Radioactive waste isolation in salt: peer review of the Office of Nuclear Waste Isolation's report on Functional Design Criteria for a Repository for High-Level Radioactive Waste

International Nuclear Information System (INIS)

Hambley, D.F.; Russell, J.E.; Busch, J.S.; Harrison, W.; Edgar, D.E.; Tisue, M.W.

1984-08-01

This report summarizes Argonne's review of the Office of Nuclear Waste Isolation's (ONWI's) draft report entitled Functional Design Criteria for High-Level Nuclear Waste Repository in Salt, dated January 23, 1984. Recommendations are given for improving the ONWI draft report

Production and properties of solidified high-level waste

International Nuclear Information System (INIS)

Brodersen, K.

1980-08-01

Available information on production and properties of solidified high-level waste are presented. The review includes literature up to the end of 1979. The feasibility of production of various types of solidified high-level wast is investigated. The main emphasis is on borosilicate glass but other options are also mentioned. The expected long-term behaviour of the materials are discussed on the basis of available results from laboratory experiments. Examples of the use of the information in safety analysis of disposal in salt formations are given. The work has been made on behalf of the Danish utilities investigation of the possibilities of disposal of high-level waste in salt domes in Jutland. (author)

Cementitious Stabilization of Mixed Wastes with High Salt Loadings

International Nuclear Information System (INIS)

Spence, R.D.; Burgess, M.W.; Fedorov, V.V.; Downing, D.J.

1999-01-01

Salt loadings approaching 50 wt % were tolerated in cementitious waste forms that still met leach and strength criteria, addressing a Technology Deficiency of low salt loadings previously identified by the Mixed Waste Focus Area. A statistical design quantified the effect of different stabilizing ingredients and salt loading on performance at lower loadings, allowing selection of the more effective ingredients for studying the higher salt loadings. In general, the final waste form needed to consist of 25 wt % of the dry stabilizing ingredients to meet the criteria used and 25 wt % water to form a workable paste, leaving 50 wt % for waste solids. The salt loading depends on the salt content of the waste solids but could be as high as 50 wt % if all the waste solids are salt

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

International Nuclear Information System (INIS)

Rothfuchs, T.; Duijves, K.A.

1990-04-01

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

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

International Nuclear Information System (INIS)

Rothfuchs, T.

1991-01-01

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

Radiolytic bubble formation and level changes in simulated high-level waste salts and sludges -- application to Savannah River Site and Hanford Storage tanks

International Nuclear Information System (INIS)

Walker, D.D.; Crawford, C.L.; Bibler, N.E.

1993-01-01

Radiolytically-produced bubbles of trapped gas are observed in simulated high-level waste (HLW) damp salt cake exposed to Co-60 gamma radiation. As the damp salt cake is irradiated, its volume increases due to the formation of trapped gas bubbles. Based on the increase in volume, the rate of trapped gas generation varies between 0.04 and 0.2 molecules/100 eV of energy deposited in the damp salt cake. The maximum volume of trapped gas observed in experiments is in the range 21--26 vol %. After reaching these volumes, the gas bubbles begin to escape. The generated gas includes hydrogen, oxygen, and nitrous oxide. The ratio in which these components are produced depends on the composition of the waste. Nitrous oxide production increases with the amount of sodium nitrite. Gases trapped by this mechanism may account for some of the observed level changes in Savannah River Site and Hanford waste tanks

Temperature calculations on different configurations for disposal of high-level reprocessing waste in a salt dome model

International Nuclear Information System (INIS)

Hamstra, J.; Kevenaar, J.W.A.M.

1978-06-01

A medium size salt dome is considered as a structure in which a repository can be located for all radioactive wastes to be produced within the scope of a postulated nuclear power program. A dominating design factor for the lay-out of such a waste repository is the temperature distribution in the salt dome resulting from decay heat released from the buried solidified high-level reprocessing waste. Two numerical models are presented for the calculation of both global and local rock salt temperatures. The results of calculations performed with these models are demonstrated to be compatible. Rock salt temperatures related to several types of burial configurations, ranging from two layer configurations with various vertical distances between the layers via a three and a four layer repository to deep bore hole concepts varying from 100 to 600 m bore hole depth, can therefore be calculated with one rather simple unit cell model. The results of these calculations indicate that rock salt temperatures can be kept within acceptable limits to realize a repository using standard mining techniques. The temperatures at mine galery level prove to be a dominating factor in the selection of a repository configuration. More detailed calculations of these temperatures taking into account the loading sequence and the cooling capacity of the mine ventilation are recommended. Finally the apparent advantages of a deep bore hole concept emphasize the need for R and D work with respect to advanced drilling techniques in order to achieve deep dry disposal bore holes that can be realized from a burial mine in the salt dome. (Auth.)

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

International Nuclear Information System (INIS)

Langton, C.A.

1984-01-01

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

The function of packing materials in a high-level nuclear waste repository and some candidate materials: Salt Repository Project

International Nuclear Information System (INIS)

Bunnell, L.R.; Shade, J.W.

1987-03-01

Packing materials should be included in waste package design for a high-level nuclear waste repository in salt. A packing material barrier would increase confidence in the waste package by alleviating possible shortcomings in the present design and prolonging confinement capabilities. Packing materials have been studied for uses in other geologic repositories; appropriately chosen, they would enhance the confinement capabilities of salt repository waste packages in several ways. Benefits of packing materials include retarding or chemically modifying brines to reduce corrosion of the waste package, providing good thermal conductivity between the waste package and host rock, retarding or absorbing radionuclides, and reducing the massiveness of the waste package. These benefits are available at low percentage of total repository cost, if the packing material is properly chosen and used. Several candidate materials are being considered, including oxides, hydroxides, silicates, cement-based mixtures, and clay mixtures. 18 refs

Conceptual process for conversion of high level waste to glass

International Nuclear Information System (INIS)

1975-01-01

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

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

International Nuclear Information System (INIS)

1984-01-01

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

Comparison of temperature calculations for an arbitrary high-level waste disposal configuration in salt formations

International Nuclear Information System (INIS)

Kevenaar, J.W.A.M.; Janssen, L.G.J.; Ploumen, P.; Winske, P.

1979-05-01

The objective of this report is the comparison of the results of temperature analyses for an arbitrary high-level radioactive waste disposal configuration in salt formations. The analyses were carried out at the RWTH and ECN. The computer programs used are based on finite difference and finite element techniques. From the local temperature analyses that were intended to check the solution techniques, it could be concluded that both finite difference and finite elements are capable to analyse this type of problems. From the global temperature analyses it could be concluded that both analysis approaches: temperature dependent and iteratively determined temperature independent material properties, are suited to analyse the global temperature distribution in the salt formation

Aspects on the gas generation and migration in repositories for high level waste in salt formations

International Nuclear Information System (INIS)

Ruebel, Andre; Buhmann, Dieter; Meleshyn, Artur; Moenig, Joerg; Spiessl, Sabine

2013-07-01

In a deep geological repository for high-level waste, gases may be produced during the post-closure phase by several processes. The generated gases can potentially affect safety relevant features and processes of the repository, like the barrier integrity, the transport of liquids and gases in the repository and the release of gaseous radionuclides from the repository into the biosphere. German long-term safety assessments for repositories for high-level waste in salt which were performed prior 2010 did not explicitly consider gas transport and the consequences from release of volatile radionuclides. Selected aspects of the generation and migration of gases in repositories for high-level waste in a salt formation are studied in this report from the viewpoint of the performance assessment. The knowledge on the availability of water in the repository, in particular the migration of salt rock internal fluids in the temperature field of the radioactive waste repository towards the emplacement drifts, was compiled and the amount of water was roughly estimated. Two other processes studied in this report are on the one hand the release of gaseous radionuclides from the repository and their potential impact in the biosphere and on the other hand the transport of gases along the drifts and shafts of the repository and their interaction with the fluid flow. The results presented show that there is some gas production expected to occur in the repository due to corrosion of container material from water disposed of with the backfill and inflowing from the host rock during the thermal phase. If not dedicated gas storage areas are foreseen in the repository concept, these gases might exceed the storage capacity for gases in the repository. Consequently, an outflow of gases from the repository could occur. If there are failed containers for spent fuel, radioactive gases might be released from the containers into the gas space of the backfill and subsequently transported together

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

International Nuclear Information System (INIS)

Kuhn, K.; Rothfuchs, T.

1989-01-01

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

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

International Nuclear Information System (INIS)

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

1995-01-01

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

Expected brine movement at potential nuclear waste repository salt sites

International Nuclear Information System (INIS)

McCauley, V.S.; Raines, G.E.

1987-08-01

The BRINEMIG brine migration code predicts rates and quantities of brine migration to a waste package emplaced in a high-level nuclear waste repository in salt. The BRINEMIG code is an explicit time-marching finite-difference code that solves a mass balance equation and uses the Jenks equation to predict velocities of brine migration. Predictions were made for the seven potentially acceptable salt sites under consideration as locations for the first US high-level nuclear waste repository. Predicted total quantities of accumulated brine were on the order of 1 m 3 brine per waste package or less. Less brine accumulation is expected at domal salt sites because of the lower initial moisture contents relative to bedded salt sites. Less total accumulation of brine is predicted for spent fuel than for commercial high-level waste because of the lower temperatures generated by spent fuel. 11 refs., 36 figs., 29 tabs

High level nuclear waste repository in salt: Sealing systems status and planning report: Draft report

International Nuclear Information System (INIS)

1985-09-01

This report documents the initial conceptual design studies for a repository sealing system for a high-level nuclear waste repository in salt. The first step in the initial design studies was to review the current design level, termed schematic designs. This review identified practicality of construction and development of a design methodology as two key issues for the conceptual design. These two issues were then investigated during the initial design studies for seal system materials, seal placement, backfill emplacement, and a testing and monitoring plan. The results of these studies have been used to develop a program plan for completion of the sealing system conceptual design. 60 refs., 26 figs., 18 tabs

Preliminary investigation results as applied to utilization of Ukrainian salt formations for disposal of high-level radioactive waste

International Nuclear Information System (INIS)

Shekhunova, S.B.; Khrushchov, D.P.; Petrichenko, O.I.

1994-01-01

The salt-bearing formations have been investigated in five regions of Ukraine. Upper Devonian and Lower Permian evaporite formations in Dnieper-Donets Depression and in the NW part of Donets basin are considered to be promising for disposal of high-level radioactive waste (HLRW). Rock salt occurs there either as bedded salts or as salt pillows and salt diapirs. Preliminary studies have resulted in selection of several candidate sites that show promise for construction of a subsurface pilot lab. Ten salt domes and two sites in bedded salts have been proposed for further exploration. Based on microstructural studies it is possible to separate the body of a salt structure and to locate within its limits the rock salt structure and to locate within its limits the rock salt blocks of different genesis, i.e.: (a) blocks characteristic of initial undisturbed sedimentary structure; (b) flow zones; (c) sliding planes; (d) bodies of loose or uncompacted rock salt. Ultramicrochemical examination of inclusions in halite have shown that they are composed of more than 40 minerals. It is emphasized that to assess suitability of a structure for construction of subsurface lab, and also the potential construction depth intervals, account should be taken of the results of ultra microchemical and microstructural data

Thermal denitration of high concentration nitrate salts waste water

International Nuclear Information System (INIS)

Hwang, D. S.; Oh, J. H.; Choi, Y. D.; Hwang, S. T.; Park, J. H.; Latge, C.

2003-01-01

This study investigated the thermodynamic and the thermal decomposition properties of high concentration nitrate salts waste water for the lagoon sludge treatment. The thermodynamic property was carried out by COACH and GEMINI II based on the composition of nitrate salts waste water. The thermal decomposition property was carried out by TG-DTA and XRD. Ammonium nitrate and sodium nitrate were decomposed at 250 .deg. C and 730 . deg. C, respectively. Sodium nitrate could be decomposed at 450 .deg. C in the case of adding alumina for converting unstable Na 2 O into stable Na 2 O.Al 2 O 3 . The flow sheet for nitrate salts waste water treatment was proposed based on the these properties data. These will be used by the basic data of the process simulation

On the time-dependent behavior of a cylindrical salt dome with a high-level waste repository

International Nuclear Information System (INIS)

Prij, J.

1988-01-01

In a salt dome with a repository for high-level radioactive and heat-generating waste, thermal stresses develop. These stresses can influence the isolation capability of the salt dome if these stresses can initiate cracks or introduce movements along existing closed flaws. The influence of the thermomechanical properties of the rock salt and the surrounding rocks on the thermal stresses and the surface rise is discussed. This discussion is based on a number of finite element creep analyses of a homogeneous cylindrical salt dome. The parameters, varied in the analyses, are constants in the thermomechanical constitutive behavior of salt and rocks, and furthermore the thermal loading has been varied. It is shown that variations in the creep properties, which result in differences in creep strain rate of a factor of 100, have only a very limited influence on the thermal stresses and the surface rise. Of more importance is the elastic stiffness of the materials. In all creep analyses the thermal stresses in the salt are compressive and the shear stresses remain below 2 MPa. The results are evaluated using an analytical treatment. Based on this evaluation, it is shown that the observed trends in the numerical results have a more general character and are not strictly limited to the geometry chosen. It is concluded that the thermal stresses in the salt formation are not strongly dependent on the creep properties of the rock salt

Radiant energy dissipation during final storage of high-level radioactive waste in rock salt

International Nuclear Information System (INIS)

Ramthun, H.

1981-08-01

A final disposal concept is assumed where the high-active waste from 1400 t of uranium, remaining after conditioning, is solidified in borosilicate glass and distributed in 1.760 waste casks. These containers 1.2 m in height and 0.3 m in diameter are to be buried 10 years after the fuel is removed from the reactor in the 300 m deep boreholes of a salt dome. For this design the mean absorbed dose rates are calculated in the glass die (3.9 Gy/s), the steel mantle (0.26 Gy/s) and in the salt rock (0.12 Gy/s at a distance of 1 cm and 0.034 Gy/s at a distance of 9 cm from the container surface) valid at the beginning of disposal. The risk involved with these amounts of stored lattice energy is shortly discussed. (orig.) [de

Destruction of high explosives and wastes containing high explosives using the molten salt destruction process

International Nuclear Information System (INIS)

Upadhye, R.S.; Brummond, W.A.; Pruneda, C.O.

1992-01-01

This paper reports the Molten Salt Destruction (MSD) Process which has been demonstrated for the destruction of HE and HE-containing wastes. MSD has been used by Rockwell International and by Anti-Pollution Systems to destroy hazardous wastes. MSD converts the organic constituents (including the HE) of the waste into non-hazardous substances such as carbon dioxide, nitrogen and water. In the case of HE-containing mixed wastes, any actinides in the waste are retained in the molten salt, thus converting the mixed wastes into low-level wastes. (Even though the MSD process is applicable to mixed wastes, this paper will emphasize HE-treatment.) The destruction of HE is accomplished by introducing it, together with oxidant gases, into a crucible containing a molten salt, such as sodium carbonate, or a suitable mixture of the carbonates of sodium, potassium, lithium and calcium. The temperature of the molten salt can be between 400 to 900 degrees C. The combustible organic components of the waste react with oxygen to produce carbon dioxide, nitrogen and steam

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

International Nuclear Information System (INIS)

Wieczorek, K.; Klarr, K.

1993-01-01

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

Fracture toughness measurements on a glass bonded sodalite high-level waste form

International Nuclear Information System (INIS)

DiSanto, T.; Goff, K. M.; Johnson, S. G.; O'Holleran, T. P.

1999-01-01

The electrometallurgical treatment of metallic spent nuclear fuel produces two high-level waste streams; cladding hulls and chloride salt. Argonne National Laboratory is developing a glass bonded sodalite waste form to immobilize the salt waste stream. The waste form consists of 75 Vol.% crystalline sodalite (containing the salt) with 25 Vol.% of an ''intergranular'' glassy phase. Microindentation fracture toughness measurements were performed on representative samples of this material using a Vickers indenter. Palmqvist cracking was confirmed by post-indentation polishing of a test sample. Young's modulus was measured by an acoustic technique. Fracture toughness, microhardness, and Young's modulus values are reported, along with results from scanning electron microscopy studies

Disposal of Savannah River Plant waste salt

International Nuclear Information System (INIS)

Dukes, M.D.

