinger, M.; Schmucker, R.H.

    1980-01-01

    The disposal of nuclear high-active waste (HAW) is representative for the problem of burdening the environment with highly active or toxic waste products at present and in the future. Safe disposal methods on Earth are technically very difficult to achieve and the costs of establishment and maintenance of such plants are extremely high. Furthermore the emotionally based rejection by a wide sector of the population gives sufficient reason to look for new solutions. Here, space technology can offer a real alternative - a waste deposit in space. With the Space Transportation System, which shall soon be operative, and the resulting high flight frequencies it will be possible to transport all future HAW into space at economical casts. (orig.) [de

  18. Treatment and conditioning of radioactive waste solution by natural clay minerals

    International Nuclear Information System (INIS)

    El-Dessouky, M.I.; El-Massry, E.H.; Khalifa, S.M.; Aly, H.F.

    1999-01-01

    Natural inorganic exchangers. Was used to remove caesium, cobalt and europium using zinc sulfate as coagulant also different clay minerals. These calys include, feldrspare, aswanly, bentionite, hematite, mud, calcite, basalt, magnetite, kaoline sand stone, limonite and sand. The factros affecting the removal process namely PH, particle size, temperature and weight of the clay have been studied. Highest removal for Cs-137, Co-60 and Eu-152 and 154 was achived by asswanly and bentonite. Sand stone is more effective than the other clays. Removal of Cs-137 from low level waste solution is in the order the sequence, aswanly (85.5%)> bentonite (82.2%)> sandstone (65.4%). Solidified cement products have been evaluated to determine optimum conditions of mixing most sludges contained clays by testing mechanical strength and leaching rates of the waste products. The solidified waste forms were found more acceptable for handing, storage and ultimate disposal

  19. Textile Dye Removal from Aqueous Solution using Modified Graphite Waste/Lanthanum/Chitosan Composite

    Science.gov (United States)

    Kusrini, E.; Wicaksono, B.; Yulizar, Y.; Prasetyanto, EA; Gunawan, C.

    2018-03-01

    We investigated various pre-treatment processes of graphite waste using thermal, mechanical and chemical methods. The aim of this work is to study the performance of modified graphite waste/lanthanum/chitosan composite (MG) as adsorbent for textile dye removal from aqueous solution. Effect of graphite waste resources, adsorbent size and lanthanum concentration on the dye removal were studied in batch experiments. Selectivity of MG was also investigated. Pre-heated graphite waste (NMG) was conducted at 80°C for 1 h, followed by mechanical crushing of the resultant graphite to 75 μm particle size, giving adsorption performance of ˜58%, ˜67%, ˜93% and ˜98% of the model dye rhodamine B (concentration determined by UV-vis spectroscopy at 554 nm), methyl orange (464 nm), methylene blue (664 nm) and methyl violet (580 nm), respectively from aqueous solution. For this process, the system required less than ˜5 min for adsorbent material to be completely saturated with the adsorbate. Further chemical modification of the pre-treated graphite waste (MG) with lanthanum (0.01 – V 0.03 M) and chitosan (0.5% w/w) did not improve the performance of dye adsorption. Under comparable experimental conditions, as those of the ‘thermal-mechanical-pre-treated-only’ (NMG), modification of graphite waste (MG) with 0.03 M lanthanum and 0.5% w/w chitosan resulted in ˜14%, ˜47%, ˜72% and ˜85% adsorption of rhodamine B, methyl orange, methylene blue and methyl violet, respectively. Selective adsorption of methylene blue at most to ˜79%, followed by methyl orange, methyl violet and rhodamine B with adsorption efficiency ˜67, ˜38, and ˜9% sequentially using MG with 0.03 M lanthanum and 0.5% w/w chitosan.

  20. 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

  1. Disposal of by-products in olive oil industry: waste-to-energy solutions

    International Nuclear Information System (INIS)

    Caputo, Antonio C.; Scacchia, Federica; Pelagagge, Pacifico M.

    2003-01-01

    Olive oil production industry is characterized by relevant amounts of liquid and solid by-products [olive mill wastewater (OMW) and olive husk (OH)], and by economical, technical and organizational constraints that make difficult the adoption of environmentally sustainable waste disposal approaches. In this context, waste treatment technologies aimed at energy recovery represent an interesting alternative. In the paper, a technical and economical analysis of thermal disposal plant solutions with energy recovery has been carried out. The considered plants enable the combined treatment of OMW and OH which, although penalizes the energy recovery, proves to be feasible and profitable in a future legislative scenario when stricter limitation on OMW disposal will force oil producers to bear high disposal costs. Results are compared by using economic performance measures, including revenues from produced energy and avoided disposal costs. A sensitivity and risk analysis is also performed in order to assess the economic profitability of the proposed solutions

  2. Long-term effects of waste solutions on concrete and reinforcing steel

    International Nuclear Information System (INIS)

    Daniel, J.I.; Stark, D.C.; Kaar, P.H.

    1982-04-01

    This report has been prepared for the In Situ Waste Disposal Program Tank Assessment Task (WG-11) as part of an investigation to evaluate the long-term performance of waste storage tanks at the Hanford Site. This report, prepared by the Portland Cement Association, presents the results of four years of concrete degradation studies which exposed concrete and reinforcing steel, under load and at 180 0 F, to simulated double-shell slurry, simulated salt cake solution, and a control solution. Exposure length varied from 3 months to 36 months. In all cases, examination of the concrete and reinforcing steel at the end of the exposure indicated there was no attack, i.e., no evidence of rusting, cracking, disruption of mill scale or loss of strength

  3. Recovery of thorium along with uranium 233 from Thorex waste solution employing Chitosan

    International Nuclear Information System (INIS)

    Priya, S.; Reghuram, D.; Kumaraguru, K.; Vijayan, K.; Jambunathan, U.

    2003-01-01

    The low level waste solution, generated from Thorex process during the processing of U 233 , contains thorium along with traces of Th 228 and U 233 . Chitosan, a natural bio-polymer derived from Chitin, was earlier used to recover the uranium and americium. The studies were extended to find out its thorium sorption characteristics. Chitosan exhibited very good absorption of thorium (350 mg/g). Chitosan was equilibrated directly with the low level waste solution at different pH after adjusting its pH, for 60 minutes with a Chitosan to aqueous ratio of 1:100 and the raffinates were filtered and analysed. The results showed more than 99% of thorium and U 233 could be recovered by Chitosan between pH 4 and 5. Loaded thorium and uranium could be eluted from the Chitosan by 1M HNO 3 quantitatively. (author)

  4. Evaluation of extractant-coated magnetic microparticles for the recovery of hazardous metals from waste solution

    International Nuclear Information System (INIS)

    Kaminski, M. D.

    1998-01-01

    A magnetically assisted chemical separation (MACS) process was developed earlier at Argonne National Laboratory (ANL). This compact process was designed for the separation of transuranics (TRU) and radionuclides from the liquid waste streams that exist at many DOE sites, with an overall reduction in waste volume requiring disposal. The MACS process combines the selectivity afforded by solvent extractant/ion exchange materials with magnetic separation to provide an efficient chemical separation. Recently, the MACS process has been evaluated with acidic organophosphorus extractants for hazardous metal recovery from waste solutions. Moreover, process scale-up design issues have been addressed with respect to particle filtration and recovery. Two acidic organophosphorus compounds have been investigated for hazardous metal recovery, bis(2,4,4-trimethylpentyl) phosphinic acid (Cyanexreg-sign 272) and bis(2,4,4-trimethylpentyl) dithiophosphinic acid (Cyanexreg-sign 301). Coated onto magnetic microparticles, these extractants demonstrated superior recovery of hazardous metals from solution, relative to what was expected on the basis of results from solvent extraction experiments. The results illustrate the diverse applications of MACS technology for dilute waste streams. Preliminary process scale-up experiments with a high-gradient magnetic separator at Oak Ridge National Laboratory have revealed that very low microparticle loss rates are possible

  5. Applying Lean Techniques to Reduce Intravenous Waste Through Premixed Solutions and Increasing Production Frequency.

    Science.gov (United States)

    Lin, Alex C; Penm, Jonathan; Ivey, Marianne F; Deng, Yihong; Commins, Monica

    This study aims to use lean techniques and evaluate the impact of increasing the use of premixed IV solutions and increased IV production frequency on IV waste. Study was conducted at a tertiary hospital pharmacy department in three phases. Phase I included evaluation of IV waste when IV production occurred three times a day and eight premixed IV products were used. Phase II increased the number of premixed IV products to 16. Phase III then increased IV production to five times a day. During Phase I, an estimate of 2,673 IV doses were wasted monthly, accounting for 6.14% of overall IV doses. This accounted for 688 L that cost $60,135. During Phase II, the average monthly IV wastage reduced significantly to 1,069 doses (2.84%), accounting for 447 L and $34,003. During Phase III, the average monthly IV wastage was further decreased to 675 doses (1.69%), accounting for 78 L and $3,431. Hence, a potential annual saving of $449,208 could result from these changes. IV waste was reduced through the increased use of premixed solutions and increasing IV production frequency.

  6. Processing results of 1800 gallons of mercury and radioactively contaminated mixed waste rinse solution

    International Nuclear Information System (INIS)

    Thiesen, B.P.

    1993-01-01

    Mercury-contaminated rinse solution was successfully treated at the Idaho National Engineering Laboratory. This waste was generated during the decontamination of the Heat Transfer Reactor Experiment 3 reactor shield tank. Approximately 6.8 m 3 (1,800 pi) of waste was generated and placed into 33 drums. Each drum contained precipitated sludge material ranging from 2--5 cm in depth, with the average depth of about 6 cm. The pH of each drum varied from 3--11. The bulk liquid waste had a mercury level of 7.0 mg/l, which exceeded the Resource Conservation and Recovery Act limit of 0.2 mg/l. The average liquid bulk radioactivity was about 2.1 pCi/mL while the average sludge contamination was about 13,800 pCi/g. Treatment of the waste required separation of the liquid from the sludge, filtration, pH adjustment, and ion exchange. The resulting solution after treatment had mercury levels at 0.0186 mg/l and radioactivity of 0.282 pCi/ml

  7. Removal of radioruthenium from alkaline intermediate level radioactive waste solution : a laboratory investigation

    International Nuclear Information System (INIS)

    Samanta, S.K.; Theyyunni, T.K.

    1994-01-01

    Various methods were investigated in the laboratory for the removal of radioruthenium from alkaline intermediate level radioactive waste solutions of reprocessing plant origin. The methods included batch equilibration with different ion exchangers and sorbents, column testing and chemical precipitation. A column method using zinc-activated carbon mixture and a chemical precipitation method using ferrous salt along with sodium sulphite were found to be promising for plant scale application. (author). 10 refs., 3 figs., 7 tabs

  8. Pore solution chemistry of simulated low-level liquid waste incorporated in cement grouts

    International Nuclear Information System (INIS)

    Kruger, A.A.

    1995-12-01

    Expressed pore solutions from simulated low level liquid waste cement grouts cured at room temperature, 50 degree C and 90 degree C for various duration were analyzed by standard chemical methods and ion chromatography. The solid portions of the grouts were formulated with portland cement, fly ash, slag, and attapulgite clay in the ratios of 3:3:3:1. Two different solutions simulating off-gas condensates expected from vitrification of Hanford low level tank wastes were made. One is highly alkaline and contains the species Na + , P0 4 3- , N0 2 - , NO 3 - and OH - . The other is carbonated and contains the species, Na + , PO 4 3- , NO 2 - , NO 3 - , and CO 3 2- . In both cases phosphate rapidly disappeared from the pore solution, leaving behind sodium in the form of hydroxide. The carbonates were also removed from the pore solution to form calcium carbonate and possibly calcium monocarboaluminate. These reactions resulted in the increase of hydroxide ion concentration in the early period. Subsequently there was a significant reduction OH - and Na + ion concentrations. In contrast high concentration of N0 2 - and N0 3 - were retained in the pore solution indefinitely

  9. Copper-Sulfate Pentahydrate as a Product of the Waste Sulfuric Acid Solution Treatment

    Science.gov (United States)

    Marković, Radmila; Stevanović, Jasmina; Avramović, Ljiljana; Nedeljković, Dragutin; Jugović, Branimir; Stajić-Trošić, Jasna; Gvozdenović, Milica

    2012-12-01

    The aim of this study is synthesis of copper-sulfate pentahydrate from the waste sulfuric acid solution-mother liquor generated during the regeneration process of copper bleed solution. Copper is removed from the mother liquor solution in the process of the electrolytic treatment using the insoluble lead anodes alloyed with 6 mass pct of antimony on the industrial-scale equipment. As the result of the decopperization process, copper is removed in the form of the cathode sludge and is precipitated at the bottom of the electrolytic cell. By this procedure, the content of copper could be reduced to the 20 mass pct of the initial value. Chemical characterization of the sludge has shown that it contains about 90 mass pct of copper. During the decopperization process, the very strong poison, arsine, can be formed, and the process is in that case terminated. The copper leaching degree of 82 mass pct is obtained using H2SO4 aqueous solution with the oxygen addition during the cathode sludge chemical treatment at 80 °C ± 5 °C. Obtained copper salt satisfies the requirements of the Serbian Standard for Pesticide, SRPS H.P1. 058. Therefore, the treatment of waste sulfuric acid solutions is of great economic and environmental interest.

  10. 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

  11. Idaho National Engineering Laboratory High-Level Waste Roadmap. Revision 2

    Energy Technology Data Exchange (ETDEWEB)

    1993-08-01

    The Idaho National Engineering Laboratory (INEL) High-Level Waste (HLW) Roadmap takes a strategic look at the entire HLW life-cycle starting with generation, through interim storage, treatment and processing, transportation, and on to final disposal. The roadmap is an issue-based planning approach that compares ``where we are now`` to ``where we want and need to be.`` The INEL has been effectively managing HLW for the last 30 years. Calcining operations are continuing to turn liquid HLW into a more manageable form. Although this document recognizes problems concerning HLW at the INEL, there is no imminent risk to the public or environment. By analyzing the INEL current business operations, pertinent laws and regulations, and committed milestones, the INEL HLW Roadmap has identified eight key issues existing at the INEL that must be resolved in order to reach long-term objectives. These issues are as follows: A. The US Department of Energy (DOE) needs a consistent policy for HLW generation, handling, treatment, storage, and disposal. B. The capability for final disposal of HLW does not exist. C. Adequate processes have not been developed or implemented for immobilization and disposal of INEL HLW. D. HLW storage at the INEL is not adequate in terms of capacity and regulatory requirements. E. Waste streams are generated with limited consideration for waste minimization. F. HLW is not adequately characterized for disposal nor, in some cases, for storage. G. Research and development of all process options for INEL HLW treatment and disposal are not being adequately pursued due to resource limitations. H. HLW transportation methods are not selected or implemented. A root-cause analysis uncovered the underlying causes of each of these issues.

  12. Use of almond endocarp shell in sorption of radioactive 152+154Europium from waste solutions

    International Nuclear Information System (INIS)

    Dakroury, G.A.; Khalil, T.; Abou El-Nour, F.H.

    2007-01-01

    In an attempt to remove radioactive ( 152 + 154 )Eu from waste solutions, the present study was tried to explore the possibility of using a natural by-product. Almond endocarp (AEC) shell produced from Sinai (El-Arish area) was selected as agricultural by-product in treatment of waste solutions containing ( 152 + 154 )Eu through a batch technique. The different physico-chemical characteristics of AEC such as specific surface area, total pore volume, average pore diameter, apparent density, porosity and pore size distribution were calculated. The adsorption process was described by a Freundlich type isotherm. The uptake percent of the metal ion was determined for the sorbent material as a function of contact time, pH-value, mass of the sorbent material, metal ion concentration and the effect of competing ions on the sorption process. The obtained data were analyzed and showed that almond endocarp shell powder can be considered as an efficient natural material to be used for sorption of radioactive ( 152 + 154 )Eu from their radioactive waste solutions

  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. DEMONSTRATION OF THE NEXT-GENERATION CAUSTIC-SIDE SOLVENT EXTRACTION SOLVENT WITH 2-CM CENTRIGUGAL CONTRACTORS USING TANK 49H WASTE AND WASTE SIMULANT

    Energy Technology Data Exchange (ETDEWEB)

    Pierce, R.; Peters, T.; Crowder, M.; Pak, D.; Fink, S.; Blessing, R.; Washington, A.; Caldwell, T.

    2011-11-29

    Researchers successfully demonstrated the chemistry and process equipment of the Caustic-Side Solvent Extraction (CSSX) flowsheet using MaxCalix for the decontamination of high level waste (HLW). The demonstration was completed using a 12-stage, 2-cm centrifugal contactor apparatus at the Savannah River National Laboratory (SRNL). This represents the first CSSX process demonstration of the MaxCalix solvent system with Savannah River Site (SRS) HLW. Two tests lasting 24 and 27 hours processed non-radioactive simulated Tank 49H waste and actual Tank 49H HLW, respectively. A solvent extraction system for removal of cesium from alkaline solutions was developed utilizing a novel solvent invented at the Oak Ridge National Laboratory (ORNL). This solvent consists of a calix[4]arene-crown-6 extractant dissolved in an inert hydrocarbon matrix. A modifier is added to the solvent to enhance the extraction power of the calixarene and to prevent the formation of a third phase. An additional additive is used to improve stripping performance and to mitigate the effects of any surfactants present in the feed stream. The process that deploys this solvent system is known as Caustic Side Solvent Extraction (CSSX). The solvent system has been deployed at the Savannah River Site (SRS) in the Modular CSSX Unit (MCU) since 2008.

  16. 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

  17. 3D numerical modelling of the thermal state of deep geological nuclear waste repositories

    Science.gov (United States)

    Butov, R. A.; Drobyshevsky, N. I.; Moiseenko, E. V.; Tokarev, Yu. N.

    2017-09-01

    One of the important aspects of the high-level radioactive waste (HLW) disposal in deep geological repositories is ensuring the integrity of the engineered barriers which is, among other phenomena, considerably influenced by the thermal loads. As the HLW produce significant amount of heat, the design of the repository should maintain the balance between the cost-effectiveness of the construction and the sufficiency of the safety margins, including those imposed on the thermal conditions of the barriers. The 3D finite-element computer code FENIA was developed as a tool for simulation of thermal processes in deep geological repositories. Further the models for mechanical phenomena and groundwater hydraulics will be added resulting in a fully coupled thermo-hydro-mechanical (THM) solution. The long-term simulations of the thermal state were performed for two possible layouts of the repository. One was based on the proposed project of Russian repository, and another features larger HLW amount within the same space. The obtained results describe the spatial and temporal evolution of the temperature filed inside the repository and in the surrounding rock for 3500 years. These results show that practically all generated heat was ultimately absorbed by the host rock without any significant temperature increase. Still in the short time span even in case of smaller amount of the HLW the temperature maximum exceeds 100 °C, and for larger amount of the HLW the local temperature remains above 100 °C for considerable time. Thus, the substantiation of the long-term stability of the repository would require an extensive study of the materials properties and behaviour in order to remove the excessive conservatism from the simulations and to reduce the uncertainty of the input data.

  18. Solution exchange corrosion testing with the glass-zeolite ceramic waste form in demineralized water at 900C

    International Nuclear Information System (INIS)

    Simpson, L. J.

    1998-01-01

    A ceramic waste form of glass-bonded zeolite is being developed for the long-term disposition of fission products and transuranic elements in wastes from the U.S. Department of Energy's spent nuclear fuel conditioning activities. Solution exchange corrosion tests were performed on the ceramic waste form and its potential base constituents of glass, zeolite 5A, and sodalite as part of an effort to qualify the ceramic waste form for acceptance into the Civilian Radioactive Waste Management System. Solution exchange tests were performed at 90 C by replacing 80 to 90% of the leachate with fresh demineralized water after set time intervals. The results from these tests provide information about corrosion mechanisms and the ability of the ceramic waste form and its constituent materials to retain waste components. The results from solution exchange tests indicate that radionuclides will be preferentially retained in the zeolites without the glass matrix and in the ceramic waste form, with respect to cations like Li, K, and Na. Release results have been compared for simulated waste from candidate ceramic waste forms with zeolite 5A and its constituent materials to determine the corrosion behavior of each component

  19. Siting Process for HLW Repository in Japan

    International Nuclear Information System (INIS)

    Masuda, S.; Kitayama, K.; Umeki, H.; Naito, M.

    2002-01-01

    In the year 2000, the geological disposal program for high-level radioactive waste in Japan moved from the phase of generic research and development (R and D) into the phase of implementation. Following legislation entitled the ''Specified Radioactive Waste Final Disposal Act'', the Nuclear Waste Management Organization of Japan (NUMO) was established as the implementing organization. The assigned activities of NUMO include selection of the repository site, demonstration of disposal technology at the site, developing relevant licensing applications and construction, operation and closure of the repository. As the first milestone of siting process, NUMO announced to the public an overall procedure for selection of preliminary investigation areas for potential candidate sites on October 29, 2001. The procedure specifies that NUMO will solicit volunteer municipalities for preliminary investigation areas with publishing four documents as an information package. These documents are tentatively entitled ''Instructions for Application'', ''Siting Factors for the Preliminary Investigation Areas'', a ''Repository Concepts'' as well as an ''Site Investigation Community Outreach Scheme''

  20. Current Understanding and Remaining Challenges in Modeling Long-Term Degradation of Borosilicate Nuclear Waste Glasses

    International Nuclear Information System (INIS)

    Vienna, John D.; Ryan, Joseph V.; Gin, Stephane; Inagaki, Yaohiro

    2013-01-01

    Chemical durability is not a single material property that can be uniquely measured. Instead it is the response to a host of coupled material and environmental processes whose rates are estimated by a combination of theory, experiment, and modeling. High-level nuclear waste (HLW) glass is perhaps the most studied of any material yet there remain significant technical gaps regarding their chemical durability. The phenomena affecting the long-term performance of HLW glasses in their disposal environment include surface reactions, transport properties to and from the reacting glass surface, and ion exchange between the solid glass and the surrounding solution and alteration products. The rates of these processes are strongly influenced and are coupled through the solution chemistry, which is in turn influenced by the reacting glass and also by reaction with the near-field materials and precipitation of alteration products. Therefore, those processes must be understood sufficiently well to estimate or bound the performance of HLW glass in its disposal environment over geologic time-scales. This article summarizes the current state of understanding of surface reactions, transport properties, and ion exchange along with the near-field materials and alteration products influences on solution chemistry and glass reaction rates. Also summarized are the remaining technical gaps along with recommended approaches to fill those technical gaps

  1. Tank Waste Remediation System optimized processing strategy

    International Nuclear Information System (INIS)

    Slaathaug, E.J.; Boldt, A.L.; Boomer, K.D.; Galbraith, J.D.; Leach, C.E.; Waldo, T.L.

    1996-03-01

    This report provides an alternative strategy evolved from the current Hanford Site Tank Waste Remediation System (TWRS) programmatic baseline for accomplishing the treatment and disposal of the Hanford Site tank wastes. This optimized processing strategy performs the major elements of the TWRS Program, but modifies the deployment of selected treatment technologies to reduce the program cost. The present program for development of waste retrieval, pretreatment, and vitrification technologies continues, but the optimized processing strategy reuses a single facility to accomplish the separations/low-activity waste (LAW) vitrification and the high-level waste (HLW) vitrification processes sequentially, thereby eliminating the need for a separate HLW vitrification facility

  2. Preoperational assessment of solute release from waste rock at proposed mining operations

    International Nuclear Information System (INIS)

    Lapakko, Kim A.

    2015-01-01

    Highlights: • Modeling to estimate solute release from waste rock at proposed mines is described. • Components of the modeling process are identified and described. • Modeling inputs required are identified and described. • Examples of data generated and their application are presented. • Challenges inherent to environmental review are identified. - Abstract: Environmental assessments are conducted prior to mineral development at proposed mining operations. Among the objectives of these assessments is prediction of solute release from mine wastes projected to be generated by the proposed mining and associated operations. This paper provides guidance to those engaged in these assessments and, in more detail, provides insights on solid-phase characterization and application of kinetic test results for predicting solute release from waste rock. The logic guiding the process is consistent with general model construction practices and recent publications. Baseline conditions at the proposed site are determined and a detailed operational plan is developed and imposed upon the site. Block modeling of the mine geology is conducted to identify the mineral assemblages present, their masses and compositional variations. This information is used to select samples, representative of waste rock to be generated, that will be analyzed and tested to describe characteristics influencing waste rock drainage quality. The characterization results are used to select samples for laboratory dissolution testing (kinetic tests). These tests provide empirical data on dissolution of the various mineral assemblages present as waste rock. The data generated are used, in conjunction with environmental conditions, the proposed method of mine waste storage, and scientific and technical principles, to estimate solute release rates for the operational scale waste rock. Common concerns regarding waste rock are generation of acidic drainage and release of heavy metals and sulfate. Key solid

  3. 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

  4. 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

  5. Waste acid/metal solution reduction and recovery by vacuum distillation

    International Nuclear Information System (INIS)

    Jones, E.O.; Wilcox, W.A.; Johnson, N.T.; Bowdish, F.W.