1982-01-01

Approximately 26-million gallons of soluble low-level waste salts will be produced during solidification of 6-million gallons of high-level defense waste in the proposed Defense Waste Processing Facility (DWPF) at the Savannah River Plant (SRP). Soluble wastes (primarily NaNO 3 , NaNO 2 , and NaOH) stored in the waste tanks will be decontaminated by ion exchange and solidified in concrete. The resulting salt-concrete mixture, saltcrete, will be placed in a landfill on the plantsite such that all applicable federal and state disposal criteria are met. Proposed NRC guidelines for the disposal of waste with the radionuclide content of SRP salt would permit shallow land burial. Federal and state rules require that potentially hazardous chemical wastes (mainly nitrate-nitrate salts in the saltcrete) be contained to the degree necessary to meet drinking water standards in the ground water beneath the landfill boundary. This paper describes the proposed saltcrete landfill and tests under way to ensure that the landfill will meet these criteria. The work includes laboratory and field tests of the saltcrete itself, a field test of a one-tenth linear scale model of the entire landfill system, and a numerical model of the system

Independent Assessment of the Savannah River Site High-Level Waste Salt Disposition Alternatives Evaluation

International Nuclear Information System (INIS)

Case, J. T.; Renfro, M. L.

1998-01-01

This report presents the results of the Independent Project Evaluation (IPE) Team assessment of the Westinghouse Savannah River Company High-Level Waste Salt Disposition Systems Engineering (SE) Team's deliberations, evaluations, and selections. The Westinghouse Savannah River Company concluded in early 1998 that production goals and safety requirements for processing SRS HLW salt to remove Cs-137 could not be met in the existing In-Tank Precipitation Facility as currently configured for precipitation of cesium tetraphenylborate. The SE Team was chartered to evaluate and recommend an alternative(s) for processing the existing HLW salt to remove Cs-137. To replace the In-Tank Precipitation process, the Savannah River Site HLW Salt Disposition SE Team down-selected (October 1998) 140 candidate separation technologies to two alternatives: Small-Tank Tetraphenylborate (TPB) Precipitation (primary alternative) and Crystalline Silicotitanate (CST) Nonelutable Ion Exchange (backup alternative). The IPE Team, commissioned by the Department of Energy, concurs that both alternatives are technically feasible and should meet all salt disposition requirements. But the IPE Team judges that the SE Team's qualitative criteria and judgments used in their down-selection to a primary and a backup alternative do not clearly discriminate between the two alternatives. To properly choose between Small-Tank TPB and CST Ion Exchange for the primary alternative, the IPE Team suggests the following path forward: Complete all essential R and D activities for both alternatives and formulate an appropriate set of quantitative decision criteria that will be rigorously applied at the end of the R and D activities. Concurrent conceptual design activities should be limited to common elements of the alternatives

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

International Nuclear Information System (INIS)

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

1988-01-01

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

Sandia studies of high-level waste canisters and overpacks applicable for a salt repository

International Nuclear Information System (INIS)

Molecke, M.A.; Schaefer, D.W.; Glass, R.S.; Ruppen, J.A.

1982-01-01

An experimental program to develop candidate materials for use as high-level waste (HLW) overpacks or canisters in a salt repository has been in progress at Sandia National Laboratories since 1976. The main objective of this program is to provide a waste package barrier having a long lifetime in the chemical and physical environment of a repository. This paper summarizes the recent corrosion and metallurgical study results for the prime overpack material, TiCode-12, in the areas of uniform corrosion (extremely low rate and extent); local attack, e.g., pits and crevices (none were found); stress corrosion cracking susceptibility (no significant changes in macroscopic tensile properties were detected); hydrogen sorption-embrittlement effects; effects of gamma irradiation in solution; and sensitization effects (testing is still in process in the last three areas). Previous candidate screening analyses on other alloys and recent work on alternate overpack alloys are reviewed. All phases of these interrelated laboratory, hot-cell, and field experimental studies are described. 16 references, 8 figures, 4 tables

Hanford long-term high-level waste management program overview

International Nuclear Information System (INIS)

Reep, I.E.

1978-05-01

The objective is the long-term disposition of the defense high-level radioactive waste which will remain upon completion of the interim waste management program in the mid-1980s, plus any additional high-level defense waste resulting from the future operation of N Reactor and the Purex Plant. The high-level radioactive waste which will exist in the mid-1980s and is addressed by this plan consists of approximately 3,300,000 ft 3 of damp salt cake stored in single-shell and double-shell waste tanks, 1,500,000 ft 3 of damp sludge stored in single-shell and double-shell waste tanks, 11,000,000 gallons of residual liquor stored in double-shell waste tanks, 3,000,000 gallons of liquid wastes stored in double-shell waste tanks awaiting solidification, and 2,900 capsules of 90 SR and 137 Cs compounds stored in water basins. Final quantities of waste may be 5 to 10% greater, depending on the future operation of N Reactor and the Purex Plant and the application of waste treatment techniques currently under study to reduce the inventory of residual liquor. In this report, the high-level radioactive waste addressed by this plan is briefly described, the major alternatives and strategies for long-term waste management are discussed, and a description of the long-term high-level waste management program is presented. Separate plans are being prepared for the long-term management of radioactive wastes which exist in other forms. 14 figures

Alternatives for definse waste-salt disposal

International Nuclear Information System (INIS)

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

1983-01-01

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

Preliminary evaluation of alternative forms for immobilization of Hanford high-level defense wastes

International Nuclear Information System (INIS)

Schulz, W.W.; Beary, M.M.; Gallagher, S.A.; Higley, B.A.; Johnston, R.G.; Jungfleisch, F.M.; Kupfer, M.J.; Palmer, R.A.; Watrous, R.A.; Wolf, G.A.

1980-09-01

A preliminary evaluation of solid waste forms for immobilization of Hanford high-level radioactive defense wastes is presented. Nineteen different waste forms were evaluated and compared to determine their applicability and suitability for immobilization of Hanford salt cake, sludge, and residual liquid. This assessment was structured to address waste forms/processes for several different leave-retrieve long-term Hanford waste management alternatives which give rise to four different generic fractions: (1) sludge plus long-lived radionuclide concentrate from salt cake and residual liquid; (2) blended wastes (salt cake plus sludge plus residual liquid); (3) residual liquid; and (4) radionuclide concentrate from residual liquid. Waste forms were evaluated and ranked on the basis of weighted ratings of seven waste form and seven process characteristics. Borosilicate Glass waste forms, as marbles or monoliths, rank among the first three choices for fixation of all Hanford high-level wastes (HLW). Supergrout Concrete (akin to Oak Ridge National Laboratory Hydrofracture Process concrete) and Bitumen, low-temperature waste forms, rate high for bulk disposal immobilization of high-sodium blended wastes and residual liquid. Certain multi-barrier (e.g., Coated Ceramic) and ceramic (SYNROC Ceramic, Tailored Ceramics, and Supercalcine Ceramic) waste forms, along with Borosilicate Glass, are rated as the most satisfactory forms in which to incorporate sludges and associated radionuclide concentrates. The Sol-Gel process appears superior to other processes for manufacture of a generic ceramic waste form for fixation of Hanford sludge. Appropriate recommendations for further research and development work on top ranking waste forms are made

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

International Nuclear Information System (INIS)

Rothfuchs, T.

1986-01-01

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

Conditioning matrices from high level waste resulting from pyrochemical processing in fluorine salt

International Nuclear Information System (INIS)

Grandjean, Agnes; Advocat, Thierry; Bousquet, Nicolas; Jegou, Christophe

2007-01-01

Separating the actinides from the fission products through reductive extraction by aluminium in a LiF/AlF 3 medium is a process investigated for pyrometallurgical reprocessing of spent fuel. The process involves separation by reductive salt-metal extraction. After dissolving the fuel or the transmutation target in a salt bath, the noble metal fission products are first extracted by contacting them with a slightly reducing metal. After extracting the metal fission products, then the actinides are selectively separated from the remaining fission products. In this hypothesis, all the unrecoverable fission products would be conditioned as fluorides. Therefore, this process will generate first a metallic waste containing the 'reducible' fission products (Pd, Mo, Ru, Rh, Tc, etc.) and a fluorine waste containing alkali-metal, alkaline-earth and rare earth fission products. Immobilization of these wastes in classical borosilicate glasses is not feasible due to the very low solubility of noble metals, and of fluoride in these hosts. Alternative candidates have therefore been developed including silicate glass/ceramic system for fluoride fission products and metallic ones for noble metal fission products. These waste-forms were evaluated for their confinement properties like homogeneity, waste loading, volatility during the elaboration process, chemical durability, etc. using appropriate techniques. (authors)

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

International Nuclear Information System (INIS)

1984-01-01

This report covers the contract period 1976-77, as well as some of the tasks carried out during the extension in 1978, in the framework of the R and D programme for disposal of radioactive waste in salt formations. With regard to the in-situ tests for the liberation and migration of brine, the testing devices were examined successfully. Laboratory examinations carried out showed a stepwise liberation of the water contents in halite in dependence on the temperature. The amount of brine liberated stood in good agreement with the in situ results. A temperature test for borehole convergence resulted in definite convergence rates. Simultaneously no influence was registered in the stability of the surrounding rocks. For the realization of an integrated major experiment, temperature test field IV was mined on the 750 m level of the Asse Salt Mine and heater- as well as measurement drillings were carried out. Extensive rheological examinations are concentrated particularly on the halite and secondly on the Carnallite. They are chiefly based on uni- and multiaxial pressure tests. Computer programmes are developed to examine the heat generation in wastes as well as in salt. In comparison, the programme development of computer codes for the stability behaviour of rocks is still at a relatively early stage, because it has to build up on the results of heat generation. The works for the development of a transport container with a shielding combination are at a very advanced stage. An integrated disposal- and retrieval system was developed, tested and successfully demonstrated. A monitoring system in the mine has also been developed in its essential parts

Salt splitting of sodium-dominated radioactive waste using ceramic membranes

International Nuclear Information System (INIS)

Hollenberg, G.W.; Carlson, C.D.; Virkar, A.; Joshi, A.

1994-08-01

The potential for salt splitting of sodium dominated radioactive wastes by use of a ceramic membrane is reviewed. The technical basis for considering this processing technology is derived from the technology developed for battery and chlor-alkali chemical industry. Specific comparisons are made with the commercial organic membranes which are the standard in nonradioactive salt splitting. Two features of ceramic membranes are expected to be especially attractive: high tolerance to gamma irradiation and high selectivity between sodium and other ions. The objective of the salt splitting process is to separate nonradioactive sodium from contaminated sodium salts prior to other pretreatment processes in order to: (1) concentrate the waste in order to reduce the volume of subsequent additives and capacity of equipment, (2) decrease the pH of the waste in preparation for further processing, and (3) provide sodium with very low radioactivity levels for caustic washing of sludge or low level and mixed waste vitrification

Disposal of high-level waste from nuclear power plants in Denmark. Salt dome investigations. v.2

International Nuclear Information System (INIS)

1981-01-01

The present report deals with the geological investigations performed for determing the feasibility of a repository for high-level waste in a salt dome. It is volume 2 of five volumes that together constitute the final report of the Danish utilities' salt dome investigations. The purpose of the work was to procure a more detailed knowledge of the geology of salt domes in North Jutland on example of Mors. The Mors dome is oval with the two axes of approx. 12.5 km and 8 km respectively. Two deep wells have been drilled into the salt. These wells reach 3400-3500 m below surface. Until a depth of about 3200 m Erslev 2 passes through rock salt of Zechstein 1 which is the oldest evaporite series. However, it could also be interlayed with the slightly younger Zechstein 2. At about 3200 m a marker layer was met with Zechstein 2 salt below. Interpretation of cores and results of downhole electromagnetic and borehole gravimetric measurements show that there is a large area around Erslev 2 which consists of very pure sodium chloride with traces of anhydrite (calcium, sulphate) 1-3%. This area is used for the repository design and safety evaluation. The hydrological conditions existing in the strata above the salt dome (caprock) have been investigated with the help of four hydrogeological wells, placed two each, on two different sites. The cores themselves were taken at various depths in all four holes. With these laboratory methods it has been possible to measure data relevant to hydrology - such as porosity and permeability - as well as geochemistry. (BP)

In-situ nitrite analysis in high level waste tanks

International Nuclear Information System (INIS)

O'Rourke, P.E.; Prather, W.S.; Livingston, R.R.

1992-01-01

The Savannah River Site produces special nuclear materials used in the defense of the United States. Most of the processes at SRS are primarily chemical separations and purifications. In-situ chemical analyses help improve the safety, efficiency and quality of these operations. One area where in situ fiberoptic spectroscopy can have a great impact is the management of high level radioactive waste. High level radioactive waste at SRS is stored in more than 50 large waste tanks. The waste exists as a slurry of nitrate salts and metal hydroxides at pH's higher than 10. Sodium Nitrite is added to the tanks as a corrosion inhibitor. In-situ fiberoptic probes are being developed to measure the nitrate, nitrite and hydroxide concentrations in both liquid and solid fractions. Nitrite levels can be measured between 0.01M and 1M in a 1mm pathlength optical cell

Site characterization plan conceptual design report for a high-level nuclear waste repository in salt, vertical emplacement mode: Volume 1

International Nuclear Information System (INIS)

1987-12-01

This Conceptual Design Report describes the conceptual design of a high-level nuclear waste repository in salt at a proposed site in Deaf Smith County, Texas. Waste receipt, processing, packing, and other surface facility operations are described. Operations in the shafts underground are described, including waste hoisting, transfer, and vertical emplacement. This report specifically addresses the vertical emplacement mode, the reference design for the repository. Waste retrieval capability is described. The report includes a description of the layout of the surface, shafts, and underground. Major equipment items are identified. The report includes plans for decommissioning and sealing of the facility. The report discusses how the repository will satisfy performance objectives. Chapters are included on basis for design, design analyses, and data requirements for completion of future design efforts. 105 figs., 52 tabs

Site characterization plan conceptual design report for a high-level nuclear waste repository in salt, vertical emplacement mode: Volume 1

Energy Technology Data Exchange (ETDEWEB)

1987-12-01

This Conceptual Design Report describes the conceptual design of a high-level nuclear waste repository in salt at a proposed site in Deaf Smith County, Texas. Waste receipt, processing, packing, and other surface facility operations are described. Operations in the shafts underground are described, including waste hoisting, transfer, and vertical emplacement. This report specifically addresses the vertical emplacement mode, the reference design for the repository. Waste retrieval capability is described. The report includes a description of the layout of the surface, shafts, and underground. Major equipment items are identified. The report includes plans for decommissioning and sealing of the facility. The report discusses how the repository will satisfy performance objectives. Chapters are included on basis for design, design analyses, and data requirements for completion of future design efforts. 105 figs., 52 tabs.

Site characterization plan conceptual design report for a high-level nuclear waste repository in salt, horizontal emplacment mode: Volume 1

International Nuclear Information System (INIS)

1987-12-01

This Conceptual Design Report describes the conceptual design of a high-level nuclear waste repository in salt at a proposed site in Deaf Smith County, Texas. Waste receipt, processing, packaging, and other surface facility operations are described. Operations in the shafts and underground are described, including waste hoisting, transfer, and horizontal emplacement. This report specifically addresses the horizontal emplacement mode, the passive alternate design for the repository. Waste retrieval capability is described. The report includes a description of the layout of the surface, shafts, and underground. Major equipment items are identified. The report includes plans for decommissioning and sealing of the facility. The report discusses how the repository will satisfy performance objectives. Chapters are included on basis for design, design analyses, and data requirements for completion of future design efforts. 105 figs., 52 tabs

Salt disposal of heat-generating nuclear waste

International Nuclear Information System (INIS)

Leigh, Christi D.; Hansen, Francis D.

2011-01-01

This report summarizes the state of salt repository science, reviews many of the technical issues pertaining to disposal of heat-generating nuclear waste in salt, and proposes several avenues for future science-based activities to further the technical basis for disposal in salt. There are extensive salt formations in the forty-eight contiguous states, and many of them may be worthy of consideration for nuclear waste disposal. The United States has extensive experience in salt repository sciences, including an operating facility for disposal of transuranic wastes. The scientific background for salt disposal including laboratory and field tests at ambient and elevated temperature, principles of salt behavior, potential for fracture damage and its mitigation, seal systems, chemical conditions, advanced modeling capabilities and near-future developments, performance assessment processes, and international collaboration are all discussed. The discussion of salt disposal issues is brought current, including a summary of recent international workshops dedicated to high-level waste disposal in salt. Lessons learned from Sandia National Laboratories' experience on the Waste Isolation Pilot Plant and the Yucca Mountain Project as well as related salt experience with the Strategic Petroleum Reserve are applied in this assessment. Disposal of heat-generating nuclear waste in a suitable salt formation is attractive because the material is essentially impermeable, self-sealing, and thermally conductive. Conditions are chemically beneficial, and a significant experience base exists in understanding this environment. Within the period of institutional control, overburden pressure will seal fractures and provide a repository setting that limits radionuclide movement. A salt repository could potentially achieve total containment, with no releases to the environment in undisturbed scenarios for as long as the region is geologically stable. Much of the experience gained from United

Salt disposal of heat-generating nuclear waste.