    1995-01-01

    Processes involving distillation under reduced pressure were developed at the Pacific Northwest Laboratory several years ago to recover spent acid solutions generated during the manufacture of nuclear fuel for the N-Reactor at the Hanford site. Following construction and testing of a pilot-plant, the technology was licensed to Viatec Recovery Systems, Inc. for commercialization. The technology developed included specialized distillation and rectification of volatile acids, removal of water and/or volatile acid from sulfuric acid, and precipitation of salts. A key feature of the Waste Acid Detoxification and Reclamation (WADR) technology is the development and use of advanced thermoplastic and fluoropolymer materials of construction in all critical process equipment. The technology was then expanded to include crystallization to recover metal salts for possible reuse. Economic and environmental advantages of the procedures include recovery of acids for reuse, simplification or elimination of the disposal of waste solutions, and possible recovery of metals. Industries expected to benefit from such applications include galvanizing, electroplating, sand leaching and any where metals are cleaned in acid solutions. Currently a modular system has been assembled for recovery of several different spent acid solutions

  6. 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

  7. 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

  8. 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

  9. 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.

  10. The basic corrosion mechanisms of HLW glasses

    International Nuclear Information System (INIS)

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

    1986-01-01

    During the years 1975 to 1984, the Commission of the European Communities organized and promoted an R and D programme on the testing and evaluation of solidified high-level waste forms with the purpose of providing a scientific basis for the management and storage of radioactive waste. A fair number of materials were tested under a broad variation of experimental data. The Fraunhofer-Institut fuer Silicatforschung, Wuerzburg, has undertaken to perform a synoptic evaluation of the above data. The purpose of this evaluation is: - to compile the data from the individual national contributors (as presented in the joint annual reports of the EC) with respect to: the materials, or the experimental parameters, or further aspects, and to harmonize them with respect to their presentation, choice of units, etc., - to compare the results to the international state of information, - to elaborate and demonstrate common features of the diverse materials, e.g. common patterns of the corrosion behaviour, - to check the validity of present models, - to define shortcomings and questions that are still open

  11. Recovery of gold from hydrometallurgical leaching solution of electronic waste via spontaneous reduction by polyaniline

    Directory of Open Access Journals (Sweden)

    Yuanzhao Wu

    2017-08-01

    Full Text Available The present study is primarily designed to develop an environmentally-benign approach for the recovery of precious metals, especially gold, from the ever increasingly-discarded electronic wastes (e-waste. By coupling the metal reduction process with an increase in the intrinsic oxidation state of the aniline polymers, and the subsequent re-protonation and reduction of the intrinsically oxidized polymer to the protonated emeraldine (EM salt, polyaniline (PANi films and polyaniline coated cotton fibers are able to recover metallic gold from acid/halide leaching solutions of electronic wastes spontaneously and sustainably. The current technique, which does not require the use of extensive extracting reagents or external energy input, can recover as much as 90% of gold from the leaching acidic solutions. The regeneration of polyaniline after gold recovery, as confirmed by the X-ray photoelectron spectroscopy measurements, promises the continuous operation using the current approach. The as-recovered elemental gold can be further concentrated and purified by incineration in air.

  12. Removal of phenol from radioactive waste solutions using activated granular Carbon and activated vermiculite

    International Nuclear Information System (INIS)

    Ezz El-Din, M.R.; Atta, E.R.

    2006-01-01

    The efficiency of both activated granular carbon (AGC) and activated vermiculite (AV) in removal of phenol from aqueous waste solutions is of great interest. The aim of the present study is to compare the absorbance capacities of both AGC and AV for the removal of phenol from radioactive waste solutions and to identify the factors affecting the sorption process. The experimental results were in the form of batch sorption measurements for the removal of phenol at ambient temperature (29 ± 1 degree C) and for times up to 40 min and 180 min for AGC and AV, respectively. The results indicated that activated carbon has good efficiency to adsorb phenol. Freundlich equation has been fitted to both AGC and AV for the contaminant removal. The adsorption capacities of both AGC and AV to phenol were 17.4 mg g-1 and 4.5 mg g-1, respectively. The maximum desorption percent of phenol from both loaded AGC and loaded AV were 9 % and 0 %, respectively, and it attained within about 200 min. accordingly, it is recommended that activated carbon is preferred in the applied field for removing phenol from radioactive aqueous wastes

  13. Treatment and Conditioning of Radioactive Waste Solution by Natural Clay Minerals

    International Nuclear Information System (INIS)

    El-Dessouky, M.I.; Abdel-Raouf, M.W.; El-Massry, E.H.; Khalifa, S.M.; Aly, H.F.

    1999-01-01

    Chemical precipitation processes have been used for the treatment of radioactive elements from aqueous solution. The volume reduction is not very great and storage facilities are expensive. There are some radionuclides which are so difficult to be precipitated by this common method, so they may be precipitated by adding solid materials such as natural inorganic exchangers. In this woek, improvement the removal of caesium, cobalt and europium with zinc sulfate as coagulant and different clay minerals have been investigated. These include, Feldespare, Aswanly, Bentionite, Hematite, Mud, Calcite, Basalt, Magnetite, Kaoline, Sand stone, Limonite and Sand. The parameters affecting the precipitation process such as pH, particle size, temperature and weight of the clay have been studied. The results indicate that, the highest removal for Cs-137, Co-60 and Eu-152 and154 by Asswanly, Bentonite and Sand stone is more than the other clays. Removal of Cs-137 from low level waste solution with these three natural clays took the sequence, Aswanly (85.5%) > Bentonite (82.2%) > Sandstone (65.4%). Solidified cement products have been evaluated to determine mechanical strength and leaching rates of the waste products. The solidified waste forms were found more acceptable for handling ,storage and ultimate disposal

  14. DOUBLE SHELL TANK INTEGRITY PROJECT HIGH LEVEL WASTE CHEMISTRY OPTIMIZATION

    International Nuclear Information System (INIS)

    WASHENFELDER DJ

    2008-01-01

    The U.S. Department of Energy's Office (DOE) of River Protection (ORP) has a continuing program for chemical optimization to better characterize corrosion behavior of High-Level Waste (HLW). The DOE controls the chemistry in its HLW to minimize the propensity of localized corrosion, such as pitting, and stress corrosion cracking (SCC) in nitrate-containing solutions. By improving the control of localized corrosion and SCC, the ORP can increase the life of the Double-Shell Tank (DST) carbon steel structural components and reduce overall mission costs. The carbon steel tanks at the Hanford Site are critical to the mission of safely managing stored HLW until it can be treated for disposal. The DOE has historically used additions of sodium hydroxide to retard corrosion processes in HLW tanks. This also increases the amount of waste to be treated. The reactions with carbon dioxide from the air and solid chemical species in the tank continually deplete the hydroxide ion concentration, which then requires continued additions. The DOE can reduce overall costs for caustic addition and treatment of waste, and more effectively utilize waste storage capacity by minimizing these chemical additions. Hydroxide addition is a means to control localized and stress corrosion cracking in carbon steel by providing a passive environment. The exact mechanism that causes nitrate to drive the corrosion process is not yet clear. The SCC is less of a concern in the newer stress relieved double shell tanks due to reduced residual stress. The optimization of waste chemistry will further reduce the propensity for SCC. The corrosion testing performed to optimize waste chemistry included cyclic potentiodynamic volarization studies. slow strain rate tests. and stress intensity factor/crack growth rate determinations. Laboratory experimental evidence suggests that nitrite is a highly effective:inhibitor for pitting and SCC in alkaline nitrate environments. Revision of the corrosion control

  15. Removal of actinides from nuclear fuel reprocessing waste solutions with bidentate organophosphorus extractants

    International Nuclear Information System (INIS)

    Schulz, W.W.; McIsaac, L.D.

    1975-08-01

    The neutral bidentate organophosphorus reagents DBDECMP (dibutyl-N,N-diethylcarbamylmethylenephosphonate) and its dihexyl analogue DHDECMP are candidate extractants for removal of actinides from certain acidic waste streams produced at the U. S. ERDA Hanford and Idaho Falls sites. Various chemical and physical properties including availability, cost, purification, alpha radiolysis, and aqueous phase solubility of DBDECMP and DHDECMP are reviewed. A conceptual flowsheet employing a 15 percent DBDECMP (or DHDECMP)--CCl 4 extractant for removal (and recovery) of Am and Pu from Hanford's Plutonium Reclamation Facility acid waste stream (CAW solution) was successfully demonstrated in laboratory-scale mixer-settler tests; this extraction scheme can be used to produce an actinide-free waste. A 30 percent DBDECMP-xylene flowsheet is being tested at the Idaho Falls site for removal of U, Np, Pu, and Am from Idaho Chemical Processing Plant first-cycle high-level raffinate to produce an actinide-free (less than 10 nCi alpha activity/gram) waste. (auth)

  16. 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)

  17. 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

  18. 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

  19. 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

  20. 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

  1. [Bioregeneration of the solutions obtained during the leaching of nonferrous metals from waste slag by acidophilic microorganisms].

    Science.gov (United States)

    Fomchenko, N V; Murav'ev, M I; Kondrat'eva, T F

    2014-01-01

    The bioregeneration of the solutions obtained after the leaching of copper and zinc from waste slag by sulfuric solutions of ferric sulfate is examined. For bioregeneration, associations of mesophilic and moderately thermqophilic acidophilic chemolithotrophic microorganisms were made. It has been shown that the complete oxidation of iron ions in solutions obtained after the leaching of nonferrous metals from waste slag is possible at a dilution of the pregnant solution with a nutrient medium. It has been found that the maximal rate of oxidation of iron ions is observed at the use of a mesophilic association of microorganisms at a threefold dilution of the pregnant solution with a nutrient medium. The application ofbioregeneration during the production of nonferrous metals from both waste and converter slags would make it possible to approach the technology of their processing using the closed cycle of workflows.

  2. Supplemental Immobilization Cast Stone Technology Development and Waste Form Qualification Testing Plan

    Energy Technology Data Exchange (ETDEWEB)

    Westsik, Joseph H. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Serne, R. Jeffrey [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Pierce, Eric M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Cozzi, Alex [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Chung, Chul-Woo [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Swanberg, David J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2013-05-31

    The Hanford Tank Waste Treatment and Immobilization Plant (WTP) is being constructed to treat the 56 million gallons of radioactive waste stored in 177 underground tanks at the Hanford Site. The WTP includes a pretreatment facility to separate the wastes into high-level waste (HLW) and low-activity waste (LAW) fractions for vitrification and disposal. The LAW will be converted to glass for final disposal at the Integrated Disposal Facility (IDF). The pretreatment facility will have the capacity to separate all of the tank wastes into the HLW and LAW fractions, and the HLW Vitrification Facility will have the capacity to vitrify all of the HLW. However, a second immobilization facility will be needed for the expected volume of LAW requiring immobilization. A number of alternatives, including Cast Stone—a cementitious waste form—are being considered to provide the additional LAW immobilization capacity.

  3. 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

  4. Bioprecipitation of uranium from alkaline waste solutions using recombinant Deinococcus radiodurans

    Energy Technology Data Exchange (ETDEWEB)

    Kulkarni, Sayali; Ballal, Anand; Apte, Shree Kumar, E-mail: aptesk@barc.gov.in

    2013-11-15

    Highlights: • Deinococcus radiodurans was genetically engineered to overexpress alkaline phosphatase (PhoK). • Deino-PhoK bioprecipitated U efficiently over a wide range of input U concentration. • A maximal loading of 10.7 g U/g of biomass at 10 mM input U was observed. • Radioresistance and U precipitation by Deino-PhoK remained unaffected by γ radiation. • Immobilization of Deino-PhoK facilitated easy separation of precipitated U. -- Abstract: Bioremediation of uranium (U) from alkaline waste solutions remains inadequately explored. We engineered the phoK gene (encoding a novel alkaline phosphatase, PhoK) from Sphingomonas sp. for overexpression in the radioresistant bacterium Deinococcus radiodurans. The recombinant strain thus obtained (Deino-PhoK) exhibited remarkably high alkaline phosphatase activity as evidenced by zymographic and enzyme activity assays. Deino-PhoK cells could efficiently precipitate uranium over a wide range of input U concentrations. At low uranyl concentrations (1 mM), the strain precipitated >90% of uranium within 2 h while a high loading capacity of around 10.7 g U/g of dry weight of cells was achieved at 10 mM U concentration. Uranium bioprecipitation by Deino-PhoK cells was not affected in the presence of Cs and Sr, commonly present in intermediate and low level liquid radioactive waste, or after exposure to very high doses of ionizing radiation. Transmission electron micrographs revealed the extracellular nature of bioprecipitated U, while X-ray diffraction and fluorescence analysis identified the precipitated uranyl phosphate species as chernikovite. When immobilized into calcium alginate beads, Deino-PhoK cells efficiently removed uranium, which remained trapped in beads, thus accomplishing physical separation of precipitated uranyl phosphate from solutions. The data demonstrate superior ability of Deino-PhoK, over earlier reported strains, in removal of uranium from alkaline solutions and its potential use in

  5. Waste treatment process for removal of contaminants from aqueous, mixed-waste solutions using sequential chemical treatment and crossflow microfiltration, followed by dewatering

    Science.gov (United States)

    Vijayan, S.; Wong, C.F.; Buckley, L.P.

    1994-11-22

    In processes of this invention aqueous waste solutions containing a variety of mixed waste contaminants are treated to remove the contaminants by a sequential addition of chemicals and adsorption/ion exchange powdered materials to remove the contaminants including lead, cadmium, uranium, cesium-137, strontium-85/90, trichloroethylene and benzene, and impurities including iron and calcium. Staged conditioning of the waste solution produces a polydisperse system of size enlarged complexes of the contaminants in three distinct configurations: water-soluble metal complexes, insoluble metal precipitation complexes, and contaminant-bearing particles of ion exchange and adsorbent materials. The volume of the waste is reduced by separation of the polydisperse system by cross-flow microfiltration, followed by low-temperature evaporation and/or filter pressing. The water produced as filtrate is discharged if it meets a specified target water quality, or else the filtrate is recycled until the target is achieved. 1 fig.

  6. 75 FR 81037 - Waste Confidence Decision Update

    Science.gov (United States)

    2010-12-23

    ... radioactive wastes produced by NPPs ``can be safely disposed of, to determine when such disposal or offsite... safe permanent disposal of high-level radioactive waste (HLW) would be available when they were needed... proceedings designed to assess the degree of assurance that radioactive wastes generated by nuclear power...

  7. 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

  8. Engineering solution for the backfilling and sealing of radioactive waste repositories

    International Nuclear Information System (INIS)

    Jorda, M.; Gouvenot, D.; Bonne, A.; Lees, T.P.; Schmidt, M.

    1990-01-01

    To ensure the safety of radioactive waste deep disposal, backfilling and sealing materials (engineered barriers) have to be used to fill residual voids. For granite medium, stress is put on emplacement techniques for cement- and clay-based materials, including in-situ validation. For clay medium, mined repository and deep boreholes drilled from the surface are considered. In the case of the first solution, the thermomechanical behaviour of a clay backfill is studied. In the same way, backfill made of excavated crushed salt is considered and thermomechanical properties evaluated by means of laboratory tests and in-situ experiments. Finally, basic works on quality assurance procedures and historic concretes behaviour are reported

  9. Cyclohexanone solvent extraction of 99TcO4 from alkaline nuclear waste solutions

    International Nuclear Information System (INIS)

    Schulz, W.W.

    1980-01-01

    Although the 99 Tc cyclohexanone solvent extraction process is still in the bench-scale development stage, the process appears well suited for engineering-scale removal of 99 Tc from alkaline Hanford waste solutions. The most pressing process development need is to resolve the phase disengaging problems encountered during water stripping operations. Stripping tests in pulse columns and/or centrifugal contactors are particularly needed to determine the magnitude of the phase disengaging problem in engineering-scale equipment and to find suitable remedies. 5 figures, 7 tables

  10. New Engineering Solutions in Creation of Mini-BOF for Metallic Waste Recycling

    Science.gov (United States)

    Eronko, S. P.; Gorbatyuk, S. M.; Oshovskaya, E. V.; Starodubtsev, B. I.

    2017-12-01

    New engineering solutions used in design of the mini melting unit capable of recycling industrial and domestic metallic waste with high content of harmful impurities are provided. High efficiency of the process technology implemented with its use is achieved due to the possibility of the heat and mass transfer intensification in the molten metal bath, controlled charge into it of large amounts of reagents in lumps and in fines, and cut-off of remaining process slag during metal tapping into the teeming ladle.

  11. Lithological suitability for HLW repository in Korea

    International Nuclear Information System (INIS)

    Kim, C.S.; Bae, D.S.; Kim, K.S.; Koh, Y.K.

    2001-01-01

    Regional geologic conditions of Korea were summarized with emphasis on rock mass and fracture system as a part of the research program for high level radioactive wastes disposal. The eastern margin of the Korea-China platform has been regarded as stable crotonic nature. The Mesozoic tectonic activities followed by igneous intrusion were the most vigorous crustal movement in the entire Korean peninsula. During the Jurassic-Cretaceous orogeny (180-130 Ma Bp), igneous activity resulted in forming a large batholith of Dab granitic rock (Jurassic granite). Rejuvenized igneous activities during the Cretaceous period formed the Bulguksa granite which are associated with felsic volcanic rocks and NE-SW/NNE-SSW geologic structures. The primary host rock is considered to be Daebo granite batholiths intruded in the geologic age of late Triassic to early Jurassic (205±15 Ma). The emplacement depths are in the range of 10-20 km and the crystallization occurs under the geopressure of 3∼7 kb. (author)

  12. Microstructural properties of high level waste concentrates and gels with raman and infrared spectroscopies. 1997 annual progress report

    International Nuclear Information System (INIS)

    Agnew, S.F.; Coarbin, R.A.; Johnston, C.T.

    1997-01-01

    'Monosodium aluminate, the phase of aluminate found in waste tanks, is only stable over a fairly narrow range of water vapor pressure (22% relative humidity at 22 C). As a result, aluminate solids are stable at Hanford (seasonal average RH ∼20%) but are not be stable at Savannah River (seasonal average RH ∼40%). Monosodium aluminate (MSA) releases water upon precipitation from solution. In contrast, trisodium aluminate (TSA) consumes water upon precipitation. As a result, MSA precipitates gradually over time while TSA undergoes rapid accelerated precipitation, often gelling its solution. Raman spectra reported for first time for monosodium and trisodium aluminate solids. Ternary phase diagrams can be useful for showing effects of water removal, even with concentrated waste. Kinetics of monosodium aluminate precipitation are extremely slow (several months) at room temperature but quite fast (several hours) at 60 C. As a result, all waste simulants that contain aluminate need several days of cooking at 60 C in order to truly represent the equilibrium state of aluminate. The high level waste (HLW) slurries that have been created at the Hanford and Savannah River Sites over that last fifty years constitute a large fraction of the remaining HLW volumes at both sites. In spite of the preponderance of these wastes, very little quantitative information is available about their physical and chemical properties other than elemental analyses.'

  13. Final waste management programmatic environmental impact statement for managing treatment, storage, and disposal of radioactive and hazardous waste. Summary

    International Nuclear Information System (INIS)

    1997-05-01

    This Waste Management Programmatic Environmental Impact Statement (WM PEIS) is a nationwide study examining the environmental impacts of managing five types of radioactive and hazardous wastes generated by past and future nuclear defense and research activities at a variety of sites located around the United States. The five waste types are low-level mixed waste (LLMW), low-level waste (LLW), transuranic waste (TRUW), high-level waste (HLW), and hazardous waste (HW)

  14. Production of furfural from waste aqueous hemicellulose solution of hardwood over ZSM-5 zeolite.

    Science.gov (United States)

    Gao, Hongling; Liu, Haitang; Pang, Bo; Yu, Guang; Du, Jian; Zhang, Yuedong; Wang, Haisong; Mu, Xindong

    2014-11-01

    This study aimed to produce furfural from waste aqueous hemicellulose solution of a hardwood kraft-based dissolving pulp production processing in a green method. The maximum furfural yield of 82.4% and the xylose conversion of 96.8% were achieved at 463K, 1.0g ZSM-5, 1.05g NaCl and organic solvent-to-aqueous phase ratio of 30:15 (V/V) for 3h. The furfural yield was just 51.5% when the same concentration of pure xylose solution was used. Under the optimized condition, furfural yield was still up to 67.1% even after the fifth reused of catalyst. Catalyst recycling study showed that ZSM-5 has a certain stability and can be efficiently reused. Copyright © 2014 Elsevier Ltd. All rights reserved.

  15. Fluorescent Lamp Glass Waste Incorporation into Clay Ceramic: A Perfect Solution

    Science.gov (United States)

    Morais, Alline Sardinha Cordeiro; Vieira, Carlos Maurício Fontes; Rodriguez, Rubén Jesus Sanchez; Monteiro, Sergio Neves; Candido, Veronica Scarpini; Ferreira, Carlos Luiz

    2016-09-01

    The mandatory use of fluorescent lamps as part of a Brazilian energy-saving program generates a huge number of spent fluorescent lamps (SFLs). After operational life, SFLs cannot be disposed as common garbage owing to mercury and lead contamination. Recycling methods separate contaminated glass tubes and promote cleaning for reuse. In this work, glass from decontaminated SFLs was incorporated into clay ceramics, not only as an environmental solution for such glass wastes and clay mining reduction but also due to technical and economical advantages. Up to 30 wt.% of incorporation, a significant improvement in fired ceramic flexural strength and a decrease in water absorption was observed. A prospective analysis showed clay ceramic incorporation as an environmentally correct and technical alternative for recycling the enormous amount of SFLs disposed of in Brazil. This could also be a solution for other world clay ceramic producers, such as US, China and some European countries.

  16. Predisposal management of high level radioactive waste. Safety guide

    International Nuclear Information System (INIS)

    2005-01-01

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

  17. Investigations regarding the wet decontamination of fluorescent lamp waste using iodine in potassium iodide solutions

    International Nuclear Information System (INIS)

    Tunsu, Cristian; Ekberg, Christian; Foreman, Mark; Retegan, Teodora

    2015-01-01

    Highlights: • A wet-based decontamination process for fluorescent lamp waste is proposed. • Mercury can be leached using iodine in potassium iodide solution. • The efficiency of the process increases with an increase in leachant concentration. • Selective leaching of mercury from rare earth elements is achieved. • Mercury is furthered recovered using ion exchange, reduction or solvent extraction. - Abstract: With the rising popularity of fluorescent lighting, simple and efficient methods for the decontamination of discarded lamps are needed. Due to their mercury content end-of-life fluorescent lamps are classified as hazardous waste, requiring special treatment for disposal. A simple wet-based decontamination process is required, especially for streams where thermal desorption, a commonly used but energy demanding method, cannot be applied. In this study the potential of a wet-based process using iodine in potassium iodide solution was studied for the recovery of mercury from fluorescent lamp waste. The influence of the leaching agent’s concentration and solid/liquid ratio on the decontamination efficiency was investigated. The leaching behaviour of mercury was studied over time, as well as its recovery from the obtained leachates by means of anion exchange, reduction, and solvent extraction. Dissolution of more than 90% of the contained mercury was achieved using 0.025/0.05 M I 2 /KI solution at 21 °C for two hours. The efficiency of the process increased with an increase in leachant concentration. 97.3 ± 0.6% of the mercury contained was dissolved at 21 °C, in two hours, using a 0.25/0.5 M I 2 /KI solution and a solid to liquid ratio of 10% w/v. Iodine and mercury can be efficiently removed from the leachates using Dowex 1X8 anion exchange resin or reducing agents such as sodium hydrosulphite, allowing the disposal of the obtained solution as non-hazardous industrial wastewater. The extractant CyMe 4 BTBP showed good removal of mercury, with an

  18. Investigations regarding the wet decontamination of fluorescent lamp waste using iodine in potassium iodide solutions

    Energy Technology Data Exchange (ETDEWEB)

    Tunsu, Cristian, E-mail: tunsu@chalmers.se; Ekberg, Christian; Foreman, Mark; Retegan, Teodora

    2015-02-15

    Highlights: • A wet-based decontamination process for fluorescent lamp waste is proposed. • Mercury can be leached using iodine in potassium iodide solution. • The efficiency of the process increases with an increase in leachant concentration. • Selective leaching of mercury from rare earth elements is achieved. • Mercury is furthered recovered using ion exchange, reduction or solvent extraction. - Abstract: With the rising popularity of fluorescent lighting, simple and efficient methods for the decontamination of discarded lamps are needed. Due to their mercury content end-of-life fluorescent lamps are classified as hazardous waste, requiring special treatment for disposal. A simple wet-based decontamination process is required, especially for streams where thermal desorption, a commonly used but energy demanding method, cannot be applied. In this study the potential of a wet-based process using iodine in potassium iodide solution was studied for the recovery of mercury from fluorescent lamp waste. The influence of the leaching agent’s concentration and solid/liquid ratio on the decontamination efficiency was investigated. The leaching behaviour of mercury was studied over time, as well as its recovery from the obtained leachates by means of anion exchange, reduction, and solvent extraction. Dissolution of more than 90% of the contained mercury was achieved using 0.025/0.05 M I{sub 2}/KI solution at 21 °C for two hours. The efficiency of the process increased with an increase in leachant concentration. 97.3 ± 0.6% of the mercury contained was dissolved at 21 °C, in two hours, using a 0.25/0.5 M I{sub 2}/KI solution and a solid to liquid ratio of 10% w/v. Iodine and mercury can be efficiently removed from the leachates using Dowex 1X8 anion exchange resin or reducing agents such as sodium hydrosulphite, allowing the disposal of the obtained solution as non-hazardous industrial wastewater. The extractant CyMe{sub 4}BTBP showed good removal of mercury

  19. Investigations regarding the wet decontamination of fluorescent lamp waste using iodine in potassium iodide solutions.