Energy Technology Data Exchange (ETDEWEB)

Leigh, Christi D. (Sandia National Laboratories, Carlsbad, NM); Hansen, Francis D.

2011-01-01

This report summarizes the state of salt repository science, reviews many of the technical issues pertaining to disposal of heat-generating nuclear waste in salt, and proposes several avenues for future science-based activities to further the technical basis for disposal in salt. There are extensive salt formations in the forty-eight contiguous states, and many of them may be worthy of consideration for nuclear waste disposal. The United States has extensive experience in salt repository sciences, including an operating facility for disposal of transuranic wastes. The scientific background for salt disposal including laboratory and field tests at ambient and elevated temperature, principles of salt behavior, potential for fracture damage and its mitigation, seal systems, chemical conditions, advanced modeling capabilities and near-future developments, performance assessment processes, and international collaboration are all discussed. The discussion of salt disposal issues is brought current, including a summary of recent international workshops dedicated to high-level waste disposal in salt. Lessons learned from Sandia National Laboratories' experience on the Waste Isolation Pilot Plant and the Yucca Mountain Project as well as related salt experience with the Strategic Petroleum Reserve are applied in this assessment. Disposal of heat-generating nuclear waste in a suitable salt formation is attractive because the material is essentially impermeable, self-sealing, and thermally conductive. Conditions are chemically beneficial, and a significant experience base exists in understanding this environment. Within the period of institutional control, overburden pressure will seal fractures and provide a repository setting that limits radionuclide movement. A salt repository could potentially achieve total containment, with no releases to the environment in undisturbed scenarios for as long as the region is geologically stable. Much of the experience gained from

Effects of waste content of glass waste forms on Savannah River high-level waste disposal costs

International Nuclear Information System (INIS)

McDonell, W.R.; Jantzen, C.M.

1985-01-01

Effects of the waste content of glass waste forms of Savannah River high-level waste disposal costs are evaluated by their impact on the number of waste canisters produced. Changes in waste content affect onsite Defense Waste Processing Facility (DWPF) costs as well as offsite shipping and repository emplacement charges. A nominal 1% increase over the 28 wt % waste loading of DWPF glass would reduce disposal costs by about $50 million for Savannah River wastes generated to the year 2000. Waste form modifications under current study include adjustments of glass frit content to compensate for added salt decontamination residues and increased sludge loadings in the DWPF glass. Projected cost reductions demonstrate significant incentives for continued optimization of the glass waste loadings. 13 refs., 3 figs., 3 tabs

Disposal of high-level waste from nuclear power plants in Denmark. v.3

International Nuclear Information System (INIS)

1981-01-01

The present report deals with material testing as part of the investigations regarding a repository for high-level waste in a salt dome. It is volume 3 of five volumes that together constitute the final report of the Danish utilities' salt dome investigations. The material testing programme chiefly comprised laboratory investigations and analyses of material samples, partly obtained from the salt dome in question and partly from the overlying geological formation. The test programme was fully completed for Eslev salt dome on Mors. Chemical analyses and mineralogical investigations were carried out in order to determine the type and extent of the impurities confined in the rock salt. Moreover, a programme was carried out to investigate the water content of the salt. Temperature conditions around a repository, the strenght and creep properties of the salt were investigated in order to determine whether drill holes, shafts and mine galleries could be constructed and maintained intact within the period of time required to complete the disposal of high-level waste. Chemical analysis were carried out in order to determine which substances are found dissolved in the water contained by the geological formation overlying the salt dome, as well as chemical/physical investigations regarding the water content, porosity, velocity of a possible leak of radioactive waste products, etc. Materials that would be introduced into a repository were studied with regard to their corrosion resistance in the salt environment. Concrete materials were investigated and characterized for their use in the final sealing of the access routes to the repository through the geological formation. (BP)

The Canadian program for management of spent fuel and high level wastes

International Nuclear Information System (INIS)

Barnes, R.W.; Mayman, S.A.

A brief history and description of the nuclear power program in Canada is given. Schedules and programs are described for storing spent fuel in station fuel bays, centralized water pool storage facilities, concrete canisters, convection vaults, and rock or salt formations. High-level wastes will be retrievable initially, therefore the focus is on storage in mined cavities. The methods developed for high-level waste storage/disposal will ideally be flexible enough to accommodate spent fuel. (E.C.B.)

Permian salt dissolution, alkaline lake basins, and nuclear-waste storage, Southern High Plains, Texas and New Mexico

International Nuclear Information System (INIS)

Reeves, C.C. Jr.; Temple, J.M.

1986-01-01

Areas of Permian salt dissolution associated with 15 large alkaline lake basins on and adjacent to the Southern High Plains of west Texas and eastern New Mexico suggest formation of the basins by collapse of strata over the dissolution cavities. However, data from 6 other alkaline basins reveal no evidence of underlying salt dissolution. Thus, whether the basins were initiated by subsidence over the salt dissolution areas or whether the salt dissolution was caused by infiltration of overlying lake water is conjectural. However, the fact that the lacustrine fill in Mound Lake greatly exceeds the amount of salt dissolution and subsidence of overlying beds indicates that at least Mound Lake basin was antecedent to the salt dissolution. The association of topography, structure, and dissolution in areas well removed from zones of shallow burial emphasizes the susceptibility of Permian salt-bed dissolution throughout the west Texas-eastern New Mexico area. Such evidence, combined with previous studies documenting salt-bed dissolution in areas surrounding a proposed high-level nuclear-waste repository site in Deaf Smith County, Texas, leads to serious questions about the rationale of using salt beds for nuclear-waste storage

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

International Nuclear Information System (INIS)

Jones, V.D.

1997-01-01

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

Proposal for basic safety requirements regarding the disposal of high-level radioactive waste

International Nuclear Information System (INIS)

1980-04-01

A working group commissioned to prepare proposals for basic safety requirements for the storage and transport of radioactive waste prepared its report to the Danish Agency of Environmental Protection. The proposals include: radiation protection requirements, requirements concerning the properties of high-level waste units, the geological conditions of the waste disposal location, the supervision of waste disposal areas. The proposed primary requirements for safety evaluation of the disposal of high-level waste in deep geological formations are of a general nature, not being tied to specific assumptions regarding the waste itself, the geological and other conditions at the place of disposal, and the technical methods of disposal. It was impossible to test the proposals for requirements on a working repository. As no country has, to the knowledge of the working group, actually disposed of hifg-level radioactive waste or approved of plans for such disposal. Methods for evaluating the suitability of geological formations for waste disposal, and background material concerning the preparation of these proposals for basic safety requirements relating to radiation, waste handling and geological conditions are reviewed. Appended to the report is a description of the phases of the fuel cycle that are related to the storage of spent fuel and the disposal of high-level reprocessing waste in a salt formation. It should be noted that the proposals of the working group are not limited to the disposal of reprocessed fuel, but also include the direct disposal of spent fuel as well as disposal in geological formations other than salt. (EG)

Other-than-high-level waste

International Nuclear Information System (INIS)

Bray, G.R.

1976-01-01

The main emphasis of the work in the area of partitioning transuranic elements from waste has been in the area of high-level liquid waste. But there are ''other-than-high-level wastes'' generated by the back end of the nuclear fuel cycle that are both large in volume and contaminated with significant quantities of transuranic elements. The combined volume of these other wastes is approximately 50 times that of the solidified high-level waste. These other wastes also contain up to 75% of the transuranic elements associated with waste generated by the back end of the fuel cycle. Therefore, any detailed evaluation of partitioning as a viable waste management option must address both high-level wastes and ''other-than-high-level wastes.''

Waste isolation facility description: bedded salt

Energy Technology Data Exchange (ETDEWEB)

1976-09-01

The waste isolation facility is designed to receive and store three basic types of solidified wastes: high-level wastes, intermediate level high-gamma transuranic waste, and low-gamma transuranic wastes. The facility under consideration in this report is designed for bedded salt at a depth of approximately 1800 ft. The present design for the facility includes an area which would be used initially as a pilot facility to test the viability of the concept, and a larger facility which would constitute the final storage area. The total storage area in the pilot facility is planned to be 77 acres and in the fuel facility 1601 acres. Other areas for shaft operations and access would raise the overall size of the total facility to slightly less than 2,000 acres. The following subjects are discussed in detail: surface facilities, shaft design and characteristics, design and construction of the underground waste isolation facility, ventilation systems, and design requirements and criteria. (LK)

Waste isolation facility description: bedded salt

International Nuclear Information System (INIS)

1976-09-01

The waste isolation facility is designed to receive and store three basic types of solidified wastes: high-level wastes, intermediate level high-gamma transuranic waste, and low-gamma transuranic wastes. The facility under consideration in this report is designed for bedded salt at a depth of approximately 1800 ft. The present design for the facility includes an area which would be used initially as a pilot facility to test the viability of the concept, and a larger facility which would constitute the final storage area. The total storage area in the pilot facility is planned to be 77 acres and in the fuel facility 1601 acres. Other areas for shaft operations and access would raise the overall size of the total facility to slightly less than 2,000 acres. The following subjects are discussed in detail: surface facilities, shaft design and characteristics, design and construction of the underground waste isolation facility, ventilation systems, and design requirements and criteria

Mixed Waste Salt Encapsulation Using Polysiloxane - Final Report

International Nuclear Information System (INIS)

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

1997-01-01

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

Atmospheric Pressure Effect of Retained Gas in High Level Waste

International Nuclear Information System (INIS)

Weber, A.H.

1999-01-01

Isolated high level waste tanks in H-Area have unexplained changes in waste-level which have been attributed to environmental effects including pressure, temperature, and relative humidity. Previous studies at SRS have considered waste-level changes from causes not including the presence of gas in the salt cake. This study was undertaken to determine the effect of atmospheric pressure on gas in the salt cake and resultant changes in the supernate level of Tank 41H, and to model that effect if possible. A simple theory has been developed to account for changes in the supernate level in a high level waste tank containing damp salt cake as the response of trapped gases to changes in the ambient pressure. The gas is modeled as an ideal gas retained as bubbles within the interstitial spaces in the salt cake and distributed uniformly throughout the tank. The model does not account for consistent long term increases or decreases in the tank level. Any such trend in the tank level is attributed to changes in the liquid content in the tank (from condensation, evaporation, etc.) and is removed from the data prior to the void estimation. Short term fluctuations in the tank level are explained as the response of the entrained gas volume to changes in the ambient pressure. The model uses the response of the tank level to pressure changes to estimate an average void fraction for the time period of interest. This estimate of the void is then used to predict the expected level response. The theory was applied to three separate time periods of the level data for tank 41H as follows: (1) May 3, 1993 through August 3, 1993, (2) January 23, 1994 through April 21, 1994, and (3) June 4, 1994 through August 24, 1994. A strong correlation was found between fluctuations in the tank level and variations in the ambient pressure. This correlation is a clear marker of the presence of entrained gases in the tank. From model calculations, an average void fraction of 11 percent was estimated to

High-level waste processing and disposal

International Nuclear Information System (INIS)

Crandall, J.L.; Krause, H.; Sombret, C.; Uematsu, K.

1984-11-01

Without reprocessing, spent LWR fuel itself is generally considered an acceptable waste form. With reprocessing, borosilicate glass canisters, have now gained general acceptance for waste immobilization. The current first choice for disposal is emplacement in an engineered structure in a mined cavern at a depth of 500-1000 meters. A variety of rock types are being investigated including basalt, clay, granite, salt, shale, and volcanic tuff. This paper gives specific coverage to the national high level waste disposal plans for France, the Federal Republic of Germany, Japan and the United States. The French nuclear program assumes prompt reprocessing of its spent fuels, and France has already constructed the AVM. Two larger borosilicate glass plants are planned for a new French reprocessing plant at La Hague. France plans to hold the glass canisters in near-surface storage for a forty to sixty year cooling period and then to place them into a mined repository. The FRG and Japan also plan reprocessing for their LWR fuels. Both are currently having some fuel reprocessed by France, but both are also planning reprocessing plants which will include waste vitrification facilities. West Germany is now constructing the PAMELA Plant at Mol, Belgium to vitrify high level reprocessing wastes at the shutdown Eurochemic Plant. Japan is now operating a vitrification mockup test facility and plans a pilot plant facility at the Tokai reprocessing plant by 1990. Both countries have active geologic repository programs. The United State program assumes little LWR fuel reprocessing and is thus primarily aimed at direct disposal of spent fuel into mined repositories. However, the US have two borosilicate glass plants under construction to vitrify existing reprocessing wastes

Repository environmental parameters and models/methodologies relevant to assessing the performance of high-level waste packages in basalt, tuff, and salt

Energy Technology Data Exchange (ETDEWEB)

Claiborne, H.C.; Croff, A.G.; Griess, J.C.; Smith, F.J.

1987-09-01

This document provides specifications for models/methodologies that could be employed in determining postclosure repository environmental parameters relevant to the performance of high-level waste packages for the Basalt Waste Isolation Project (BWIP) at Richland, Washington, the tuff at Yucca Mountain by the Nevada Test Site, and the bedded salt in Deaf Smith County, Texas. Guidance is provided on the identify of the relevant repository environmental parameters; the models/methodologies employed to determine the parameters, and the input data base for the models/methodologies. Supporting studies included are an analysis of potential waste package failure modes leading to identification of the relevant repository environmental parameters, an evaluation of the credible range of the repository environmental parameters, and a summary of the review of existing models/methodologies currently employed in determining repository environmental parameters relevant to waste package performance. 327 refs., 26 figs., 19 tabs.

Repository environmental parameters and models/methodologies relevant to assessing the performance of high-level waste packages in basalt, tuff, and salt

International Nuclear Information System (INIS)

Claiborne, H.C.; Croff, A.G.; Griess, J.C.; Smith, F.J.

1987-09-01

This document provides specifications for models/methodologies that could be employed in determining postclosure repository environmental parameters relevant to the performance of high-level waste packages for the Basalt Waste Isolation Project (BWIP) at Richland, Washington, the tuff at Yucca Mountain by the Nevada Test Site, and the bedded salt in Deaf Smith County, Texas. Guidance is provided on the identify of the relevant repository environmental parameters; the models/methodologies employed to determine the parameters, and the input data base for the models/methodologies. Supporting studies included are an analysis of potential waste package failure modes leading to identification of the relevant repository environmental parameters, an evaluation of the credible range of the repository environmental parameters, and a summary of the review of existing models/methodologies currently employed in determining repository environmental parameters relevant to waste package performance. 327 refs., 26 figs., 19 tabs

High-level waste canister envelope study: structural analysis

International Nuclear Information System (INIS)

1977-11-01

The structural integrity of waste canisters, fabricated from standard weight Type 304L stainless steel pipe, was analyzed for sizes ranging from 8 to 24 in. diameter and 10 to 16 feet long under normal, abnormal, and improbable life cycle loading conditions. The canisters are assumed to be filled with vitrified high-level nuclear waste, stored temporarily at a fuel reprocessing plant, and then transported for storage in an underground salt bed or other geologic storage. In each of the three impact conditions studies, the resulting impact force is far greater than the elastic limit capacity of the material. Recommendations are made for further study

Waste salt disposal at the Savannah River Plant

International Nuclear Information System (INIS)

Langton, C.A.; Oblath, S.B.; Pepper, D.W.; Wilhite, E.L.

1986-01-01

Waste salt solution, produced during processing of high-level nuclear waste, will be incorporated in a cement matrix for emplacement in an engineered disposal facility. Wasteform characteristics and disposal facility details will be presented along with results of a field test of wasteform contaminant release and of modeling studies to predict releases. 5 refs., 11 figs., 5 tabs

Alternative processes for managing existing commercial high-level radioactive wastes

International Nuclear Information System (INIS)

1976-04-01

A number of alternatives are discussed for managing high-level radioactive waste presently stored at the West Valley, New York, plant owned by Nuclear Fuel Services, Inc. These alternatives (liquid storage, conversion to cement, shale fracturing, shale cement, calcination, aqueous silicate, conversion to glass, and salt cake) are limited to concepts presently under active investigation by ERDA. Each waste management option is described and examined regarding the status of the technology; its applications to managing NFS waste; its advantages and disadvantages; the research and development needed to implement the option; safety considerations; and estimated costs and time to implement the process

Testing of high-level waste forms under repository conditions

International Nuclear Information System (INIS)

Mc Menamin, T.

1989-01-01

The workshop on testing of high-level waste forms under repository conditions was held on 17 to 21 October 1988 in Cadarache, France, and sponsored by the Commission of the European Communities (CEC), the Commissariat a l'energie atomique (CEA) and the Savannah River Laboratory (US DOE). Participants included representatives from Australia, Belgium, Denmark, France, Germany, Italy, Japan, the Netherlands, Sweden, Switzerland, The United Kingdom and the United States. The first part of the conference featured a workshop on in situ testing of simulated nuclear waste forms and proposed package components, with an emphasis on the materials interface interactions tests (MIIT). MIIT is a sevent-part programme that involves field testing of 15 glass and waste form systems supplied by seven countries, along with potential canister and overpack materials as well as geologic samples, in the salt geology at the Waste Isolation Pilot Plant (WIPP) in Carlsbad, New Mexico, USA. This effort is still in progress and these proceedings document studies and findings obtained thus far. The second part of the meeting emphasized multinational experimental studies and results derived from repository systems simulation tests (RSST), which were performed in granite, clay and salt environments

Salt Repository Project waste emplacement mode decision paper: Revison 1

International Nuclear Information System (INIS)

1987-08-01

This paper provides a recommendation as to the mode of waste emplacement to be used as the current basis for site characterization activity for the Deaf Smith County, Texas, high level nuclear waste repository site. It also presents a plan for implementing the recommendation so as to provide a high level of confidence in the project's success. Since evaluations of high-level waste disposal in geologic repositories began in the 1950s, most studies emplacement in salt formations employed the vertical orientation for emplacing waste packages in boreholes in the floor of the underground facility. This orientation was used in trials at Project Salt Vault in the 1960s. The Waste Isolation Pilot Plant (WIPP) has recently settled on a combination of vertical and horizontal modes for various waste types. This paper analyzes the information available and develops a project position upon which to base current site characterization activities. The position recommended is that the SRP should continue to use the vertical waste emplacement mode as the reference design and to carry the horizontal mode as a ''passive'' alternative. This position was developed based upon the conclusions of a decision analysis, risk assessment, and cost/schedule impact assessment. 52 refs., 6 figs., 1 tab

Salt Repository Project Waste Package Program Plan: Draft

International Nuclear Information System (INIS)

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

1986-01-01

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

Comments on a letter by George D. DeBuchananne (US Geological Survey) regarding the use of salt domes for high-level waste disposal

International Nuclear Information System (INIS)

1984-08-01

The US Geological Survey (USGS) concluded in a letter to the US Department of Energy, dated March 7, 1981, that subsurface geologic conditions in bedded salt are more predictable and less complex than those in domal salt. This predictability is equated with the relative suitability of bedded and domal salt as repository host media. This report comments on the USGS letter. The key points made are as follows: Complexities which may exist in the geologic setting of a salt dome (or other potential host medium) should not a priori preclude the dome from being an acceptable host medium for a high-level waste (HLW) repository. Predictability, as used by the USGS, focused on the spatial extrapolation of information on geologic conditions and should not be confused with predicting the performance of a repository. Notwithstanding the general characteristics of bedded and domal salt, there are salt domes whose individual characteristics should make them as acceptable as potential bedded salt areas for HLW repository sites. Complexities which may occur in the geologic setting of a salt dome can be explored and characterized with sufficient accuracy by available techniques

Stabilization of high-level waste from a chloride volatility nuclear fuel reprocessing system

International Nuclear Information System (INIS)

Smith, L.A.; Thornton, T.A.

1979-01-01

Methods for stabilizing high-level waste from a chloride volatility thorium-based fuel coprocessing system have been studied. The waste, which is present as chloride salts, is combined with SiO 2 or Al 2 O 3 and pyrohydrolyzed to remove the chloride ions. The resulting solid is then combined with a flux and glassified. 3 figures, 4 tables

The management of high-level radioactive wastes

International Nuclear Information System (INIS)

Lennemann, Wm.L.

1979-01-01

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

Performance of high level waste forms and engineered barriers under repository conditions

International Nuclear Information System (INIS)

1991-02-01

The IAEA initiated in 1977 a co-ordinated research programme on the ''Evaluation of Solidified High-Level Waste Forms'' which was terminated in 1983. As there was a continuing need for international collaboration in research on solidified high-level waste form and spent fuel, the IAEA initiated a new programme in 1984. The new programme, besides including spent fuel and SYNROC, also placed greater emphasis on the effect of the engineered barriers of future repositories on the properties of the waste form. These engineered barriers included containers, overpacks, buffer and backfill materials etc. as components of the ''near-field'' of the repository. The Co-ordinated Research Programme on the Performance of High-Level Waste Forms and Engineered Barriers Under Repository Conditions had the objectives of promoting the exchange of information on the experience gained by different Member States in experimental performance data and technical model evaluation of solidified high level waste forms, components of the waste package and the complete waste management system under conditions relevant to final repository disposal. The programme includes studies on both irradiated spent fuel and glass and ceramic forms as the final solidified waste forms. The following topics were discussed: Leaching of vitrified high-level wastes, modelling of glass behaviour in clay, salt and granite repositories, environmental impacts of radionuclide release, synroc use for high--level waste solidification, leachate-rock interactions, spent fuel disposal in deep geologic repositories and radionuclide release mechanisms from various fuel types, radiolysis and selective leaching correlated with matrix alteration. Refs, figs and tabs

Performance analysis of conceptual waste package designs in salt repositories

International Nuclear Information System (INIS)

Jansen, G. Jr.; Raines, G.E.; Kircher, J.F.

1984-01-01

A performance analysis of commercial high-level waste and spent fuel conceptual package designs in reference repositories in three salt formations was conducted with the WAPPA waste package code. Expected conditions for temperature, stress, brine composition, radiation level, and brine flow rate were used as boundary conditions to compute expected corrosion of a thick-walled overpack of 1025 wrought steel. In all salt formations corrosion by low Mg salt-dissolution brines typical of intrusion scenarios was too slow to cause the package to fail for thousands of years after burial. In high Mg brines judged typical of thermally migrating brines in bedded salt formations, corrosion rates which would otherwise have caused the packages to fail within a few hundred years were limited by brine availability. All of the brine reaching the package was consumed by reaction with the iron in the overpack, thus preventing further corrosion. Uniform brine distribution over the package surface was an important factor in predicting long package lifetimes for the high Mg brines. 14 references, 15 figures

Potential host media for a high-level waste repository

Energy Technology Data Exchange (ETDEWEB)

Hustrulid, W

1982-01-01

Earlier studies of burial of radioactive wastes in geologic repositories had concentrated on salt formations for well-publicized reasons. However, under the Carter administration, significant changes were made in the US nuclear waste management program. Changes which were made were: (1) expansion of the number of rock types under consideration; (2) adoption of the multiple-barrier approach to waste containment; (3) additional requirements for waste retrieval; and (4) new criteria proposed by the Nuclear Regulatory Commission for the isolation of high-level waste in geologic repositories. Results of the studies of different types of rocks as repository sites are summarized herein. It is concluded that each generic rock type has certain advantages and disadvantages when considered from various aspects of the waste disposal problem and that characteristics of rocks are so varied that a most favorable or least favorable rock type cannot be easily identified. This lack of definitive characteristics of rocks makes site selection and good engineering barriers very important for containment of the wastes. (BLM)

Comparison of SRP high-level waste disposal costs for borosilicate glass and crystalline ceramic waste forms

International Nuclear Information System (INIS)

McDonell, W.R.

1982-04-01

An evaluation of costs for the immobilization and repository disposal of SRP high-level wastes indicates that the borosilicate glass waste form is less costly than the crystalline ceramic waste form. The wastes were assumed immobilized as glass with 28% waste loading in 10,300 reference 24-in.-diameter canisters or as crystalline ceramic with 65% waste loading in either 3400 24-in.-diameter canisters or 5900 18-in.-diameter canisters. After an interim period of onsite storage, the canisters would be transported to the federal repository for burial. Total costs in undiscounted 1981 dollars of the waste disposal operations, excluding salt processing for which costs are not yet well defined, were about $2500 million for the borosilicate glass form in reference 24-in.-diameter canisters, compared to about $2900 million for the crystalline ceramic form in 24-in.-diameter canisters and about $3100 million for the crystalline ceramic form in 18-in.-diameter canisters. No large differences in salt processing costs for the borosilicate glass and crystalline ceramic forms are expected. Discounting to present values, because of a projected 2-year delay in startup of the DWPF for the crystalline ceramic form, preserved the overall cost advantage of the borosilicate glass form. The waste immobilization operations for the glass form were much less costly than for the crystalline ceramic form. The waste disposal operations, in contrast, were less costly for the crystalline ceramic form, due to fewer canisters requiring disposal; however, this advantage was not sufficient to offset the higher development and processing costs of the crystalline ceramic form. Changes in proposed Nuclear Regulatory Commission regulations to permit lower cost repository packages for defense high-level wastes would decrease the waste disposal costs of the more numerous borosilicate glass forms relative to the crystalline ceramic forms

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

International Nuclear Information System (INIS)

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

1990-04-01

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

Criticality considerations for salt-cake disolution in DOE waste tanks

International Nuclear Information System (INIS)

Trumble, E.F.; Niemer, K.A.

1995-01-01

A large amount of high-level waste is being stored in the form of salt cake at the Savannah River site (SRS) in large (1.3 x 106 gal) underground tanks awaiting startup of the Defense Waste Processing Facility (DWPF). This salt cake will be dissolved with water, and the solution will be fed to DWPF for immobilization in borosilicate glass. Some of the waste that was transferred to the tanks contained enriched uranium and plutonium from chemical reprocessing streams. As water is added to these tanks to dissolve the salt cake, the insoluble portion of this fissile material will be left behind in the tank as the salt solution is pumped out. Because the salt acts as a diluent to the fissile material, the process of repeated water addition, salt dissolution, and salt solution removal will act as a concentrating mechanism for the undissolved fissile material that will remain in the tank. It is estimated that tank 41 H at SRS contains 20 to 120 kg of enriched uranium, varying from 10 to 70% 235 U, distributed nonuniformly throughout the tank. This paper discusses the criticality concerns associated with the dissolution of salt cake in this tank. These concerns are also applicable to other salt cake waste tanks that contain significant quantities of enriched uranium and/or plutonium

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

International Nuclear Information System (INIS)

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

1985-07-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1985-07-01

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

Stabilization Using Phosphate Bonded Ceramics. Salt Containing Mixed Waste Treatment. Mixed Waste Focus Area. OST Reference No. 117

International Nuclear Information System (INIS)

1999-01-01

Throughout the Department of Energy (DOE) complex there are large inventories of homogeneous mixed waste solids, such as wastewater treatment residues, fly ashes, and sludges that contain relatively high concentrations (greater than 15% by weight) of salts. The inherent solubility of salts (e.g., nitrates, chlorides, and sulfates) makes traditional treatment of these waste streams difficult, expensive, and challenging. One alternative is low-temperature stabilization by chemically bonded phosphate ceramics (CBPCs). The process involves reacting magnesium oxide with monopotassium phosphate with the salt waste to produce a dense monolith. The ceramic makes a strong environmental barrier, and the metals are converted to insoluble, low-leaching phosphate salts. The process has been tested on a variety of surrogates and actual mixed waste streams, including soils, wastewater, flyashes, and crushed debris. It has also been demonstrated at scales ranging from 5 to 55 gallons. In some applications, the CBPC technology provides higher waste loadings and a more durable salt waste form than the baseline method of cementitious grouting. Waste form test specimens were subjected to a variety of performance tests. Results of waste form performance testing concluded that CBPC forms made with salt wastes meet or exceed both RCRA and recommended Nuclear Regulatory Commission (NRC) low-level waste (LLW) disposal criteria. Application of a polymer coating to the CBPC may decrease the leaching of salt anions, but continued waste form evaluations are needed to fully assess the deteriorating effects of this leaching, if any, over time.

Evaluation of process alternatives for solidification of the West Valley high-level liquid wastes

International Nuclear Information System (INIS)

Holton, L.K.; Larson, D.E.

1982-01-01

The Department of Energy (DOE) established the West Valley Solidification Project (WVSP) in 1980. The project purpose is to demonstrate removal and solidification of the high-level liquid wastes (HLLW) presently stored in tanks at the Western New York Nuclear Service Center (WNYNSC), West Valley, New York. As part of this effort, the Pacific Northwest Laboratory (PNL) conducted a study to evaluate process alternatives for solidifcation of the WNYNSC wastes. Two process approaches for waste handling before solidification, together with solidification processes for four terminal and four interim waste forms, were considered. The first waste-handling approach, designated the salt/sludge separation process, involves separating the bulk of the nonradioactive nuclear waste constituents from the radioactive waste constituents, and the second waste-handling approach, designated the combined-waste process, involves no waste segregation prior to solidification. The processes were evaluated on the bases of their (1) readiness for plant startup by 1987, (2) relative technical merits, and (3) process cost. The study has shown that, based on these criteria, the salt/sludge separation process with a borosilicate glass waste form is preferred when producing a terminal waste form. It was also concluded that if an interim waste form is to be used, the preferred approach would be the combined waste process with a fused-salt waste form

Electropolishing decontamination system for high-level waste canisters

International Nuclear Information System (INIS)

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

1988-10-01

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

Salton Sea geothermal field as a natural analog for the near-field in a salt high-level nuclear waste repository

International Nuclear Information System (INIS)

Elders, W.A.; Moody, J.B.; Battelle Memorial Inst., Columbus, OH)

1984-01-01

The Salton Sea Geothermal Field (SSGF), on the delta of the Colorado River in southern California, is being studied as a natural analog for the near-field environment of proposed nuclear waste repositories in salt. A combination of mineralogical and geochemical methods is being employed to develop a three-dimenisonal picture of temperature, salinity, lithology, mineralogy, and chemistry of reactions between the reservoir rocks and the hot brines. Our aim is to obtain quantitative data on mineral stabilities and on mobilities of the naturally occurring radionuclides of concern in Commercial High-Level Waste (CHLW). These data will be used to validate the EQ3/6 geochemical code under development to model the salt near-field repository behavior. Maximum temperatures encountered in wells in the SSGF equal or exceed peak temperatures expected in a salt repository. Brines produced from these wells have major element chemistry similar to brines from candidate salt sites. Relative to the rocks, these brines are enriched in Na, Mn, Sr, Ra, and Po, depleted in Ba, Si, Mg, Ti, and Al, and strongly depleted in U and Th. However, the unaltered rocks contain only about 2 to 3 ppm of U and 4 to 12 ppm of Th, largely in detrital epidotes and zircons. Samples of hydrothermally altered rocks from a wide range of temperature and salinity show rather similar uniform low concentrations of these elements, even when authigenic illite, chlorite, ipidote and feldspar are present. These observations suggest that U and Th are relatively immobile in these hot brines. However, Ra, Po, Cs, and Sr are relatively mobile. Work is continuing to document naturally occurring radionuclide partitioning between SSGF minears and brine over a range of temperature, salinity, and lithology. 8 refs., 7 figs., 2 tabs

Heat transfer analysis of the geologic disposal of spent fuel and high-level waste storage canisters

International Nuclear Information System (INIS)

Allen, G.K.