    Science.gov (United States)

    Tunsu, Cristian; Ekberg, Christian; Foreman, Mark; Retegan, Teodora

    2015-02-01

    With the rising popularity of fluorescent lighting, simple and efficient methods for the decontamination of discarded lamps are needed. Due to their mercury content end-of-life fluorescent lamps are classified as hazardous waste, requiring special treatment for disposal. A simple wet-based decontamination process is required, especially for streams where thermal desorption, a commonly used but energy demanding method, cannot be applied. In this study the potential of a wet-based process using iodine in potassium iodide solution was studied for the recovery of mercury from fluorescent lamp waste. The influence of the leaching agent's concentration and solid/liquid ratio on the decontamination efficiency was investigated. The leaching behaviour of mercury was studied over time, as well as its recovery from the obtained leachates by means of anion exchange, reduction, and solvent extraction. Dissolution of more than 90% of the contained mercury was achieved using 0.025/0.05 M I2/KI solution at 21 °C for two hours. The efficiency of the process increased with an increase in leachant concentration. 97.3 ± 0.6% of the mercury contained was dissolved at 21 °C, in two hours, using a 0.25/0.5M I2/KI solution and a solid to liquid ratio of 10% w/v. Iodine and mercury can be efficiently removed from the leachates using Dowex 1X8 anion exchange resin or reducing agents such as sodium hydrosulphite, allowing the disposal of the obtained solution as non-hazardous industrial wastewater. The extractant CyMe4BTBP showed good removal of mercury, with an extraction efficiency of 97.5 ± 0.7% being achieved in a single stage. Better removal of mercury was achieved in a single stage using the extractants Cyanex 302 and Cyanex 923 in kerosene, respectively. Copyright © 2014 Elsevier Ltd. All rights reserved.

  20. NEXT GENERATION MELTER(S) FOR VITRIFICATION OF HANFORD WASTE: STATUS AND DIRECTION

    International Nuclear Information System (INIS)

    Ramsey, W.G.; Gray, M.F.; Calmus, R.B.; Edge, J.A.; Garrett, B.G.

    2011-01-01

    Vitrification technology has been selected to treat high-level waste (HLW) at the Hanford Site, the West Valley Demonstration Project and the Savannah River Site (SRS), and low activity waste (LAW) at Hanford. In addition, it may potentially be applied to other defense waste streams such as sodium bearing tank waste or calcine. Joule-heated melters (already in service at SRS) will initially be used at the Hanford Site's Waste Treatment and Immobilization Plant (WTP) to vitrify tank waste fractions. The glass waste content and melt/production rates at WTP are limited by the current melter technology. Significant reductions in glass volumes and mission life are only possible with advancements in melter technology coupled with new glass formulations. The Next Generation Melter (NGM) program has been established by the U.S. Department of Energy's (DOE's), Environmental Management Office of Waste Processing (EM-31) to develop melters with greater production capacity (absolute glass throughput rate) and the ability to process melts with higher waste fractions. Advanced systems based on Joule-Heated Ceramic Melter (JHCM) and Cold Crucible Induction Melter (CCIM) technologies will be evaluated for HLW and LAW processing. Washington River Protection Solutions (WRPS), DOE's tank waste contractor, is developing and evaluating these systems in cooperation with EM-31, national and university laboratories, and corporate partners. A primary NGM program goal is to develop the systems (and associated flowsheets) to Technology Readiness Level 6 by 2016. Design and testing are being performed to optimize waste glass process envelopes with melter and balance of plant requirements. A structured decision analysis program will be utilized to assess the performance of the competing melter technologies. Criteria selected for the decision analysis program will include physical process operations, melter performance, system compatibility and other parameters.

  1. Optimizing High Level Waste Disposal

    International Nuclear Information System (INIS)

    Dirk Gombert

    2005-01-01

    If society is ever to reap the potential benefits of nuclear energy, technologists must close the fuel-cycle completely. A closed cycle equates to a continued supply of fuel and safe reactors, but also reliable and comprehensive closure of waste issues. High level waste (HLW) disposal in borosilicate glass (BSG) is based on 1970s era evaluations. This host matrix is very adaptable to sequestering a wide variety of radionuclides found in raffinates from spent fuel reprocessing. However, it is now known that the current system is far from optimal for disposal of the diverse HLW streams, and proven alternatives are available to reduce costs by billions of dollars. The basis for HLW disposal should be reassessed to consider extensive waste form and process technology research and development efforts, which have been conducted by the United States Department of Energy (USDOE), international agencies and the private sector. Matching the waste form to the waste chemistry and using currently available technology could increase the waste content in waste forms to 50% or more and double processing rates. Optimization of the HLW disposal system would accelerate HLW disposition and increase repository capacity. This does not necessarily require developing new waste forms, the emphasis should be on qualifying existing matrices to demonstrate protection equal to or better than the baseline glass performance. Also, this proposed effort does not necessarily require developing new technology concepts. The emphasis is on demonstrating existing technology that is clearly better (reliability, productivity, cost) than current technology, and justifying its use in future facilities or retrofitted facilities. Higher waste processing and disposal efficiency can be realized by performing the engineering analyses and trade-studies necessary to select the most efficient methods for processing the full spectrum of wastes across the nuclear complex. This paper will describe technologies being

  2. Waste-Mixes Study for space disposal

    International Nuclear Information System (INIS)

    McCallum, R.F.; Blair, H.T.; McKee, R.W.; Silviera, D.J.; Swanson, J.L.

    1983-01-01

    The Wastes Mixes Study is a component of Cy-1981 and 1982 research activities to determine if space disposal could be a feasible complement to geologic disposal for certain high-level (HLW) and transuranic wastes (TRU). The objectives of the study are: to determine if removal of radionuclides from HLW and TRU significantly reduces the long-term radiological risks of geologic disposal; to determine if chemical partitioning of the waste for space disposal is technically feasible; to identify acceptable waste forms for space disposal; and to compare improvements in geologic disposal system performance to impacts of additional treatment, storage, and transportation necessary for space disposal. To compare radiological effects, five system alternatives are defined: Reference case - All HLW and TRU to a repository. Alternative A - Iodine to space, the balance to a repository. Alternative B - Technetium to space, the balance to a repository. Alternative C - 95% of cesium and strontium to a repository; the balance of HLW aged first, then to space; plutonium separated from TRU for recycle; the balance of the TRU to a repository. Alternative D - HLW aged first, then to space, plutonium separated from TRU for recycle; the balance of the TRU to a repository. The conclusions of this study are: the incentive for space disposal is that it offers a perception of reduced risks rather than significant reduction. Suitable waste forms for space disposal are cermet for HLW, metallic technetium, and lead iodide. Space disposal of HLW appears to offer insignificant safety enhancements when compared to geologic disposal; the disposal of iodine and technetium wastes in space does not offer risk advantages. Increases in short-term doses for the alternatives are minimal; however, incremental costs of treating, storing and transporting wastes for space disposal are substantial

  3. Removal of cesium and strontium from low active waste solutions by zeolites

    International Nuclear Information System (INIS)

    Jain, Savita; Ramaswamy, M.; Theyyunni, T.K.

    1994-01-01

    Ion exchange, crystallographic and thermal characteristics of sodium, cesium and strontium forms of locally available synthetic zeolites have been investigated. X-ray and differential thermal analyses have confirmed that the synthetic materials AR1 and 4A belonged to the mordenite and A type families of zeolites respectively. Equilibrium uptake of cesium and strontium ions by sodium forms of zeolite was studied as a function of time, pH and sodium concentration. It was found that the rate of sorption by AR1 was higher than that by 4A. In regard to pH, distribution of nuclides on zeolites was found to pass through maxima at a pH value of around 9. Sodium ion interfered with the sorption of cesium and strontium by zeolites. However, at sodium concentration ≤ 0.01 M, distribution coefficient values for these nuclides were sufficiently high to merit consideration of these zeolites for low level waste treatment. Lab-scale column runs using 5 ml beds of materials showed that the zeolites AR1 and 4A were very effective in removing cesium and strontium nuclides respectively from large volumes (a decontamination factor of 50 for a throughput of 6000 bed volumes) of actual low level waste solutions. Thus, the zeolite system has a potential future for large scale application in the treatment of low level wastes. (author). 6 refs., 5 figs., 6 tabs

  4. Uranium extraction from aqueous solution using dried and pyrolyzed tea and coffee wastes

    International Nuclear Information System (INIS)

    Zaynab Aly

    2013-01-01

    The adsorption of U(VI) onto dried and pyrolyzed tea and coffee wastes was investigated. The adsorption properties of the materials were characterized by measuring uranium uptake as a function of solution pH, kinetics and adsorption isotherms. pH profile of uranium adsorption where UO 2 2+ is expected to be the predominant species was measured between pH 0 and 4. Both Langmuir and Freundlich adsorption models were used to describe adsorption equilibria, and corresponding constants evaluated. Using the Langmuir model, the maximum adsorption capacity of uranium by dried tea and coffee wastes was 59.5 and 34.8 mg/g, respectively at 291 K. Adsorption thermodynamic constants, ΔHdeg ΔSdeg and ΔGdeg were also calculated from adsorption data obtained at three different temperatures. Adsorption thermodynamics of uranyl ions on dried tea and coffee systems indicated spontaneous and endothermic processes. Additionally, a Lagergren pseudo-second-order kinetic model was used to fit the kinetic experimental data for both adsorbents and the constants evaluated. Dried tea and coffee wastes proved to be effective adsorbents with high capacities and significant advantage of a very low cost. (author)

  5. Reuse of waste beer yeast sludge for biosorptive decolorization of reactive blue 49 from aqueous solution.

    Science.gov (United States)

    Wang, Baoe; Guo, Xiu

    2011-06-01

    Reactive blue 49 was removed from aqueous solution by biosorption using powder waste sludge composed of Saccharomyces cerevisiae from the beer-brewing industry. The effect of initial pH, temperature and the biosorption thermodynamics, equilibrium, kinetics was investigated in this study. It was found that the biosorption capacity was at maximum at initial pH 3, that the effect of temperature on biosorption of reactive blue 49 was only slight in relation to the large biosorption capacity (25°C, 361 mg g(-1)) according as the biosorption capacity decreased only 43 mg g(-1) at the temperature increased from 25 to 50°C. The biosorption was spontaneous, exothermic in nature and the dye molecules movements decreased slightly in random at the solid/liquid interface during the biosorption of dye on biosorbents. The biosorption equilibrium data could be described by Freundich isotherm model. The biosorption rates were found to be consistent with a pseudo-second-order kinetics model. The functional group interaction analysis between waste beer yeast sludge and reactive blue 49 by the aid of Fourier transform infrared (abbr. FTIR) spectroscopy indicated that amino components involved in protein participated in the biosorption process, which may be achieved by the mutual electrostatic adsorption process between the positively charged amino groups in waste beer yeast sludge with negatively charged sulfonic groups in reactive blue 49.

  6. DEMONSTRATION OF THE NEXT-GENERATION CAUSTIC-SIDE SOLVENT EXTRACTION SOLVENT WITH 2-CM CENTRIFUGAL CONTRACTORS USING TANK 49H WASTE AND WASTE SIMULANT

    Energy Technology Data Exchange (ETDEWEB)

    Pierce, R.; Peters, T.; Crowder, M.; Caldwell, T.; Pak, D; Fink, S.; Blessing, R.; Washington, A.

    2011-09-27

    Researchers successfully demonstrated the chemistry and process equipment of the Caustic-Side Solvent Extraction (CSSX) flowsheet using MaxCalix for the decontamination of high level waste (HLW). The demonstration was completed using a 12-stage, 2-cm centrifugal contactor apparatus at the Savannah River National Laboratory (SRNL). This represents the first CSSX process demonstration of the MaxCalix solvent system with Savannah River Site (SRS) HLW. Two tests lasting 24 and 27 hours processed non-radioactive simulated Tank 49H waste and actual Tank 49H HLW, respectively. Conclusions from this work include the following. The CSSX process is capable of reducing {sup 137}Cs in high level radioactive waste by a factor of more than 40,000 using five extraction, two scrub, and five strip stages. Tests demonstrated extraction and strip section stage efficiencies of greater than 93% for the Tank 49H waste test and greater than 88% for the simulant waste test. During a test with HLW, researchers processed 39 liters of Tank 49H solution and the waste raffinate had an average decontamination factor (DF) of 6.78E+04, with a maximum of 1.08E+05. A simulant waste solution ({approx}34.5 liters) with an initial Cs concentration of 83.1 mg/L was processed and had an average DF greater than 5.9E+03, with a maximum DF of greater than 6.6E+03. The difference may be attributable to differences in contactor stage efficiencies. Test results showed the solvent can be stripped of cesium and recycled for {approx}25 solvent turnovers without the occurrence of any measurable solvent degradation or negative effects from minor components. Based on the performance of the 12-stage 2-cm apparatus with the Tank 49H HLW, the projected DF for MCU with seven extraction, two scrub, and seven strip stages operating at a nominal efficiency of 90% is {approx}388,000. At 95% stage efficiency, the DF in MCU would be {approx}3.2 million. Carryover of organic solvent in aqueous streams (and aqueous in organic

  7. High-level waste melter alternatives assessment report

    Energy Technology Data Exchange (ETDEWEB)

    Calmus, R.B.