1980-08-01

Near-field temperatures resulting from the storage of high-level waste canisters and spent unreprocessed fuel assembly canisters in geologic formations were determined. Preliminary design of the repository was modeled for a heat transfer computer code, HEATING5, which used the Crank-Nicolson finite difference method to evaluate transient heat transfer. The heat transfer system was evaluated with several two- and three-dimensional models which transfer heat by a combination of conduction, natural convention, and radiation. Physical properties of the materials in the model were based upon experimental values for the various geologic formations. The effects of canister spacing, fuel age, and use of an overpack were studied for the analysis of the spent fuel canisters; salt, granite, and basalt were considered as the storage media for spent fuel canisters. The effects of canister diameter and use of an overpack were studied for the analysis of the high-level waste canisters; salt was considered as the only storage media for high-level waste canisters. Results of the studies on spent fuel assembly canisters showed that the canisters could be stored in salt formations with a maximum heat loading of 134 kw/acre without exceeding the temperature limits set for salt stability. The use of an overpack had little effect on the peak canister temperatures. When the total heat load per acre decreased, the peak temperatures reached in the geologic formations decreased; however, the time to reach the peak temperatures increased. Results of the studies on high-level waste canisters showed that an increased canister diameter will increase the canister interior temperatures considerably; at a constant areal heat loading, a 381 mm diameter canister reached almost a 50 0 C higher temperature than a 305 mm diameter canister. An overpacked canister caused almost a 30 0 C temperature rise in either case

Minor component study for simulated high-level nuclear waste glasses (Draft)

International Nuclear Information System (INIS)

Li, H.; Langowskim, M.H.; Hrma, P.R.; Schweiger, M.J.; Vienna, J.D.; Smith, D.E.

1996-02-01

Hanford Site single-shell tank (SSI) and double-shell tank (DSI) wastes are planned to be separated into low activity (or low-level waste, LLW) and high activity (or high-level waste, HLW) fractions, and to be vitrified for disposal. Formulation of HLW glass must comply with glass processibility and durability requirements, including constraints on melt viscosity, electrical conductivity, liquidus temperature, tendency for phase segregation on the molten glass surface, and chemical durability of the final waste form. A wide variety of HLW compositions are expected to be vitrified. In addition these wastes will likely vary in composition from current estimates. High concentrations of certain troublesome components, such as sulfate, phosphate, and chrome, raise concerns about their potential hinderance to the waste vitrification process. For example, phosphate segregation in the cold cap (the layer of feed on top of the glass melt) in a Joule-heated melter may inhibit the melting process (Bunnell, 1988). This has been reported during a pilot-scale ceramic melter run, PSCM-19, (Perez, 1985). Molten salt segregation of either sulfate or chromate is also hazardous to the waste vitrification process. Excessive (Cr, Fe, Mn, Ni) spinel crystal formation in molten glass can also be detrimental to melter operation

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

International Nuclear Information System (INIS)

Chaturvedi, L.

1985-01-01

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

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

International Nuclear Information System (INIS)

1986-01-01

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

Concepts and Technologies for Radioactive Waste Disposal in Rock Salt

Directory of Open Access Journals (Sweden)

Wernt Brewitz

2007-01-01

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

Designing Advanced Ceramic Waste Forms for Electrochemical Processing Salt Waste

International Nuclear Information System (INIS)

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

2016-01-01

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

Designing Advanced Ceramic Waste Forms for Electrochemical Processing Salt Waste

Energy Technology Data Exchange (ETDEWEB)

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

2016-03-01

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

High-Level Radioactive Waste.

Science.gov (United States)

Hayden, Howard C.

1995-01-01

Presents a method to calculate the amount of high-level radioactive waste by taking into consideration the following factors: the fission process that yields the waste, identification of the waste, the energy required to run a 1-GWe plant for one year, and the uranium mass required to produce that energy. Briefly discusses waste disposal and…

New high-level waste management technology for IFR pyroprocessing wastes

International Nuclear Information System (INIS)

Ackerman, J.P.; Johnson, T.R.

1993-01-01

The pyrochemical electrorefining process for recovery of actinides in spent fuel from the Integral Fast Reactor accumulates fission product wastes as chlorides dissolved in molten LiCI-KCI and as metals, some of which are in molten cadmium. Pyrochemical processes are being developed to recover uranium and transuranium elements for return to the reactor, and to separate and immobilize fission products in suitable waste forms. Solvent cadmium is recycled within the process. Electrolyte salt is treated in a series of salt/cadmium extraction steps; it is also returned to the process. Salt-borne fission products are concentrated on a zeolite bed that is converted to a stable, leach-resistant mineral. Rare earth fission products from the salt, noble metal fission products, and cladding hulls are dispersed in a metal matrix

Selection of the host rock for high level radioactive waste repository in China

International Nuclear Information System (INIS)

Jin Yuanxin; Wang Wenguang; Chen Zhangru

2001-01-01

The authors has briefly introduced the experiences of the host rock selection and the host rock types in other countries for high level radioactive waste repository. The potential host rocks in China are investigated. They include granite, tuff, clay, basalt, salt, and loess. The report has expounded the distributions, scale, thickness, mineral and chemical composition, construction, petrogenesis and the ages of the rock. The possibility of these rocks as the host rock has been studied. The six pieces of distribution map of potential rocks have been made up. Through the synthetical study, it is considered that granite as the host rock of high level radioactive waste repository is possible

Hydrological methods preferentially recover cesium from nuclear waste salt cake

International Nuclear Information System (INIS)

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

1997-01-01

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

High Level Radioactive Waste Management

International Nuclear Information System (INIS)

1991-01-01

The proceedings of the second annual international conference on High Level Radioactive Waste Management, held on April 28--May 3, 1991, Las Vegas, Nevada, provides information on the current technical issue related to international high level radioactive waste management activities and how they relate to society as a whole. Besides discussing such technical topics as the best form of the waste, the integrity of storage containers, design and construction of a repository, the broader social aspects of these issues are explored in papers on such subjects as conformance to regulations, transportation safety, and public education. By providing this wider perspective of high level radioactive waste management, it becomes apparent that the various disciplines involved in this field are interrelated and that they should work to integrate their waste management activities. Individual records are processed separately for the data bases

West Valley demonstration project: alternative processes for solidifying the high-level wastes

International Nuclear Information System (INIS)

Holton, L.K.; Larson, D.E.; Partain, W.L.; Treat, R.L.

1981-10-01

In 1980, the US Department of Energy (DOE) established the West Valley Solidification Project as the result of legislation passed by the US Congress. The purpose of this project was to carry out a high level nuclear waste management demonstration project at the Western New York Nuclear Service Center in West Valley, New York. The DOE authorized the Pacific Northwest Laboratory (PNL), which is operated by Battelle Memorial Institute, to assess alternative processes for treatment and solidification of the WNYNSC high-level wastes. The Process Alternatives Study is the suject of this report. Two pretreatment approaches and several waste form processes were selected for evaluation in this study. The two waste treatment approaches were the salt/sludge separation process and the combined waste process. Both terminal and interim waste form processes were studied. The terminal waste form processes considered were: borosilicate glass, low-alkali glass, marbles-in-lead matrix, and crystallinolecular potential and molecular dynamics calculations of the effect are yet to be completed. Cous oxide was also investigated. The reaction is first order in nitrite ion, second order in hydrogen ion, and between zero and first order in hydroxylamine monosulfonate, depending on the concentration

Heat transfer in high-level waste management

International Nuclear Information System (INIS)

Dickey, B.R.; Hogg, G.W.

1979-01-01

Heat transfer in the storage of high-level liquid wastes, calcining of radioactive wastes, and storage of solidified wastes are discussed. Processing and storage experience at the Idaho Chemical Processing Plant are summarized for defense high-level wastes; heat transfer in power reactor high-level waste processing and storage is also discussed

Cerebral salt wasting following tuberculous meningoencephalitis in an infant

Directory of Open Access Journals (Sweden)

Syed Ahmed Zaki

2012-01-01

Full Text Available In patients with central nervous system disease, life-threatening hyponatremia can result from either the syndrome of inappropriate secretion of antidiuretic hormone or cerebral salt wasting. Clinical manifestations of the two conditions may be similar, but their pathogeneses and management protocols are different. Cerebral salt wasting syndrome is a disorder in which excessive natriuresis and hyponatremia occurs in patients with intracranial diseases. We report a 6-month-old girl with CSWS associated with tuberculous meningoencephalitis. She was diagnosed as having CSWS on the basis of hypovolemia, polyuria, natriuresis, and the relatively high level of fractional excretion of uric acid. Aggressive replacement of urine salt and water losses using 0.9% or 3% sodium chloride was done. Fludrocortisone was started at 0.1 mg twice daily on the seventh day of admission and was continued for 17 days.

Corrosion behaviour of selected high-level waste packaging materials under gamma irradiation and in-situ disposal conditions in rock salt

International Nuclear Information System (INIS)

Smailos, E.; Schwarzkopf, W.; Koester, R.

1988-07-01

Corrosion studies performed until now on a number of materials have shown that unalloyed steels, Hastelloy C4 and Ti 99.8-Pd are the most promising materials for a long-term resistant packaging to be used in high-level waste (HLW) canister disposal in rock salt formations. To characterize their corrosion behaviour in more detail, additional studies have been performed. The influence has been examined which is exerted by the gamma dose rate (1 Gy/h to 100 Gy/h) on the corrosion of three preselected steels and Hastelloy C4 at 90 0 C in a salt brine (Q-brine) rich in MgCl 2 , i.e., conditions relevant to accident scenarios in a repository. In addition, in-situ corrosion experiments have been carried out in the Asse salt mine at elevated temperatures (120 0 C to 210 0 C) in the absence and in the presence of a gamma radiation field of 3 x 10 2 Gy/h, within the framework of the German/US Brine Migration Test. Under the test conditions the gamma radiation did not exert a significant influence on the corrosion of the steels investigated, whereas Hastelloy C4, exposed to dose rates of 10 Gy/h and 100 Gy/h, underwent pitting and crevice corrosion (20 μm/a at the maximum).The low amounts of migrated salt brine (140 ml after 900 days) in the in-situ- experiment did not produce noticeable corrosion of the materials. (orig./RB) [de

High-level-waste immobilization

International Nuclear Information System (INIS)

Crandall, J.L.

1982-01-01

Analysis of risks, environmental effects, process feasibility, and costs for disposal of immobilized high-level wastes in geologic repositories indicates that the disposal system safety has a low sensitivity to the choice of the waste disposal form

Backfill barriers for nuclear waste repositories in salt

Energy Technology Data Exchange (ETDEWEB)

Nowak, E J; Odoj, R; Merz, E [eds.

1981-06-01

Backfill materials were evaluated for containment of radionuclides, chemical modification of brine, and sensitivity to hydrothermal conditions. Experimental conditions were relevant to nuclear waste isolation in bedded salt. They were based on geologic conditions at the site of the Waste Isolation Pilot Plant (WIPP) in southeastern New Mexico, USA. Conclusions are: backfill mixtures surrounding the waste form and canister can provide a neutral or slightly acidic, potentially reducing environment, prevent convective aqueous flow, and act as an effective radionuclide migration barrier; bentonite is likely to remain hydrothermally stable but potentially sensitive to waste package interactions which could alter the pH, the ratio of dissolved ions, or the sorption properties of radionuclide species; effects of irradiation from high level waste should be investigated.

Experimental results on salt concrete for barrier elements made of salt concrete in a repository for radioactive waste in a salt mine

International Nuclear Information System (INIS)

Gutsch, Alex-W.; Preuss, Juergen; Mauke, Ralf

2012-01-01

The Bartensleben rock salt mine in Germany was used as a repository for low and intermediate level radioactive waste from 1971 to 1991 and from 1994 to 1998. The repository with an overall volume of about 6 million m 3 has to be closed. Salt concrete is used for the refill of the voids of the repository. The concrete mixtures contain crushed salt instead of natural aggregates as the void filling material should be as similar to the salt rock as possible. Very high requirements regarding low heat development and little or even no cracking during concrete hardening had to be fulfilled even for the barrier elements made from salt concrete which separate the radioactive waste from the environment. Requirements for the salt concrete were set up with regard to the fluidity of the fresh concrete during the hardening process and its durability. In the view of a comprehensive numerical calculations of the temperature development and thermal stresses in the massive salt concrete elements of the backfill of the voids, experimental results for material properties of the salt concrete are presented: mixture of the salt concrete, thermodynamic properties (adiabatic heat release, thermal dilatation, thermal conductivity and heat capacity), mechanical short term properties, creep (under tension, under compression), autogenous shrinkage

Chemotoxic materials in a final repository for high-level radioactive wastes. CHEMOTOX concept for defence in depth concerning ground water protection from chemotoxic materials in a final high-level waste repository

International Nuclear Information System (INIS)

Alt, Stefan; Sailer, Michael; Schmidt, Gerhard; Herbert, Horst-Juergen; Krone, Juergen; Tholen, Marion

2009-01-01

The disposal of high-level radioactive wastes in a final repository includes chemotoxic materials. The chemotoxic materials are either part of the radioactive material or part of the packaging material, or the structures within the repository. In the frame of the licensing procedure it has to be demonstrated that no hazardous pollution of the ground water or other disadvantageous changes can occur. The report describes the common project of the Oeko-Institut e.V., the DBE Technology GmbH and the GRS mbH concerning the possible demonstration of a systematic protection of the groundwater against chemotoxic materials in case of a final high-level-radioactive waste repository in the host materials salt and clay stone.

Waste treatment using molten salt oxidation

International Nuclear Information System (INIS)

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

1996-01-01

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

Hot-wall corrosion testing of simulated high level nuclear waste

International Nuclear Information System (INIS)

Chandler, G.T.; Zapp, P.E.; Mickalonis, J.I.

1995-01-01

Three materials of construction for steam tubes used in the evaporation of high level radioactive waste were tested under heat flux conditions, referred to as hot-wall tests. The materials were type 304L stainless steel alloy C276, and alloy G3. Non-radioactive acidic and alkaline salt solutions containing halides and mercury simulated different high level waste solutions stored or processed at the United States Department of Energy's Savannah River Site. Alloy C276 was also tested for corrosion susceptibility under steady-state conditions. The nickel-based alloys C276 and G3 exhibited excellent corrosion resistance under the conditions studied. Alloy C276 was not susceptible to localized corrosion and had a corrosion rate of 0.01 mpy (0.25 μm/y) when exposed to acidic waste sludge and precipitate slurry at a hot-wall temperature of 150 degrees C. Type 304L was susceptible to localized corrosion under the same conditions. Alloy G3 had a corrosion rate of 0.1 mpy (2.5 μm/y) when exposed to caustic high level waste evaporator solution at a hot-wall temperature of 220 degrees C compared to 1.1 mpy (28.0 μ/y) for type 304L. Under extreme caustic conditions (45 weight percent sodium hydroxide) G3 had a corrosion rate of 0.1 mpy (2.5 μm/y) at a hot-wall temperature of 180 degrees C while type 304L had a high corrosion rate of 69.4 mpy (1.8 mm/y)

Defense waste salt disposal at the Savannah River Plant

International Nuclear Information System (INIS)

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

1984-01-01

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

Laboratory characterization and vitrification of Hanford radioactive high-level waste

International Nuclear Information System (INIS)

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

1991-05-01

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

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

International Nuclear Information System (INIS)

1984-01-01

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

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

International Nuclear Information System (INIS)

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

1976-01-01

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

In situ corrosion studies on selected high level waste packaging materials under simulated disposal conditions in rock salt

International Nuclear Information System (INIS)

Smailos, E.; Schwarzkopf, W.; Koester, R.

1988-01-01

In order to qualify corrosion resistant materials for high level waste (HLW) packagings acting as a long-term barrier in a rock salt repository, the corrosion behavior of preselected materials is being investigated in laboratory-scale and in-situ experiments. This work reports about in-situ corrosion experiments on unalloyed steels, Ti 99.8-Pd, Hastelloy C4, and iron-base alloys, as nodular cast iron, Ni-Resist D4 and Si-cast iron, under simulated disposal conditions. The results of the investigations can be summarized as follows: (1) all materials investigated exhibited high resistance to corrosion under the conditions prevailing in the Brine Migration Test; (2) all materials and above all the materials with passivating oxide layers such as Ti 99.8-Pd and Hastelloy C4 which may corrode selectively already in the presence of minor amounts of brine had been resistant with respect to any type of local corrosion attack; the gamma-radiation of 3 · 10 2 Gy/h did not exert an influence on the corrosion behavior of the materials

High-level waste management research and development program at Oak Ridge National Laboratory

International Nuclear Information System (INIS)

Blomeke, J.O.; Bond, W.D.