    1995-02-01

    This document describes the Tank Waste Remediation System (TWRS) High-Level Waste (HLW) Program`s (hereafter referred to as HLW Program) Melter Candidate Assessment Activity performed in fiscal year (FY) 1994. The mission of the TWRS Program is to store, treat, and immobilize highly radioactive Hanford Site waste (current and future tank waste and encapsulated strontium and cesium isotopic sources) in an environmentally sound, safe, and cost-effective manner. The goal of the HLW Program is to immobilize the HLW fraction of pretreated tank waste into a vitrified product suitable for interim onsite storage and eventual offsite disposal at a geologic repository. Preparation of the encapsulated strontium and cesium isotopic sources for final disposal is also included in the HLW Program. As a result of trade studies performed in 1992 and 1993, processes planned for pretreatment of tank wastes were modified substantially because of increasing estimates of the quantity of high-level and transuranic tank waste remaining after pretreatment. This resulted in substantial increases in needed vitrification plant capacity compared to the capacity of original Hanford Waste Vitrification Plant (HWVP). The required capacity has not been finalized, but is expected to be four to eight times that of the HWVP design. The increased capacity requirements for the HLW vitrification plant`s melter prompted the assessment of candidate high-capacity HLW melter technologies to determine the most viable candidates and the required development and testing (D and T) focus required to select the Hanford Site HLW vitrification plant melter system. An assessment process was developed in early 1994. This document describes the assessment team, roles of team members, the phased assessment process and results, resulting recommendations, and the implementation strategy.

  8. High-level waste melter alternatives assessment report

    International Nuclear Information System (INIS)

    Calmus, R.B.

    1995-02-01

    This document describes the Tank Waste Remediation System (TWRS) High-Level Waste (HLW) Program's (hereafter referred to as HLW Program) Melter Candidate Assessment Activity performed in fiscal year (FY) 1994. The mission of the TWRS Program is to store, treat, and immobilize highly radioactive Hanford Site waste (current and future tank waste and encapsulated strontium and cesium isotopic sources) in an environmentally sound, safe, and cost-effective manner. The goal of the HLW Program is to immobilize the HLW fraction of pretreated tank waste into a vitrified product suitable for interim onsite storage and eventual offsite disposal at a geologic repository. Preparation of the encapsulated strontium and cesium isotopic sources for final disposal is also included in the HLW Program. As a result of trade studies performed in 1992 and 1993, processes planned for pretreatment of tank wastes were modified substantially because of increasing estimates of the quantity of high-level and transuranic tank waste remaining after pretreatment. This resulted in substantial increases in needed vitrification plant capacity compared to the capacity of original Hanford Waste Vitrification Plant (HWVP). The required capacity has not been finalized, but is expected to be four to eight times that of the HWVP design. The increased capacity requirements for the HLW vitrification plant's melter prompted the assessment of candidate high-capacity HLW melter technologies to determine the most viable candidates and the required development and testing (D and T) focus required to select the Hanford Site HLW vitrification plant melter system. An assessment process was developed in early 1994. This document describes the assessment team, roles of team members, the phased assessment process and results, resulting recommendations, and the implementation strategy

  9. A Study of the applicability of biomineralization process to the treatment of radioactive waste solutions

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chang Heon; Choi, Kwang Soon; Kim, Yong Bok

    2004-01-01

    A study has been carried out on the removal of uranium and radionuclides like {sup 137}Cs, {sup 60}Co and {sup 90}Sr from radioactive waste solutions using a bioreactor of Serratia sp.-biofilm which is capable of producing an acid type phosphatase enzyme that liberates HPO{sub 4}{sup 2-} from a suitable organic phosphate donor and of stoichiometrically forming precipitates of the metal ions (M{sup 2+}) such as insoluble MHPO{sub 4}(HUP) at the cell surface. Sorption behaviours of the Serratia sp.-biofilm for UO{sub 2}{sup 2+}, Cs{sup +}, Sr{sup 2+} and Co{sup 2+} were investigated using synthetic radioactive waste solutions containing {sup 137}Cs, {sup 60}Co and {sup 85}Sr. Uranium was effectively removed as hydrogen uranyl phosphate (HUP) in the presence of glycerol-2-phosphate. Cs{sup +}, Sr{sup 2+} and Co{sup 2+} were removed via an intercalation by either continuous co-crystal(precipitate) growth along with U or by a pre-formed HUP host crystal as an inorganic ion exchanger. Effect of sodium ion on the sorption behaviour of serratia sp.-HUP-bioreactor and its stability were investigated. In addition, Sorption behaviours of the serratia sp.-HUP-bioreactor were compared with those of Amberlite IRN 77,cation exchanger and ammoniumphosphotungstate/silical gel which shows selective sorption property for Cs{sup +}.

  10. A Study of the applicability of biomineralization process to the treatment of radioactive waste solutions

    International Nuclear Information System (INIS)

    Lee, Chang Heon; Choi, Kwang Soon; Kim, Yong Bok

    2004-01-01

    A study has been carried out on the removal of uranium and radionuclides like 137 Cs, 60 Co and 90 Sr from radioactive waste solutions using a bioreactor of Serratia sp.-biofilm which is capable of producing an acid type phosphatase enzyme that liberates HPO 4 2- from a suitable organic phosphate donor and of stoichiometrically forming precipitates of the metal ions (M 2+ ) such as insoluble MHPO 4 (HUP) at the cell surface. Sorption behaviours of the Serratia sp.-biofilm for UO 2 2+ , Cs + , Sr 2+ and Co 2+ were investigated using synthetic radioactive waste solutions containing 137 Cs, 60 Co and 85 Sr. Uranium was effectively removed as hydrogen uranyl phosphate (HUP) in the presence of glycerol-2-phosphate. Cs + , Sr 2+ and Co 2+ were removed via an intercalation by either continuous co-crystal(precipitate) growth along with U or by a pre-formed HUP host crystal as an inorganic ion exchanger. Effect of sodium ion on the sorption behaviour of serratia sp.-HUP-bioreactor and its stability were investigated. In addition, Sorption behaviours of the serratia sp.-HUP-bioreactor were compared with those of Amberlite IRN 77,cation exchanger and ammoniumphosphotungstate/silical gel which shows selective sorption property for Cs +

  11. Licensing information needs for a high-level waste repository

    International Nuclear Information System (INIS)

    Wright, R.J.; Greeves, J.T.; Logsdon, M.J.

    1985-01-01

    The information needs for licensing findings during the development of a repository for high-level waste (HLW) are described. In particular, attention is given to the information and needs to demonstrate, for construction authorization purposes: repository constructibility, waste retrievability, waste containment, and waste isolation

  12. Synroc - a multiphase ceramic for high level nuclear waste immobilisation

    International Nuclear Information System (INIS)

    Reeve, K.D.; Vance, E.R.; Hart, K.P.; Smith, K.L.; Lumpkin, G.R.; Mercer, D.J.

    1992-01-01

    Many natural minerals - particularly titanates - are very durable geochemically, having survived for millions of years with very little alteration. Moreover, some of these minerals have quantitatively retained radioactive elements and their daughter products over this time. The Synroc concept mimics nature by providing an all-titanate synthetic mineral phase assemblage to immobilise high level waste (HLW) from nuclear fuel reprocessing operations for safe geological disposal. In principle, many chemically hazardous inorganic wastes arising from industry could also be immobilised in highly durable ceramics and disposed of geologically, but in practice the cost structure of most industries is such that lower cost waste management solutions - for example, the development of reusable by-products or the use of cements rather than ceramics - have to be devised. In many thousands of aqueous leach tests at ANSTO, mostly at 70-90 deg C, Synroc has been shown to be exceptionally durable. The emphases of the recent ANSTO program have been on tailoring of the Synroc composition to varying HLW compositions, leach testing of Synroc containing radioactive transuranic actinides, study of leaching mechanisms by SEM and TEM, and the development and costing of a conceptual fully active Synroc fabrication plant design. A summary of recent results on these topics will be presented. 29 refs., 4 figs

  13. Inorganic sorbents for radiostrontium removal from waste solutions: selectivity and role of calixarenes

    International Nuclear Information System (INIS)

    Vijayan, S.; Belikov, K.; Drapailo, A.

    2011-01-01

    The challenge in the remediation of 90 Sr-contaminated waters arises from the need to achieve very high removal efficiencies to meet discharge targets from waste effluents containing relatively high concentrations of non-radioactive cations. Low-cost natural zeolites are not selective for strontium over other divalent cations, notably such ions as calcium; and produce low 90 Sr removal performance, and large volumes of spent sorbent waste. The synthesis and use of selective, synthetic inorganic sorbents could prove to be a feasible approach for high 90 Sr removal efficiencies, and much smaller volumes of secondary solid waste generation. The essential advantages of inorganic sorbents include their stability and resistance to radiation, and the potential for producing stable waste forms such as vitrified glass or ceramics for disposal. However, the cost of strontium-specific sorbents is prohibitive for large-scale applications at present. This paper is a review of the reported information on removal mechanisms and performance of Sr-specific inorganic sorbents. The analysis has revealed promising performance, efficiency and selectivity for strontium removal from solutions containing low and high concentrations of salts. The leading sorbents are crystalline silicotitanate and oxides of metals such as titanium. An initial assessment has also been made of the performance of calixarene-based macrocyclic compounds. These are known for their selectivity for strontium in solvent extraction processes. From the initial strontium removal results in bench-scale tests using different solid substrates, impregnated with calixarene derivatives, only sodium-mordenite impregnated with calyx[8]arene octamide gave an overall strontium removal efficiency in the range of 90 to 95% in the presence of 3.5 ppm calcium. There was no improvement observed for strontium-removal efficiency or selectivity over calcium in the calixarene-impregnated inorganic sorbent matrix. In several tests, the

  14. Development of knowledge building program concerning about high-level radioactive waste disposal

    International Nuclear Information System (INIS)

    Kimura, Hiroshi; Yamada, Kazuhiro; Takase, Hiroyasu

    2005-01-01

    Acquirement of knowledge about the high-level radioactive waste (HLW) disposal is one of the important factors for public to determine the social acceptance of HLW disposal. However in Japan, public do not have knowledge about HLW and its disposal sufficiently. In this work, we developed the knowledge building program concerning about HLW disposal based on Nonaka, and Takeuchi's SECI spiral model in knowledge management, and carried to the experiment on this program. In the results, we found that the participants' knowledge about the HLW disposal increased and changed from misunderstanding' or 'assuming' to 'facts' or 'consideration' through this experimental program. These results said that the experimental program leads participants to have higher quality of knowledge about the HLW disposal. In consequence, this knowledge building program may be effective in the acquirement of high quality knowledge. (author)

  15. Numerical investigation of high level nuclear waste disposal in deep anisotropic geologic repositories

    KAUST Repository

    Salama, Amgad; El Amin, Mohamed F.; Sun, Shuyu

    2015-01-01

    One of the techniques that have been proposed to dispose high level nuclear waste (HLW) has been to bury them in deep geologic formations, which offer relatively enough space to accommodate the large volume of HLW accumulated over the years since

  16. Conceptual design report for immobilized high-level waste interim storage facility (Phase 1)

    International Nuclear Information System (INIS)

    Burgard, K.C.

    1998-01-01

    The Hanford Site Canister Storage Building (CSB Bldg. 212H) will be utilized to interim store Phase 1 HLW products. Project W-464, Immobilized High-Level Waste Interim Storage, will procure an onsite transportation system and retrofit the CSB to accommodate the Phase 1 HLW products. The Conceptual Design Report establishes the Project W-464 technical and cost basis

  17. Conceptual design report for immobilized high-level waste interim storage facility (Phase 1)

    Energy Technology Data Exchange (ETDEWEB)

    Burgard, K.C.

    1998-04-09

    The Hanford Site Canister Storage Building (CSB Bldg. 212H) will be utilized to interim store Phase 1 HLW products. Project W-464, Immobilized High-Level Waste Interim Storage, will procure an onsite transportation system and retrofit the CSB to accommodate the Phase 1 HLW products. The Conceptual Design Report establishes the Project W-464 technical and cost basis.

  18. The comparison of DYNA3D to approximate solutions for a partially- full waste storage tank subjected to seismic loading

    International Nuclear Information System (INIS)

    Zaslawsky, M.; Kennedy, W.N.

    1992-01-01

    Mathematical solutions to the problem consisting of a partially-full waste tank subjected to seismic loading, embedded in soil, is classically difficult in that one has to address: soil-structure interaction, fluid-structure interaction, non-linear behavior of material, dynamic effects. Separating the problem and applying numerous assumptions will yield approximate solutions. This paper explores methods for generating these solutions accurately

  19. Specifications of the International Atomic Energy Agency's international project on safety assessment driven radioactive waste management solutions

    International Nuclear Information System (INIS)

    Ghannadi, M.; Asgharizadeh, F.; Assadi, M. R.

    2008-01-01

    Radioactive waste is produced in the generation of nuclear power and the production and use of radioactive materials in the industry, research, and medicine. The nuclear waste management facilities need to perform a safety assessment in order to ensure the safety of a facility. Nuclear safety assessment is a structured and systematic way of examining a proposed facility, process, operation and activity. In nuclear waste management point of view, safety assessment is a process which is used to evaluate the safety of radioactive waste management and disposal facilities. In this regard the International Atomic Energy Agency is planed to implement an international project with cooperation of some member states. The Safety Assessment Driving Radioactive Waste Management Solutions Project is an international programme of work to examine international approaches to safety assessment in aspects of p redisposal r adioactive waste management, including waste conditioning and storage. This study is described the rationale, common aspects, scope, objectives, work plan and anticipated outcomes of the project with refer to International Atomic Energy Agency's documents, such as International Atomic Energy Agency's Safety Standards, as well as the Safety Assessment Driving Radioactive Waste Management Solutions project reports

  20. 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)

  1. 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

  2. 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

  3. Equilibrium leach tests with cobalt in the system cemented waste form/container material/aqueous solution

    International Nuclear Information System (INIS)

    Vejmelka, P.; Koester, R.; Lee, M. J.; Han, K. W.

    1991-01-01

    The equilibrium concentrations of Co in the system of cemented waste form/aqueous solutions were determined including the effect of the container material and its corrosion products under the respective conditions. The chemical conditions in the near field of the waste form were characterized by measurement of the pH and E h value. As disposal relevant solutions, saturated sodium chloride, Q-brine (main constituent MgCl 2 ) and a granitic type groundwater were used. For comparison, also experiments using deionized water were performed. In all systems investigated the cemented waste form itself has a strong influence on the chemical conditions in the near field. The pH and E h values are affected in all cases by the addition of the cemented waste form. There is no or only a slight difference between the E h values if iron powder or iron hydroxide is added to the cemented waste form/solution systems, but the E h is markedly decreased when iron powder is added to the solution free of cement. The Co concentration is decreased in all solutions by the addition of the cemented waste form, the largest effect is observed in Q-brine and this can be attributed either to the sorption of the Co-ions on the corrosion products of the cement or to the coprecipitation of Co-hydroxide and Mg-hydroxide. In the other solutions the Co concentration is decreased by precipitation of Co-hydroxide due to the high pH value of 12.5, and the concentrations are comparable for the different solutions

  4. Waste acceptance product specifications for vitrified high-level waste forms

    International Nuclear Information System (INIS)

    Applewhite-Ramsey, A.; Sproull, J.F.

    1993-01-01

    The Nuclear Waste Policy Act of 1982 mandated that all high-level waste (HLW) be sent to a federal geologic repository for permanent disposal. DOE published the Environmental Assessment in 1982 which identified borosilicate glass as the chosen HLW form. 1 In 1985 the Department of Energy instituted a Waste Acceptance Process to assure that DWPF glass waste forms would be acceptable to such a repository. This assurance was important since production of waste forms will precede repository construction and licensing. As part of this Waste Acceptance Process, the DOE Office of Civilian Radioactive Waste Management (RW) formed the Waste Acceptance Committee (WAC). The WAC included representatives from the candidate repository sites, the waste producing sites and DOE. The WAC was responsible for developing the Waste Acceptance Preliminary Specifications (WAPS) which defined the requirements the waste forms must meet to be compatible with the candidate repository geologies

  5. Cesium Removal from Savannah River Site Radioactive Waste Using the Caustic Side Solvent Extraction (CSSX) Process

    International Nuclear Information System (INIS)

    WALKER, DARREL

    2004-01-01

    Researchers at the Savannah River Technology Center (SRTC) successfully demonstrated the Caustic-Side Solvent Extraction (CSSX) process flow sheet using a 33-stage, 2-cm centrifugal contactor apparatus in two 24-hour tests using actual high level waste. Previously, we demonstrated the solvent extraction process with actual SRS HLW supernatant solution using a non-optimized solvent formulation. Following that test, the solvent system was optimized to enhance extractant solubility in the diluent by increasing the modifier concentration. We now report results of two tests with the new and optimized solvent

  6. Adsorption of lanthanides in aqueous solution aiming to study of nuclear wastes

    International Nuclear Information System (INIS)

    Belline, Jean de Brito

    2009-01-01

    The problem of radioactive wastes is a concern of world-wide scope, a time that does not still have a defined local for the construction of a repository for radioactive wastes of high level. One of the preliminary stages for the choice of the place more appropriate is the geologic study associated to the experimental studies of adsorption of the involved chemical species in the process. In this work, a sample of basaltic rock was used, of the South Region of the Formation Serra Geral, collected in Frederico Westphalen Town (RS), that it will be probably a candidate to the rock hostess for location of radioactive wastes. Two experiments have been carried out through, namely: 'Test Batch' and Percolating, both under atmospheric pressure, at the ambient temperature of 25 deg C, with the purpose to study the capacity of sorption of the rare earth elements - REE. The REE are used in this work in function of its analogy with the actinides, aiming at to investigate the chemistry behavior and the speciation of the same in natural waters, searching the possibility of geologic storage of radioactive wastes, a time that the adsorption of the REE depends on variables of the environment as pH, ionic strength, temperature and presence of ligands, as carbonates and constituent of surfaces of minerals. Experiment of percolating of the REE was carried through, 100ppb, in the basalt (with 80 mesh) in solutions with ionic strength 1= 0,025 M and 1=0,5 M of NaCl. pH was controlled in a range of 5,6 the 7,6 with HNO 3 addition. The concentrations were analyzed by ICP-MS. The 'Batch Test' is an efficient form of studying sorption/desorption isotherms, beyond values of the reason between the distributions solid/solution and estimation of the solubility. The percolating experiment, was carried through under pH controlled around 6, and allowed to verify the behaviour of heavy REE in comparison with the light REE. (author)

  7. Adsorption behavior and mechanism of Cr(VI) using Sakura waste from aqueous solution

    International Nuclear Information System (INIS)

    Qi, Wenfang; Zhao, Yingxin; Zheng, Xinyi; Ji, Min; Zhang, Zhenya

    2016-01-01

    Graphical abstract: The main chemical components of Sakura leaves are cellulose 16.6%, hemicellulose 10.4%, lignin 18.3%, ash 11.4%, and others 43.3%. The adsorption capacity of Cr(VI) onto Sakura leaves can achieve 435.25 mg g"−"1, much higher than other similar agroforestry wastes. - Highlights: • Sakura leaves were prepared to remove Cr(VI) from aqueous solution. • The maximum adsorption capacity of Cr(VI) reached 435.25 mg g"−"1. • Cr(VI) adsorption fitted pseudo-second-order kinetic model. • Isotherm models indicated Cr(VI) adsorption occurred on a monolayer surface. • The influence order of coexisting ions followed PO_4"3"− > SO_4"2"− > Cl"−. - Abstract: A forestall waste, Sakura leave, has been studied for the adsorption of Cr(VI) from aqueous solution. The materials before and after adsorption were characterized by X-ray photoelectron spectroscopy (XPS) and Fourier transform infrared spectroscopy (FTIR). To investigate the adsorption performance of Sakura waste, batch experiments were conducted under different adsorbent dosage, contact time, initial concentration of Cr(VI), and co-existing ions. Results showed the data fitted pseudo-second-order better than pseudo-first-order kinetic model. Equilibrium data was analyzed with Langmuir, Freundlich and Redlich–Peterson isotherm models at temperature ranges from 25 °C to 45 °C. The maximum adsorption capacity from the Langmuir model was 435.25 mg g"−"1 at pH 1.0. The presence of Cl"−, SO_4"2"− and PO_4"3"− would lead to an obvious negative effect on Cr(VI) adsorption, and their influence order follows PO_4"3"− > SO_4"2"− > Cl"−. The study developed a new way to reutilize wastes and showed a great potential for resource recycling.

  8. Adsorption behavior and mechanism of Cr(VI) using Sakura waste from aqueous solution

    Energy Technology Data Exchange (ETDEWEB)

    Qi, Wenfang [School of Environmental Science and Engineering, Tianjin University, Tianjin 300072 (China); Zhao, Yingxin, E-mail: yingxinzhao@tju.edu.cn [School of Environmental Science and Engineering, Tianjin University, Tianjin 300072 (China); Tianjin Engineering Center of Urban River Eco-Purification Technology, Tianjin 300072 (China); Zheng, Xinyi [School of Environmental Science and Engineering, Tianjin University, Tianjin 300072 (China); Ji, Min [School of Environmental Science and Engineering, Tianjin University, Tianjin 300072 (China); Tianjin Engineering Center of Urban River Eco-Purification Technology, Tianjin 300072 (China); Zhang, Zhenya [Graduate School of Life and Environmental Sciences, University of Tsukuba, Tsukuba 3058572 (Japan)

    2016-01-01

    Graphical abstract: The main chemical components of Sakura leaves are cellulose 16.6%, hemicellulose 10.4%, lignin 18.3%, ash 11.4%, and others 43.3%. The adsorption capacity of Cr(VI) onto Sakura leaves can achieve 435.25 mg g{sup −1}, much higher than other similar agroforestry wastes. - Highlights: • Sakura leaves were prepared to remove Cr(VI) from aqueous solution. • The maximum adsorption capacity of Cr(VI) reached 435.25 mg g{sup −1}. • Cr(VI) adsorption fitted pseudo-second-order kinetic model. • Isotherm models indicated Cr(VI) adsorption occurred on a monolayer surface. • The influence order of coexisting ions followed PO{sub 4}{sup 3−} > SO{sub 4}{sup 2−} > Cl{sup −}. - Abstract: A forestall waste, Sakura leave, has been studied for the adsorption of Cr(VI) from aqueous solution. The materials before and after adsorption were characterized by X-ray photoelectron spectroscopy (XPS) and Fourier transform infrared spectroscopy (FTIR). To investigate the adsorption performance of Sakura waste, batch experiments were conducted under different adsorbent dosage, contact time, initial concentration of Cr(VI), and co-existing ions. Results showed the data fitted pseudo-second-order better than pseudo-first-order kinetic model. Equilibrium data was analyzed with Langmuir, Freundlich and Redlich–Peterson isotherm models at temperature ranges from 25 °C to 45 °C. The maximum adsorption capacity from the Langmuir model was 435.25 mg g{sup −1} at pH 1.0. The presence of Cl{sup −}, SO{sub 4}{sup 2−} and PO{sub 4}{sup 3−} would lead to an obvious negative effect on Cr(VI) adsorption, and their influence order follows PO{sub 4}{sup 3−} > SO{sub 4}{sup 2−} > Cl{sup −}. The study developed a new way to reutilize wastes and showed a great potential for resource recycling.

  9. Sealing of rock fractures around HLW repositories, 2

    International Nuclear Information System (INIS)

    Chigira, Masahiro

    1993-01-01

    During the flow of a silica-saturated hydrothermal solution in rock with negative temperature gradients, the behavior of silica in such solution is controlled by temperature, temperature gradient, pH, flow velocity, and solid surface area/fluid mass ratio (A/M). Such behavior could not be analysed precisely and totally at present state, but 'threshold conditions' have been found experimentally, under which solution keeps in equilibrium with solid silica in a flow field with temperature gradients. Solution keeps in equilibrium with solid silica under the conditions of A/M ratios more than 700 m 2 /kg, temperatures 80 - 120degC, temperature gradients less than 50degC/m, and pH 6 - 9, if mean pore velocities are less than 100 m/y. Under the same A/M ratios, temperature gradients, and pH, mean pore velocities must be less than 5 m/y in order to keep solution in equilibrium with solid silica in a flow field with temperatures 80 - 25degC. These 'threshold conditions' are expected to be satisfied in a near field of a repository of high-level radioactive waste, which suggests that if a groundwater is once saturated with silica under a higher temperature in a near field it would flow with decreasing temperatures in equilibrium with solid silica. In this case, the precipitation rate of amorphous silica along the flow path can be estimated without kinetic consideration. (author) 54 refs

  10. Thermoelastic analysis of spent fuel and high level radioactive waste repositories in salt. A semi-analytical solution

    International Nuclear Information System (INIS)

    St John, C.M.