1976-01-01

Projections of wastes to be generated through the year 2000 portend a problem of impressive size and complexity but one which can be handled within the framework of current and planned investigative programs. Investigations of the technical feasibility of removing actinide elements from wastes to render the residuals more manageable in terms of hazards and storage requirements indicate that they can be removed from wastes by the minimally desired factors of 10 2 to 10 4 ; however, demonstrations and engineering assessments of chemical flowsheets have yet to be made. Natural salt formations are believed to offer the best prospects for disposal of high-level wastes; other promising geological formations are also being evaluated for their suitability for use in the disposal of wastes

R and D activities on the management of waste chloride salts in KAERI

International Nuclear Information System (INIS)

In-Tae, Kim; Hwan-Seo, Park; Jeong-Gook, Kim; Hee-Chul, Yang; Yong-Joon, Cho; Eung-Ho Kim

2007-01-01

Full text of publication follows. Electrochemical treatment of spent oxide fuels has been intensively studied in KAERI to reduce the volume, heat load and radiotoxicity of high-level wastes. It consists of an electrolytic reduction process to convert the oxide fuel into a metallic form and an electro-refining process to separate TRU elements from the electro-reduced metal ingot. Two types of waste salts are expected to generate from the electrochemical pyro-processes, that is, LiCl salt from the reduction process and LiCl+KCl eutectic salt form the refining process. The R and D strategy of the waste salt management in KAERI can be categorized into two parts: 1) enhancement of safety by the stabilisation/solidification of waste salt that is to be finally disposed of and 2) reduction of the waste generation by the regeneration/recycle of the spent salt after removal of radionuclides in it. A sol-gel technique and a zeolite occlusion technique are under development to stabilize the waste salt. The LiCl salt is stabilised by a low-temperature sol-gel process and then the gel product is solidified into a ceramic-like waste form with an addition of glass frit. Another method uses Zeolite-4A to occlude the LiCl salt into its cage and adsorption site to immobilize the radionuclides. The product, salt-occluded zeolite, is fabricated into another type of a ceramic waste form. For the regeneration and recycle of the spent salt, the radionuclides in the salt are removed by a zeolite process for the LiCl salt and by an oxidation/distillation process for the eutectic salt. The target nuclides to be removed in each process are Cs/Sr and rare earth (RE) elements, respectively. In the oxidation/ distillation process, the rare earth chloride nuclides are oxidised by an oxygen sparging method, and the products are precipitated in the form of oxide or oxychloride REs. After separation of the RE elements from the precipitates by distillation, the refined spent salt with a low content

Impact of thermal constraints on the optimal design of high-level waste repositories in geologic media

Energy Technology Data Exchange (ETDEWEB)

Malbrain, C; Lester, R K [Massachusetts Inst. of Tech., Cambridge (USA). Dept. of Nuclear Engineering

1982-12-01

An approximate, semi-analytical heat conduction model for predicting the time-dependent temperature distribution in the region of a high-level waste repository has been developed. The model provides the basis for a systematic, inexpensive examination of the impact of several independent thermal design constraints on key repository design parameters and for determining the optimal set of design parameters which satisfy these constraints. Illustrative calculations have been carried out for conceptual repository designs for spent pressurized water reactor (PWR) fuel and reprocessed PWR high-level waste in salt and granite media.

Treatment of high-level wastes from the IFR fuel cycle

International Nuclear Information System (INIS)

Johnson, T.R.; Lewis, M.A.; Newman, A.E.; Laidler, J.J.

1992-01-01

The Integral Fast Reactor (IFR) is being developed as a future commercial power source that promises to have important advantages over present reactors, including improved resource conservation and waste management. The spent metal alloy fuels from an IFR will be processed in an electrochemical cell operating at 500 degree C with a molten chloride salt electrolyte and cadmium metal anode. After the actinides have been recovered from several batches of core and blanket fuels, the salt cadmium in this electrorefiner will be treated to separate fission products from residual transuranic elements. This treatment produces a waste salt that contains the alkali metal, alkaline earth, and halide fission products; some of the rare earths; and less than 100 nCi/g of alpha activity. The treated metal wastes contain the rest of the fission products (except T, Kr, and Xe) small amounts of uranium, and only trace amounts of transuranic elements. The current concept for the salt waste form is an aluminosilicate matrix, and the concept for the metal waste form is a corrosion-resistant metal alloy. The processes and equipment being developed to treat and immobilize the salt and metal wastes are described

Treatment of high-level wastes from the IFR fuel cycle

Energy Technology Data Exchange (ETDEWEB)

Johnson, T.R.; Lewis, M.A.; Newman, A.E.; Laidler, J.J.

1992-01-01

The Integral Fast Reactor (IFR) is being developed as a future commercial power source that promises to have important advantages over present reactors, including improved resource conservation and waste management. The spent metal alloy fuels from an IFR will be processed in an electrochemical cell operating at 500{degree}C with a molten chloride salt electrolyte and cadmium metal anode. After the actinides have been recovered from several batches of core and blanket fuels, the salt cadmium in this electrorefiner will be treated to separate fission products from residual transuranic elements. This treatment produces a waste salt that contains the alkali metal, alkaline earth, and halide fission products; some of the rare earths; and less than 100 nCi/g of alpha activity. The treated metal wastes contain the rest of the fission products (except T, Kr, and Xe) small amounts of uranium, and only trace amounts of transuranic elements. The current concept for the salt waste form is an aluminosilicate matrix, and the concept for the metal waste form is a corrosion-resistant metal alloy. The processes and equipment being developed to treat and immobilize the salt and metal wastes are described.

Treatment of high-level wastes from the IFR fuel cycle

Energy Technology Data Exchange (ETDEWEB)

Johnson, T.R.; Lewis, M.A.; Newman, A.E.; Laidler, J.J.

1992-08-01

The Integral Fast Reactor (IFR) is being developed as a future commercial power source that promises to have important advantages over present reactors, including improved resource conservation and waste management. The spent metal alloy fuels from an IFR will be processed in an electrochemical cell operating at 500{degree}C with a molten chloride salt electrolyte and cadmium metal anode. After the actinides have been recovered from several batches of core and blanket fuels, the salt cadmium in this electrorefiner will be treated to separate fission products from residual transuranic elements. This treatment produces a waste salt that contains the alkali metal, alkaline earth, and halide fission products; some of the rare earths; and less than 100 nCi/g of alpha activity. The treated metal wastes contain the rest of the fission products (except T, Kr, and Xe) small amounts of uranium, and only trace amounts of transuranic elements. The current concept for the salt waste form is an aluminosilicate matrix, and the concept for the metal waste form is a corrosion-resistant metal alloy. The processes and equipment being developed to treat and immobilize the salt and metal wastes are described.

Initial performance assessment of the disposal of spent nuclear fuel and high-level waste stored at Idaho National Engineering Laboratory

International Nuclear Information System (INIS)

Rechard, R.P.

1993-12-01

This performance assessment characterized plausible treatment options conceived by the Idaho National Engineering Laboratory (INEL) for its spent fuel and high-level radioactive waste and then modeled the performance of the resulting waste forms in two hypothetical, deep, geologic repositories: one in bedded salt and the other in granite. The results of the performance assessment are intended to help guide INEL in its study of how to prepare wastes and spent fuel for eventual permanent disposal. This assessment was part of the Waste Management Technology Development Program designed to help the US Department of Energy develop and demonstrate the capability to dispose of its nuclear waste, as mandated by the Nuclear Waste Policy Act of 1982. The waste forms comprised about 700 metric tons of initial heavy metal (or equivalent units) stored at the INEL: graphite spent fuel, experimental low enriched and highly enriched spent fuel, and high-level waste generated during reprocessing of some spent fuel. Five different waste treatment options were studied; in the analysis, the options and resulting waste forms were analyzed separately and in combination as five waste disposal groups. When the waste forms were studied in combination, the repository was assumed to also contain vitrified high-level waste from three DOE sites for a common basis of comparison and to simulate the impact of the INEL waste forms on a moderate-sized repository, The performance of the waste form was assessed within the context of a whole disposal system, using the U.S. Environmental Protection Agency's Environmental Radiation Protection Standards for Management and Disposal of Spent Nuclear Fuel, High-Level and Transuranic Radioactive Wastes, 40 CFR 191, promulgated in 1985. Though the waste form behavior depended upon the repository type, all current and proposed waste forms provided acceptable behavior in the salt and granite repositories

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

International Nuclear Information System (INIS)

Shaw, P.; Anderson, B.

1993-07-01

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

High-level waste processing and disposal

International Nuclear Information System (INIS)

Crandall, J.L.; Krause, H.; Sombret, C.; Uematsu, K.

1984-01-01

The national high-level waste disposal plans for France, the Federal Republic of Germany, Japan, and the United States are covered. Three conclusions are reached. The first conclusion is that an excellent technology already exists for high-level waste disposal. With appropriate packaging, spent fuel seems to be an acceptable waste form. Borosilicate glass reprocessing waste forms are well understood, in production in France, and scheduled for production in the next few years in a number of other countries. For final disposal, a number of candidate geological repository sites have been identified and several demonstration sites opened. The second conclusion is that adequate financing and a legal basis for waste disposal are in place in most countries. Costs of high-level waste disposal will probably add about 5 to 10% to the costs of nuclear electric power. The third conclusion is less optimistic. Political problems remain formidable in highly conservative regulations, in qualifying a final disposal site, and in securing acceptable transport routes

Mixing Modeling Analysis For SRS Salt Waste Disposition

International Nuclear Information System (INIS)

Lee, S.

2011-01-01

Nuclear waste at Savannah River Site (SRS) waste tanks consists of three different types of waste forms. They are the lighter salt solutions referred to as supernate, the precipitated salts as salt cake, and heavier fine solids as sludge. The sludge is settled on the tank floor. About half of the residual waste radioactivity is contained in the sludge, which is only about 8 percentage of the total waste volume. Mixing study to be evaluated here for the Salt Disposition Integration (SDI) project focuses on supernate preparations in waste tanks prior to transfer to the Salt Waste Processing Facility (SWPF) feed tank. The methods to mix and blend the contents of the SRS blend tanks were evalutaed to ensure that the contents are properly blended before they are transferred from the blend tank such as Tank 50H to the SWPF feed tank. The work consists of two principal objectives to investigate two different pumps. One objective is to identify a suitable pumping arrangement that will adequately blend/mix two miscible liquids to obtain a uniform composition in the tank with a minimum level of sludge solid particulate in suspension. The other is to estimate the elevation in the tank at which the transfer pump inlet should be located where the solid concentration of the entrained fluid remains below the acceptance criterion (0.09 wt% or 1200 mg/liter) during transfer operation to the SWPF. Tank 50H is a Waste Tank that will be used to prepare batches of salt feed for SWPF. The salt feed must be a homogeneous solution satisfying the acceptance criterion of the solids entrainment during transfer operation. The work described here consists of two modeling areas. They are the mixing modeling analysis during miscible liquid blending operation, and the flow pattern analysis during transfer operation of the blended liquid. The modeling results will provide quantitative design and operation information during the mixing/blending process and the transfer operation of the blended

Disposal of high level and intermediate level radioactive wastes

International Nuclear Information System (INIS)

Flowers, R.H.

1991-01-01

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

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

International Nuclear Information System (INIS)

Cornman, W.R.

1981-12-01

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

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

International Nuclear Information System (INIS)

Jacquier, P.

1989-01-01

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

Overview: Defense high-level waste technology program

International Nuclear Information System (INIS)

Shupe, M.W.; Turner, D.A.

1987-01-01

Defense high-level waste generated by atomic energy defense activities is stored on an interim basis at three U.S. Department of Energy (DOE) operating locations; the Savannah River Plant in South Carolina, the Hanford Site in Washington, and the Idaho National Engineering Laboratory in Idaho. Responsibility for the permanent disposal of this waste resides with DOE's Office of Defense Waste and Transportation Management. The objective of the Defense High-Level Wast Technology Program is to develop the technology for ending interim storage and achieving permanent disposal of all U.S. defense high-level waste. New and readily retrievable high-level waste are immobilized for disposal in a geologic repository. Other high-level waste will be stabilized in-place if, after completion of the National Environmental Policy Act (NEPA) process, it is determined, on a site-specific basis, that this option is safe, cost effective and environmentally sound. The immediate program focus is on implementing the waste disposal strategy selected in compliance with the NEPA process at Savannah River, while continuing progress toward development of final waste disposal strategies at Hanford and Idaho. This paper presents an overview of the technology development program which supports these waste management activities and an assessment of the impact that recent and anticipated legal and institutional developments are expected to have on the program

Laboratory simulation of salt dissolution during waste removal

International Nuclear Information System (INIS)

Wiersma, B.J.; Parish, W.R.

1997-01-01

Laboratory experiments were performed to support the field demonstration of improved techniques for salt dissolution in waste tanks at the Savannah River Site. The tests were designed to investigate three density driven techniques for salt dissolution: (1) Drain-Add-Sit-Remove, (2) Modified Density Gradient, and (3) Continuous Salt Mining. Salt dissolution was observed to be a very rapid process as salt solutions with densities between 1.38-1.4 were frequently removed. Slower addition and removal rates and locating the outlet line at deeper levels below the top of the saltcake provided the best contact between the dissolution water and the saltcake. It was observed that dissolution with 1 M sodium hydroxide solution resulted in salt solutions that were within the current inhibitor requirements for the prevention of stress corrosion cracking. This result was independent of the density driven technique. However, if inhibited water (0.01 M sodium hydroxide and 0.011 M sodium nitrite) was utilized, the salt solutions were frequently outside the inhibitor requirements. Corrosion testing at conditions similar to the environments expected during waste removal was recommended

40 CFR 227.30 - High-level radioactive waste.

Science.gov (United States)

2010-07-01

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

Long-term high-level waste technology program

International Nuclear Information System (INIS)

1980-04-01

The Department of Energy (DOE) is conducting a comprehensive program to isolate all US nuclear wastes from the human environment. The DOE Office of Nuclear Energy - Waste (NEW) has full responsibility for managing the high-level wastes resulting from defense activities and additional responsiblity for providing the technology to manage existing commercial high-level wastes and any that may be generated in one of several alternative fuel cycles. Responsibilities of the Three Divisions of DOE-NEW are shown. This strategy document presents the research and development plan of the Division of Waste Products for long-term immobilization of the high-level radioactive wastes resulting from chemical processing of nuclear reactor fuels and targets. These high-level wastes contain more than 99% of the residual radionuclides produced in the fuels and targets during reactor operations. They include essentially all the fission products and most of the actinides that were not recovered for use

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

International Nuclear Information System (INIS)

Elders, W.A.

1987-01-01

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

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

International Nuclear Information System (INIS)

Kolarik, Z.

1991-11-01

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

Immobilization of IFR salt wastes in mortar

International Nuclear Information System (INIS)

Fischer, D.F.; Johnson, T.R.

1988-01-01

Portland cement-base mortars are being considered for immobilizing chloride salt wastes produced by the fuel cycles of Integral Fast Reactors (IFR). The IFR is a sodium-cooled fast reactor with metal alloy fuels. It has a close-coupled fuel cycle in which fission products are separated from the actinides in an electrochemical cell operating at 500/degree/C. This cell has a liquid cadmium anode in which the fuels are dissolved and a liquid salt electrolyte. The salt will be a mixture of either lithium, potassium, and sodium chlorides or lithium, calcium, barium, and sodium chlorides. One method being considered for immobilizing the treated nontransuranic salt waste is to disperse the salt in a portland cement-base mortar that will be sealed in corrosion-resistant containers. For this application, the grout must be sufficiently fluid that it can be pumped into canister-molds where it will solidify into a strong, leach-resistant material. The set times must be longer than a few hours to allow sufficient time for processing, and the mortar must reach a reasonable compressive strength (/approximately/7 MPa) within three days to permit handling. Because fission product heating will be high, about 0.6 W/kg for a mortar containing 10% waste salt, the effects of elevated temperatures during curing and storage on mortar properties must be considered

High-level waste immobilization program: an overview

International Nuclear Information System (INIS)

Bonner, W.R.

1979-09-01

The High-Level Waste Immobilization Program is providing technology to allow safe, affordable immobilization and disposal of nuclear waste. Waste forms and processes are being developed on a schedule consistent with national needs for immobilization of high-level wastes stored at Savannah River, Hanford, Idaho National Engineering Laboratory, and West Valley, New York. This technology is directly applicable to high-level wastes from potential reprocessing of spent nuclear fuel. The program is removing one more obstacle previously seen as a potential restriction on the use and further development of nuclear power, and is thus meeting a critical technological need within the national objective of energy independence

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

Problems of the final storage of radioactive waste in salt formations

International Nuclear Information System (INIS)

Hofrichter, E.