    1977-04-01

    An underground repository containing heat generating, High Level Waste or Spent Unreprocessed Fuel may be approximated as a finite number of heat sources distributed across the plane of the repository. The resulting temperature, displacement and stress changes may be calculated using analytical solutions, providing linear thermoelasticity is assumed. This report documents a computer program based on this approach and gives results that form the basis for a comparison between the effects of disposing of High Level Waste and Spent Unreprocessed Fuel

  11. Emerging concepts and requirements for the long-term management of non-radioactive hazardous wastes - would geological disposal be an appropriate solution for some of these wastes

    International Nuclear Information System (INIS)

    Rein, K. von

    1994-01-01

    This work deals with the emerging concepts and requirements for the long-term management of non-radioactive hazardous wastes. After some generalities on the pollution of natural environment and the legislations taken by the swedish government the author tries to answer to the question : would geological disposal be an appropriate solution for the non-radioactive hazardous wastes? Then is given the general discussion of the last three articles concerning the background to current environmental policies and their implementation and more particularly the evolution and current thoughts about environmental policies, the managing hazardous activities and substances and the emerging concepts and requirements for the long-term management of non-radioactive hazardous wastes. Comments and questions concerning the similarity or otherwise between the present position of radioactive waste disposal and the background to current environmental policies are indicated. (O.L.)

  12. Research and demonstration results for a new "Double-Solution" technology for municipal solid waste treatment.

    Science.gov (United States)

    Erping, Li; Haoyun, Chen; Yanyang, Shang; Jun, Pan; Qing, Hu

    2017-11-01

    In this paper, the pyrolysis characteristics of six typical components in municipal solid waste (MSW) were investigated through a TG-FTIR combined technique and it was concluded that the main pyrolysis process of the biomass components (including food residues, sawdust and paper) occurred at 150-600°C. The main volatiles were multi-component gas including H 2 O, CO 2 , and CO. The main pyrolysis temperatures of three artificial products (PP, PVC and leather) was ranged from 200to 500°C. The wavelength of small molecule gases (CH 4 , CO 2 and CO) and the the chemical bonds (CO and CC) were observed in the infrared spectrum Based on the pyrolysis temperature interval and volatile constituent, a new "double-solution" process of pyrolysis and oxygen-enrichment decomposition MSW was designed. To achieve this process, a double-solution project was built for the direct treatment of MSW (10t/d). The complete setup of equipment and analysis of the byproducts has been reported in this paper to indicate the performance of this process. Energy balance and economic benefits were analysed for the process supporting. It was successfully demonstrated that the double-solution process was the environmentally friendly alternative method for MSW treatment in Chinese rural areas. Copyright © 2017 Elsevier Ltd. All rights reserved.

  13. Biosorption of clofibric acid and carbamazepine in aqueous solution by agricultural waste rice straw.

    Science.gov (United States)

    Liu, Zhanguang; Zhou, Xuefei; Chen, Xiaohua; Dai, Chaomeng; Zhang, Juan; Zhang, Yalei

    2013-12-01

    Due to their widespread use, clofibric acid (CA) and carbamazepine (CBZ) have been frequently detected simultaneously at relatively high concentrations in aquatic environments. In this study, agricultural waste rice straw was employed as a potentially low-cost, effective and easy-to-operate biosorbent (RSB) to remove CA and CBZ. The adsorption of both pharmaceuticals followed pseudo second-order kinetics, and intraparticle diffusion was an important rate-limiting step. The adsorption isotherms of both drugs were fit well with Freundlich model. The adsorption of CA onto RSB was exothermic and was more likely to be dominated by physical processes, while the adsorption of CBZ was endothermic. Solution pH was determined to be the most important factor for CA adsorption, such that the adsorption capacity of CA onto RSB increased with the decline of solution pH. In the lower range of solution pH below 3.1, the CA removal efficiency was enhanced with the increase of biosorbent dosage. The CBZ removal efficiency was enhanced with the increase of RSB dosage without pH control. The maximum adsorption capacities were 126.3 mg/g for CA and 40.0 mg/g for CBZ.

  14. Study on decontamination of radioactive ruthenium by steel wool in waste solution

    Energy Technology Data Exchange (ETDEWEB)

    Sugimoto, S; Sakaki, T [Radia Industry Co. Ltd., Takasaki, Gunma (Japan)

    1979-06-01

    Tracer experiments were done in order to establish a decontamination process of /sup 106/Ru in radioactive waste solution by column method paying special attention on the solution of nitrato-nitrosyl complex of Ru which is often encountered as a low level radioactive solution. It turned out that metallic iron was the most effective decontaminating agent among the several tens of materials tested. The decontamination factor (DF) of /sup 106/Ru increased in proportion to the total surface area of iron and it sensitively depended on the oxidation state of the surface as revealed by the batchwise and columnwise tests. Iron samples with high corrosiveness gave a much larger DF than those with low corrosiveness. The decontamination process proceeded as iron was being oxidized via Fe(metal) ..-->.. Fe(II) ..-->.. Fe(III). As the results, the DF initially increased after initiating the passage of water through the column but it then decreased as the oxidation process became inactive. An excellent durability up to 10000 bed volumes was demonstrated by the column method at a high average DF of 150.

  15. Production of a High-Level Waste Glass from Hanford Waste Samples

    International Nuclear Information System (INIS)

    Crawford, C.L.; Farrara, D.M.; Ha, B.C.; Bibler, N.E.

    1998-09-01

    The HLW glass was produced from a HLW sludge slurry (Envelope D Waste), eluate waste streams containing high levels of Cs-137 and Tc-99, solids containing both Sr-90 and transuranics (TRU), and glass-forming chemicals. The eluates and Sr-90/TRU solids were obtained from ion-exchange and precipitation pretreatments, respectively, of other Hanford supernate samples (Envelopes A, B and C Waste). The glass was vitrified by mixing the different waste streams with glass-forming chemicals in platinum/gold crucibles and heating the mixture to 1150 degree C. Resulting glass analyses indicated that the HLW glass waste form composition was close to the target composition. The targeted waste loading of Envelope D sludge solids in the HLW glass was 30.7 wt percent, exclusive of Na and Si oxides. Condensate samples from the off-gas condenser and off-gas dry-ice trap indicated that very little of the radionuclides were volatilized during vitrification. Microstructure analysis of the HLW glass using Scanning Electron Microscopy (SEM) and Energy Dispersive X-Ray Analysis (EDAX) showed what appeared to be iron spinel in the HLW glass. Further X-Ray Diffraction (XRD) analysis confirmed the presence of nickel spinel trevorite (NiFe2O4). These crystals did not degrade the leaching characteristics of the glass. The HLW glass waste form passed leach tests that included a standard 90 degree C Product Consistency Test (PCT) and a modified version of the United States Environmental Protection Agency Toxicity Characteristic Leaching Procedure (TCLP)

  16. Advanced waste form and melter development for treatment of troublesome high-level wastes

    Energy Technology Data Exchange (ETDEWEB)

    Marra, James [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Kim, Dong -Sang [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Maio, Vincent [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-09-02

    A number of waste components in US defense high level radioactive wastes (HLW) have proven challenging for current Joule heated ceramic melter (JHCM) operations and have limited the ability to increase waste loadings beyond already realized levels. Many of these "troublesome" waste species cause crystallization in the glass melt that can negatively impact product quality or have a deleterious effect on melter processing. Recent efforts at US Department of Energy laboratories have focused on understanding crystallization behavior within HLW glass melts and investigating approached to mitigate the impacts of crystallization so that increases in waste loading can be realized. Advanced glass formulations have been developed to highlight the unique benefits of next-generation melter technologies such as the Cold Crucible Induction Melter (CCIM). Crystal-tolerant HLW glasses have been investigated to allow sparingly soluble components such as chromium to crystallize in the melter but pass out of the melter before accumulating.

  17. Separation processes for high-level radioactive waste treatment

    International Nuclear Information System (INIS)

    Sutherland, D.G.

    1992-11-01

    During World War II, production of nuclear materials in the United States for national defense, high-level waste (HLW) was generated as a byproduct. Since that time, further quantities of HLW radionuclides have been generated by continued nuclear materials production, research, and the commercial nuclear power program. In this paper HLW is defined as the highly radioactive material resulting from the processing of spent nuclear fuel. The HLW is the liquid waste generated during the recovery of uranium and plutonium in a fuel processing plant that generally contains more than 99% of the nonvolatile fission products produced during reactor operation. Since this paper deals with waste separation processes, spent reactor fuel elements that have not been dissolved and further processed are excluded

  18. 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)

  19. Colloid Genesis/Transport and Flow Pathway Alterations Resulting From Interactions of Reactive Waste Solutions and Hanford Vadose Zone Sediments

    International Nuclear Information System (INIS)

    Wan, Jiamin; Tokunaga, Tetsu K.

    2001-01-01

    Leakage of underground tanks containing high-level nuclear waste solutions has been identified at various DOE facilities. The Hanford Site is one the main facilities of concern, with about 2,300 to 3,400 m3 of leaked waste liquids. Radionuclides and other contaminants have been found in elevated concentrations in the vadose zone and groundwater underneath single shell tank farms. We do not currently know the mechanisms responsible for the unexpected deep migration of some contaminants through the vadose zone, and such understanding is urgently needed for planning remediation. Due to the extreme chemical conditions of the tank waste solutions (very high pH, aluminum concentration, and ionic strength), interactions between the highly reactive waste solutions and sediments underneath the tanks can result in dissolution of primary minerals of the sediments and precipitation of secondary phases including colloidal particles. Contaminants can sorb onto and/or co-precipitate with the secondary phases. Therefore transport of strongly associated contaminants on mobile colloids can be substantially greater than without colloids. The overall objective of this research is to improve our understanding on the effects of interactions between the tank waste solution and sediments on deep contaminant migration under Hanford Site conditions. This objective will be achieved through the following four tasks: (1) colloid generation and transport studies, (2) studies on sediment permeability and chemical composition alterations, (3) quantifying associations of contaminants with secondary colloids, and (4) studies on the combined effects of the aforementioned processes on deep contaminant migration

  20. 77 FR 38789 - Notice of Availability of Draft Waste Incidental to Reprocessing Evaluation for the Concentrator...

    Science.gov (United States)

    2012-06-29

    ... disposal facility, either the Area 5 Radioactive Waste Management Site at DOE's Nevada National Security... offsite LLW disposal facility, either the NNSS Area 5 Radioactive Waste Management Site or the Waste... radioactive waste (HLW) and may be managed and disposed of offsite as low-level waste (LLW). DOE prepared the...

  1. 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.

  2. Challenges in development of matrices for vitrification of old legacy waste and high-level radioactive waste generated from reprocessing of AHWR and FBR spent fuel

    International Nuclear Information System (INIS)

    Kaushik, C.P.

    2012-01-01

    Majority of radioactivity in entire nuclear fuel cycle is concentrated in HLW. A three step strategy for management of HLW has been adopted in India. This involves immobilization of waste oxides in stable and inert solid matrices, interim retrievable storage of the conditioned waste product under continuous cooling and disposal in deep geological formations. Glass has been accepted as most suitable matrix world-wide for immobilization of HLW, because of its attractive features like ability to accommodate wide range of waste constituents, modest processing temperatures, adequate chemical, thermal and radiation stability. Borosilicate glass matrix developed by BARC in collaboration with CGCRI has been adopted in India for immobilization of HLW. In view of compositional variation of HLW from site to site, tailor make changes in the glass formulations are often necessary to incorporate all the waste constituents and having the product of desirable characteristics. The vitrified waste products made with different glass formulations and simulated waste need to be characterized for chemical durability, thermal stability, homogeneity etc. before finalizing a suitable glass formulation. The present extended abstract summarises the studies carried out for development of glass formulations for vitrification of legacy waste and futuristic waste likely to be generated from AHWR and FBR having wide variations in their compositions. The presently stored HLW at Trombay is characterized by significant concentrations of uranium, sodium and sulphate in addition to fission products, corrosion products and small amount of other actinides

  3. Conceptualization of a hypothetical high-level nuclear waste repository site in unsaturated, fractured tuff

    International Nuclear Information System (INIS)

    Parsons, A.M.; Olague, N.E.; Gallegos, D.P.

    1991-01-01

    Under the sponsorship of the US Nuclear Regulatory Commission (NRC), Sandia National Laboratories (SNL) is developing a performance assessment methodology for the analysis of long-term disposal and isolation of high-level nuclear wastes (HLW) in alternative geologic media. As part of this exercise, SNL created a conceptualization of ground-water flow and radionuclide transport in the far field of a hypothetical HLW repository site located in unsaturated, fractured tuff formations. This study provides a foundation for the development of conceptual mathematical, and numerical models to be used in this performance assessment methodology. This conceptualization is site specific in terms of geometry, the regional ground-water fl