1977-01-01

The geological conditions for the final storage of radioactive waste, the occurrence of salt formations, and the tectonics of salt domes are discussed. The safety of salt rocks, the impermeability of the rocks, and the thermal problems in the storage of high-activity waste are dealt with. Possibilities and preconditions of final storage in West Germany are discussed. (HPH) [de

Projected Salt Waste Production from a Commercial Pyroprocessing Facility

Directory of Open Access Journals (Sweden)

Michael F. Simpson

2013-01-01

Full Text Available Pyroprocessing of used nuclear fuel inevitably produces salt waste from electrorefining and/or oxide reduction unit operations. Various process design characteristics can affect the actual mass of such waste produced. This paper examines both oxide and metal fuel treatment, estimates the amount of salt waste generated, and assesses potential benefit of process options to mitigate the generation of salt waste. For reference purposes, a facility is considered in which 100 MT/year of fuel is processed. Salt waste estimates range from 8 to 20 MT/year from considering numerous scenarios. It appears that some benefit may be derived from advanced processes for separating fission products from molten salt waste, but the degree of improvement is limited. Waste form production is also considered but appears to be economically unfavorable. Direct disposal of salt into a salt basin type repository is found to be the most promising with respect to minimizing the impact of waste generation on the economic feasibility and sustainability of pyroprocessing.

High-Level Waste Vitrification Facility Feasibility Study

Energy Technology Data Exchange (ETDEWEB)

D. A. Lopez

1999-08-01

A ''Settlement Agreement'' between the Department of Energy and the State of Idaho mandates that all radioactive high-level waste now stored at the Idaho Nuclear Technology and Engineering Center will be treated so that it is ready to be moved out of Idaho for disposal by a compliance date of 2035. This report investigates vitrification treatment of the high-level waste in a High-Level Waste Vitrification Facility based on the assumption that no more New Waste Calcining Facility campaigns will be conducted after June 2000. Under this option, the sodium-bearing waste remaining in the Idaho Nuclear Technology and Engineering Center Tank Farm, and newly generated liquid waste produced between now and the start of 2013, will be processed using a different option, such as a Cesium Ion Exchange Facility. The cesium-saturated waste from this other option will be sent to the Calcine Solids Storage Facilities to be mixed with existing calcine. The calcine and cesium-saturated waste will be processed in the High-Level Waste Vitrification Facility by the end of calendar year 2035. In addition, the High-Level Waste Vitrification Facility will process all newly-generated liquid waste produced between 2013 and the end of 2035. Vitrification of this waste is an acceptable treatment method for complying with the Settlement Agreement. This method involves vitrifying the waste and pouring it into stainless-steel canisters that will be ready for shipment out of Idaho to a disposal facility by 2035. These canisters will be stored at the Idaho National Engineering and Environmental Laboratory until they are sent to a national geologic repository. The operating period for vitrification treatment will be from the end of 2015 through 2035.

High-Level Waste Vitrification Facility Feasibility Study

International Nuclear Information System (INIS)

D. A. Lopez

1999-01-01

A ''Settlement Agreement'' between the Department of Energy and the State of Idaho mandates that all radioactive high-level waste now stored at the Idaho Nuclear Technology and Engineering Center will be treated so that it is ready to be moved out of Idaho for disposal by a compliance date of 2035. This report investigates vitrification treatment of the high-level waste in a High-Level Waste Vitrification Facility based on the assumption that no more New Waste Calcining Facility campaigns will be conducted after June 2000. Under this option, the sodium-bearing waste remaining in the Idaho Nuclear Technology and Engineering Center Tank Farm, and newly generated liquid waste produced between now and the start of 2013, will be processed using a different option, such as a Cesium Ion Exchange Facility. The cesium-saturated waste from this other option will be sent to the Calcine Solids Storage Facilities to be mixed with existing calcine. The calcine and cesium-saturated waste will be processed in the High-Level Waste Vitrification Facility by the end of calendar year 2035. In addition, the High-Level Waste Vitrification Facility will process all newly-generated liquid waste produced between 2013 and the end of 2035. Vitrification of this waste is an acceptable treatment method for complying with the Settlement Agreement. This method involves vitrifying the waste and pouring it into stainless-steel canisters that will be ready for shipment out of Idaho to a disposal facility by 2035. These canisters will be stored at the Idaho National Engineering and Environmental Laboratory until they are sent to a national geologic repository. The operating period for vitrification treatment will be from the end of 2015 through 2035

Embedding of solid high-level wastes into metal and non-metal matrices

International Nuclear Information System (INIS)

Geel, J. van; Eschrich, H.; Dobbels, F.; Favre, P.; Sterner, H.

1980-03-01

The primary objective of embedding solidification high-level waste forms of high specific activity into a matrix material is to obtain final waste composites with moderate inner temperatures, even at large waste loadings per meter cylinder length. Secondary objectives are to produce a non-porous, crack-free composite product with a durability superior to that of the embedded waste form itself. The temperature distribution in composite material composed of vitreous beads embedded into a metal matrix (vitromets) are compared with that in a vitreous block, of equal heat generation per meter height, during short- and long-term storage. It was found that for storage under water, inner temperatures below 100 0 C are assured in vitromets, produced from short-cooled high-level wastes, and containing high waste loadings per metercanister height. The chemical and mechanical stability, as well as the thermal conductivity have been examined for vitromets containing various matrix materials whereby emphasis is imparted to lead- and aluminum alloys. The corrosion of lead- and aluminum alloys in distilled water, brine solution and dry salt has been examined at temperatures up to 230 0 C and pressures up to 3.5 MPa. Some lead alloys were found to exhibit superior corrosion resistance in these chemical environments than certain reference borosilicate glasses. The deformation behavior of vitromets under axial compression has been investigated at different temperatures and varying height diameter ratios. The maturity of the vitromet production is finally demonstrated by presenting process data from hot-laboratory scale and cold semi-industrial scale production units. (author)

An analysis of the technical status of high level radioactive waste and spent fuel management systems

Science.gov (United States)

English, T.; Miller, C.; Bullard, E.; Campbell, R.; Chockie, A.; Divita, E.; Douthitt, C.; Edelson, E.; Lees, L.

1977-01-01

The technical status of the old U.S. mailine program for high level radioactive nuclear waste management, and the newly-developing program for disposal of unreprocessed spent fuel was assessed. The method of long term containment for both of these waste forms is considered to be deep geologic isolation in bedded salt. Each major component of both waste management systems is analyzed in terms of its scientific feasibility, technical achievability and engineering achievability. The resulting matrix leads to a systematic identification of major unresolved technical or scientific questions and/or gaps in these programs.

Radiation transport in high-level waste form

International Nuclear Information System (INIS)

Arakali, V.S.; Barnes, S.M.

1992-01-01

The waste form selected for vitrifying high-level nuclear waste stored in underground tanks at West Valley, NY is borosilicate glass. The maximum radiation level at the surface of a canister filled with the high-level waste form is prescribed by repository design criteria for handling and disposition of the vitrified waste. This paper presents an evaluation of the radiation transport characteristics for the vitreous waste form expected to be produced at West Valley and the resulting neutron and gamma dose rates. The maximum gamma and neutron dose rates are estimated to be less than 7500 R/h and 10 mRem/h respectively at the surface of a West Valley canister filled with borosilicate waste glass

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

International Nuclear Information System (INIS)

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

2004-01-01

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

The safe disposal of radioactive wastes in geologic salt formations

International Nuclear Information System (INIS)

Kuehn, K.; Proske, R.

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

Organic waste processing using molten salt oxidation

Energy Technology Data Exchange (ETDEWEB)

Adamson, M. G., LLNL

1998-03-01

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

Corrosion testing of selected packaging materials for disposal of high-level waste glass in rock salt formations

International Nuclear Information System (INIS)

Smailos, E.; Schwarzkopf, W.; Koester, R.; Fiehn, B.; Halm, G.

1990-05-01

In previous corrosion studies performed in salt brines, unalloyed steels, Ti 99.8-Pd and Hastelloy C4 have proved to be the most promising materials for long-term resistant packagings to be used in heat-generating waste (vitrified HLW, spent fuel) disposal in rock-salt formations. To characterise the corrosion behaviour of these materials in more detail, further in-depth laboratory-scale and in-situ corrosion studies have been performed in the present study. Besides the above-mentioned materials, also some in-situ investigations of the iron-base materials Ni-Resist D2 and D4, cast iron and Si-cast iron have been carried out in order to complete the results available to date. (orig.) [de

Ramifications of defining high-level waste

International Nuclear Information System (INIS)

Wood, D.E.; Campbell, M.H.; Shupe, M.W.

1987-01-01

The Nuclear Regulatory Commission (NRC) is considering rule making to provide a concentration-based definition of high-level waste (HLW) under authority derived from the Nuclear Waste Policy Act (NWPA) of 1982 and the Low Level Waste Policy Amendments Act of 1985. The Department of Energy (DOE), which has the responsibility to dispose of certain kinds of commercial waste, is supporting development of a risk-based classification system by the Oak Ridge National Laboratory to assist in developing and implementing the NRC rule. The system is two dimensional, with the axes based on the phrases highly radioactive and requires permanent isolation in the definition of HLW in the NWPA. Defining HLW will reduce the ambiguity in the present source-based definition by providing concentration limits to establish which materials are to be called HLW. The system allows the possibility of greater-confinement disposal for some wastes which do not require the degree of isolation provided by a repository. The definition of HLW will provide a firm basis for waste processing options which involve partitioning of waste into a high-activity stream for repository disposal, and a low-activity stream for disposal elsewhere. Several possible classification systems have been derived and the characteristics of each are discussed. The Defense High Level Waste Technology Lead Office at DOE - Richland Operations Office, supported by Rockwell Hanford Operations, has coordinated reviews of the ORNL work by a technical peer review group and other DOE offices. The reviews produced several recommendations and identified several issues to be addressed in the NRC rule making. 10 references, 3 figures

Sampling and analyses of SRP high-level waste sludges

International Nuclear Information System (INIS)

Stone, J.A.; Kelley, J.A.; McMillan, T.S.

1976-08-01

Twelve 3-liter samples of high-heat waste sludges were collected from four Savannah River Plant waste tanks with a hydraulically operated sample collector of unique design. Ten of these samples were processed in Savannah River Laboratory shielded cell facilities, yielding 5.3 kg of washed, dried sludge products for waste solidification studies. After initial drying, each batch was washed by settling and decantation to remove the bulk of soluble salts and then was redried. Additional washes were by filtration, followed by final drying. Conclusions from analyses of samples taken during the processing steps were: (a) the raw sludges contained approximately 80 wt percent soluble salts, most of which were removed by the washes; (b) 90 Sr and 238 , 239 Pu remained in the sludges, but most of the 137 Cs was removed by washing; (c) small amounts of sodium, sulfate, and 137 Cs remained in the sludges after thorough washing; (d) no significant differences were found in sludge samples taken from different risers of one waste tank. Chemical and radiometric compositions of the sludge product from each tank were determined. The sludges had diverse compositions, but iron, manganese, aluminum, and uranium were principal elements in each sludge. 90 Sr was the predominant radionuclide in each sludge product

Current high-level waste solidification technology

International Nuclear Information System (INIS)

Bonner, W.F.; Ross, W.A.

1976-01-01

Technology has been developed in the U.S. and abroad for solidification of high-level waste from nuclear power production. Several processes have been demonstrated with actual radioactive waste and are now being prepared for use in the commercial nuclear industry. Conversion of the waste to a glass form is favored because of its high degree of nondispersibility and safety

Initial performance assessment of the disposal of spent nuclear fuel and high-level waste stored at Idaho National Engineering Laboratory. Volume 2: Appendices

Energy Technology Data Exchange (ETDEWEB)

Rechard, R.P. [ed.

1993-12-01

This performance assessment characterized plausible treatment options conceived by the Idaho National Engineering Laboratory (INEL) for its spent fuel and high-level radioactive waste and then modeled the performance of the resulting waste forms in two hypothetical, deep, geologic repositories: one in bedded salt and the other in granite. The results of the performance assessment are intended to help guide INEL in its study of how to prepare wastes and spent fuel for eventual permanent disposal. This assessment was part of the Waste Management Technology Development Program designed to help the US Department of Energy develop and demonstrate the capability to dispose of its nuclear waste, as mandated by the Nuclear Waste Policy Act of 1982. The waste forms comprised about 700 metric tons of initial heavy metal (or equivalent units) stored at the INEL: graphite spent fuel, experimental low enriched and highly enriched spent fuel, and high-level waste generated during reprocessing of some spent fuel. Five different waste treatment options were studied; in the analysis, the options and resulting waste forms were analyzed separately and in combination as five waste disposal groups. When the waste forms were studied in combination, the repository was assumed to also contain vitrified high-level waste from three DOE sites for a common basis of comparison and to simulate the impact of the INEL waste forms on a moderate-sized repository, The performance of the waste form was assessed within the context of a whole disposal system, using the U.S. Environmental Protection Agency`s Environmental Radiation Protection Standards for Management and Disposal of Spent Nuclear Fuel, High-Level and Transuranic Radioactive Wastes, 40 CFR 191, promulgated in 1985. Though the waste form behavior depended upon the repository type, all current and proposed waste forms provided acceptable behavior in the salt and granite repositories.

Next Generation Extractants for Cesium Separation from High-Level Waste: From Fundamental Concepts to Site Implementation

International Nuclear Information System (INIS)

Moyer, Bruce A.; Bazelaire, Eve; Bonnesen, Peter V.; Bryan, Jeffrey C.; Delmau, Latitia H.; Engle, Nancy L.; Gorbunova, Maryna G.; Keever, Tamara J.; Levitskaia, Tatiana G.; Sachleben, Richard A.; Tomkins, Bruce A.; Bartsch, Richard A.

2004-01-01

General project objectives. This project seeks a fundamental understanding and major improvement in cesium separation from high-level waste by cesium-selective calixcrown extractants. Systems of particular interest involve novel solvent-extraction systems containing specific members of the calix[4]arene-crown-6 family, alcohol solvating agents, and alkylamines. Questions being addressed pertain to cesium binding strength, extraction selectivity, cesium stripping, and extractant solubility. Enhanced properties in this regard will specifically benefit cleanup projects funded by the USDOE Office of Environmental Management to treat and dispose of high-level radioactive wastes currently stored in underground tanks at the Savannah River Site (SRS), the Hanford site, and the Idaho National Environmental and Engineering Laboratory.1 The most direct beneficiary will be the SRS Salt Processing Project, which has recently identified the Caustic-Side Solvent Extraction (CSSX) process employing a calixcrown as its preferred technology for cesium removal from SRS high level tank waste.2 This technology owes its development in part to fundamental results obtained in this program

Next Generation Extractants for Cesium Separation from High-Level Waste: From Fundamental Concepts to Site Implementation

International Nuclear Information System (INIS)

Moyer, Bruce A; Bazelaire, Eve; Bonnesen, Peter V.; Bryan, Jeffrey C.; Delmau, Laetitia H.; Engle, Nancy L.; Gorbunova, Maryna G.; Keever, Tamara J.; Levitskaia, Tatiana G.; Sachleben, Richard A.; Tomkins, Bruce A.; Bartsch, Richard A.; Talanov, Vladimir S.; Gibson, Harry W.; Jones, Jason W.; Hay, Benjamin P.

2003-01-01

This project seeks a fundamental understanding and major improvement in cesium separation from high-level waste by cesium-selective calixcrown extractants. Systems of particular interest involve novel solvent-extraction systems containing specific members of the calix[4]arene-crown-6 family, alcohol solvating agents, and alkylamines. Questions being addressed pertain to cesium binding strength, extraction selectivity, cesium stripping, and extractant solubility. Enhanced properties in this regard will specifically benefit cleanup projects funded by the USDOE Office of Environmental Management to treat and dispose of high-level radioactive wastes currently stored in underground tanks at the Savannah River Site (SRS), the Hanford site, and the Idaho National Environmental and Engineering Laboratory.1 The most direct beneficiary will be the SRS Salt Processing Project, which has recently identified the Caustic-Side Solvent Extraction (CSSX) process employing a calixcrown as its preferred technology for cesium removal from SRS high-level tank waste.2 This technology owes its development in part to fundamental results obtained in this program

Handbook of high-level radioactive waste transportation

International Nuclear Information System (INIS)

Sattler, L.R.

1992-10-01

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

Underground disposal of vitrified high level radioactive waste: a review of research and development

International Nuclear Information System (INIS)

1982-11-01

A review has been undertaken of the worldwide status of research and development related to the geological disposal of vitrified high level radioactive waste. The nature and quantities of vitrified high level waste that will arise from nuclear power generation in the UK have been estimated and considered. The safety case for establishing a geological repository would have to be based on predictive models, which could adequately represent the interactions and effects of a wide range of gradual processes and possible sudden events. No detailed repository design has yet been published, but the configuration currently favoured, in the UK and in most other countries, comprises a small number of vertical shafts, from which a network of horizontal tunnels would be excavated. Waste packages would be placed in holes drilled in the floors of the tunnels. The excavation of such a repository in hard crystalline rock, in a thick homogeneous formation of rock salt, or in the less plastic argillaceous formations, appears to be within the scope of present technology. Rock types available in the UK, which are likely to prove suitable for the accommodation of a repository, have been identified. The strategies and programmes for high level waste disposal in other countries have been reviewed. (U.K.)

Risk assessment of nonhazardous oil-field waste disposal in salt caverns.

Energy Technology Data Exchange (ETDEWEB)

Elcock, D.

1998-03-10

that if caverns are sited and designed well, operated carefully, closed properly, and monitored routinely, they could, from technical and legal perspectives, be suitable for disposing of oil-field wastes. On the basis of these findings, ANL subsequently conducted a preliminary risk assessment on the possibility that adverse human health effects (carcinogenic and noncarcinogenic) could result from exposure to contaminants released from the NOW disposed of in salt caverns. The methodology for the risk assessment included the following steps: identifying potential contaminants of concern; determining how humans could be exposed to these contaminants; assessing contaminant toxicities; estimating contaminant intakes; and estimating human cancer and noncancer risks. To estimate exposure routes and pathways, four postclosure cavern release scenarios were assessed. These were inadvertent cavern intrusion, failure of the cavern seal, failure of the cavern through cracks, failure of the cavern through leaky interbeds, and partial collapse of the cavern roof. Assuming a single, generic, salt cavern and generic oil-field wastes, potential human health effects associated with constituent hazardous substances (arsenic, benzene, cadmium, and chromium) were assessed under each of these scenarios. Preliminary results provided excess cancer risk and hazard index (for noncancer health effects) estimates that were well within the EPA target range for acceptable exposure risk levels. These results lead to the preliminary conclusion that from a human health perspective, salt caverns can provide an acceptable disposal method for nonhazardous oil-field wastes.

Risk assessment of nonhazardous oil-field waste disposal in salt caverns

International Nuclear Information System (INIS)

Elcock, D.

1998-01-01

that if caverns are sited and designed well, operated carefully, closed properly, and monitored routinely, they could, from technical and legal perspectives, be suitable for disposing of oil-field wastes. On the basis of these findings, ANL subsequently conducted a preliminary risk assessment on the possibility that adverse human health effects (carcinogenic and noncarcinogenic) could result from exposure to contaminants released from the NOW disposed of in salt caverns. The methodology for the risk assessment included the following steps: identifying potential contaminants of concern; determining how humans could be exposed to these contaminants; assessing contaminant toxicities; estimating contaminant intakes; and estimating human cancer and noncancer risks. To estimate exposure routes and pathways, four postclosure cavern release scenarios were assessed. These were inadvertent cavern intrusion, failure of the cavern seal, failure of the cavern through cracks, failure of the cavern through leaky interbeds, and partial collapse of the cavern roof. Assuming a single, generic, salt cavern and generic oil-field wastes, potential human health effects associated with constituent hazardous substances (arsenic, benzene, cadmium, and chromium) were assessed under each of these scenarios. Preliminary results provided excess cancer risk and hazard index (for noncancer health effects) estimates that were well within the EPA target range for acceptable exposure risk levels. These results lead to the preliminary conclusion that from a human health perspective, salt caverns can provide an acceptable disposal method for nonhazardous oil-field wastes

High-level radioactive waste disposal: Key geochemical issues and information needs for site characterization

International Nuclear Information System (INIS)

Brooks, D.J.; Bembia, P.J.; Bradbury, J.W.; Jackson, K.C.; Kelly, W.R.; Kovach, L.A.; Mo, T.; Tesoriero, J.A.

1986-01-01

Geochemistry plays a key role in determining the potential of a high-level radioactive waste disposal site for long-term radionuclide containment and isolation. The Nuclear Regulatory Commission (NRC) has developed a set of issues and information needs important for characterizing geochemistry at the potential sites being investigated by the Department of Energy Basalt Waste Isolation Project, Nevada Nuclear Waste Storage Investigations project, and Salt Repository Project. The NRC site issues and information needs consider (1) the geochemical environment of the repository, (2) changes to the initial geochemical environment caused by construction and waste emplacement, and (3) interactions that affect the transport of waste radionuclides to the accessible environment. The development of these issues and information needs supports the ongoing effort of the NRC to identify and address areas of geochemical data uncertainty during prelicensing interactions

Initial performance assessment of the disposal of spent nuclear fuel and high-level waste stored at Idaho National Engineering Laboratory

International Nuclear Information System (INIS)

Rechard, R.P.

1993-12-01

This performance assessment characterized plausible treatment options conceived by the Idaho National Engineering Laboratory (INEL) for its spent fuel and high-level radioactive waste and then modeled the performance of the resulting waste forms in two hypothetical, deep, geologic repositories: one in bedded salt and the other in granite. The results of the performance assessment are intended to help guide INEL in its study of how to prepare wastes and spent fuel for eventual permanent disposal. This assessment was part of the Waste Management Technology Development Program designed to help the US Department of Energy develop and demonstrate the capability to dispose of its nuclear waste. Although numerous caveats must be placed on the results, the general findings were as follows: Though the waste form behavior depended upon the repository type, all current and proposed waste forms provided acceptable behavior in the salt and granite repositories

Thermal Analysis of Disposal of High-Level Nuclear Waste in a Generic Bedded Salt repository using the Semi-Analytical Method.

Energy Technology Data Exchange (ETDEWEB)

Hadgu, Teklu [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Matteo, Edward N. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2017-05-01

An example case is presented for testing analytical thermal models. The example case represents thermal analysis of a generic repository in bedded salt at 500 m depth. The analysis is part of the study reported in Matteo et al. (2016). Ambient average ground surface temperature of 15°C, and a natural geothermal gradient of 25°C/km, were assumed to calculate temperature at the near field. For generic salt repository concept crushed salt backfill is assumed. For the semi-analytical analysis crushed salt thermal conductivity of 0.57 W/m-K was used. With time the crushed salt is expected to consolidate into intact salt. In this study a backfill thermal conductivity of 3.2 W/m-K (same as intact) is used for sensitivity analysis. Decay heat data for SRS glass is given in Table 1. The rest of the parameter values are shown below. Results of peak temperatures at the waste package surface are given in Table 2.

Disposal of high-level radioactive waste

International Nuclear Information System (INIS)

Glasby, G.P.

1977-01-01

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

The immobilization of High Level Waste Into Glass

International Nuclear Information System (INIS)

Aisyah; Martono, H.

1998-01-01

High level liquid waste is generated from the first step extraction in the nuclear fuel reprocessing. The waste is immobilized with boro-silicate glass. A certain composition of glass is needed for a certain type of waste, so that the properties of waste glass would meet the requirement either for further process or for disposal. The effect of waste loading on either density, thermal expansion, softening point and leaching rate has been studied. The composition of the high level liquid waste has been determined by ORIGEN 2 and the result has been used to prepare simulated high level waste. The waste loading in the waste glass has been set to be 19.48; 22.32; 25.27; and 26.59 weight percent. The result shows that increasing the waste loading has resulted in the higher density with no thermal expansion and softening point significant change. The increase in the waste loading increase that leaching rate. The properties of the waste glass in this research have not shown any deviation from the standard waste glass properties

Evaluation of radionuclide concentrations in high-level radioactive wastes

International Nuclear Information System (INIS)

Fehringer, D.J.

1985-10-01

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

Handling and storage of conditioned high-level wastes

International Nuclear Information System (INIS)

1983-01-01

This report deals with certain aspects of the management of one of the most important wastes, i.e. the handling and storage of conditioned (immobilized and packaged) high-level waste from the reprocessing of spent nuclear fuel and, although much of the material presented here is based on information concerning high-level waste from reprocessing LWR fuel, the principles, as well as many of the details involved, are applicable to all fuel types. The report provides illustrative background material on the arising and characteristics of high-level wastes and, qualitatively, their requirements for conditioning. The report introduces the principles important in conditioned high-level waste storage and describes the types of equipment and facilities, used or studied, for handling and storage of such waste. Finally, it discusses the safety and economic aspects that are considered in the design and operation of handling and storage facilities

Disposal of high-level waste from nuclear power plants in Denmark. Salt dome investigations. v.1

International Nuclear Information System (INIS)

1981-01-01

A summary is presented of a report in five volumes on possible disposal of radioactive waste in Denmark. The investigation was made by the Danish electric utilities ELKRAFT and ELSAM at the request of the Danish Government. The investigation proved it possible to consider two alternative designs for a disposal facility, one based on the deposition of waste in individual, deep holes, the other on placing the waste in mine galleries. A safety analysis was completed with the Mors dome as example. The purpose of the analysis was to prove whether safe disposal of high-level waste in Denmark was feasible. The utilities concluded that the results of the analysis were satisfactory and the report is now being assessed by the authorities. (BP)

High level nuclear wastes

International Nuclear Information System (INIS)

Lopez Perez, B.

1987-01-01

The transformations involved in the nuclear fuels during the burn-up at the power nuclear reactors for burn-up levels of 33.000 MWd/th are considered. Graphs and data on the radioactivity variation with the cooling time and heat power of the irradiated fuel are presented. Likewise, the cycle of the fuel in light water reactors is presented and the alternatives for the nuclear waste management are discussed. A brief description of the management of the spent fuel as a high level nuclear waste is shown, explaining the reprocessing and giving data about the fission products and their radioactivities, which must be considered on the vitrification processes. On the final storage of the nuclear waste into depth geological burials, both alternatives are coincident. The countries supporting the reprocessing are indicated and the Spanish programm defined in the Plan Energetico Nacional (PEN) is shortly reviewed. (author) 8 figs., 4 tabs

Process for solidifying high-level nuclear waste

Science.gov (United States)

Ross, Wayne A.

1978-01-01

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

Characteristics of solidified high-level waste products

International Nuclear Information System (INIS)

1979-01-01

The object of the report is to contribute to the establishment of a data bank for future preparation of codes of practice and standards for the management of high-level wastes. The work currently in progress on measuring the properties of solidified high-level wastes is being studied

Application of molten salt oxidation for the minimization and recovery of plutonium-238 contaminated wastes

International Nuclear Information System (INIS)

Wishau, R.; Ramsey, K.B.; Montoya, A.

1998-01-01

This paper presents the technical and economic feasibility of molten salt oxidation technology as a volume reduction and recovery process for 238 Pu contaminated waste. Combustible low-level waste material contaminated with 238 Pu residue is destroyed by oxidation in a 900 C molten salt reaction vessel. The combustible waste is destroyed creating carbon dioxide and steam and a small amount of ash and insoluble 2328 Pu in the spent salt. The valuable 238 Pu is recycled using aqueous recovery techniques. Experimental test results for this technology indicate a plutonium recovery efficiency of 99%. Molten salt oxidation stabilizes the waste converting it to a non-combustible waste. Thus installation and use of molten salt oxidation technology will substantially reduce the volume of 238 Pu contaminated waste. Cost-effectiveness evaluations of molten salt oxidation indicate a significant cost savings when compared to the present plans to package, or re-package, certify and transport these wastes to the Waste Isolation Pilot Plant for permanent disposal. Clear and distinct cost advantages exist for MSO when the monetary value of the recovered 238 Pu is considered

National high-level waste systems analysis report

Energy Technology Data Exchange (ETDEWEB)

Kristofferson, K.; Oholleran, T.P.; Powell, R.H.

1995-09-01

This report documents the assessment of budgetary impacts, constraints, and repository availability on the storage and treatment of high-level waste and on both existing and pending negotiated milestones. The impacts of the availabilities of various treatment systems on schedule and throughput at four Department of Energy sites are compared to repository readiness in order to determine the prudent application of resources. The information modeled for each of these sites is integrated with a single national model. The report suggests a high-level-waste model that offers a national perspective on all high-level waste treatment and storage systems managed by the Department of Energy.

National high-level waste systems analysis report

International Nuclear Information System (INIS)

Kristofferson, K.; Oholleran, T.P.; Powell, R.H.

1995-09-01

This report documents the assessment of budgetary impacts, constraints, and repository availability on the storage and treatment of high-level waste and on both existing and pending negotiated milestones. The impacts of the availabilities of various treatment systems on schedule and throughput at four Department of Energy sites are compared to repository readiness in order to determine the prudent application of resources. The information modeled for each of these sites is integrated with a single national model. The report suggests a high-level-waste model that offers a national perspective on all high-level waste treatment and storage systems managed by the Department of Energy

High-Level Waste Melter Study Report

Energy Technology Data Exchange (ETDEWEB)

Perez, Joseph M.; Bickford, Dennis F.; Day, Delbert E.; Kim, Dong-Sang; Lambert, Steven L.; Marra, Sharon L.; Peeler, David K.; Strachan, Denis M.; Triplett, Mark B.; Vienna, John D.; Wittman, Richard S.

2001-07-13

At the Hanford Site in Richland, Washington, the path to site cleanup involves vitrification of the majority of the wastes that currently reside in large underground tanks. A Joule-heated glass melter is the equipment of choice for vitrifying the high-level fraction of these wastes. Even though this technology has general national and international acceptance, opportunities may exist to improve or change the technology to reduce the enormous cost of accomplishing the mission of site cleanup. Consequently, the U.S. Department of Energy requested the staff of the Tanks Focus Area to review immobilization technologies, waste forms, and modifications to requirements for solidification of the high-level waste fraction at Hanford to determine what aspects could affect cost reductions with reasonable long-term risk. The results of this study are summarized in this report.

Radiological consequences of a human intrusion in a nuclear waste repository in a salt formation

International Nuclear Information System (INIS)

Jacquier, P.; Raimbault, P.

1989-07-01

The assessment of the consequences of human intrusion scenarios for a repository is very important for salt formations, since this material has an undeniable economic interest. In this work, the scenario considers the solution mining of salt for human consumption: salt is extracted from a cavern; by leaching, this cavern enlarges and uncovers the waste, which falls down into the sump. It was assumed that the intrusion takes place either 500 years or 2500 years after the closing of the repository. High-level vitrified waste or alpha cemented waste were considered. This paper displays the assumptions made and, using a simplified modelling of the phenomena, the estimation of the radiological consequences due to ingestion of contamined sals. A sensitivity/uncertainty analysis is presented which emphasizes several fields where experimental studies have to be pursued or launched [fr

Cermets for high level waste containment

International Nuclear Information System (INIS)

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

1978-01-01

Cermet materials are currently under investigation as an alternate for the primary containment of high level wastes. The cermet in this study is an iron--nickel base metal matrix containing uniformly dispersed, micron-size fission product oxides, aluminosilicates, and titanates. Cermets possess high thermal conductivity, and typical waste loading of 70 wt % with volume reduction factors of 2 to 200 and low processing volatility losses have been realized. Preliminary leach studies indicate a leach resistance comparable to other candidate waste forms; however, more quantitative data are required. Actual waste studies have begun on NFS Acid Thorex, SRP dried sludge and fresh, unneutralized SRP process wastes

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

International Nuclear Information System (INIS)

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

2006-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

NONE

2015-10-15

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

Handling and storage of conditioned high-level wastes

International Nuclear Information System (INIS)

Heafield, W.

1984-01-01

This paper deals with certain aspects of the management of one of the most important radioactive wastes arising from the nuclear fuel cycle, i.e. the handling and storage of conditioned high-level wastes. The paper is based on an IAEA report of the same title published during 1983 in the Technical Reports Series. The paper provides illustrative background material on the characteristics of high-level wastes and, qualitatively, their requirements for conditioning. The principles important in the storage of high-level wastes are reviewed in conjunction with the radiological and socio-political considerations involved. Four fundamentally different storage concepts are described with reference to published information and the safety aspects of particular storage concepts are discussed. Finally, overall conclusions are presented which confirm the availability of technology for constructing and operating conditioned high-level waste storage facilities for periods of at least several decades. (author)

Characteristics of solidified products containing radioactive molten salt waste.

Science.gov (United States)

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

2007-11-01

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

Overview of high-level waste management accomplishments

International Nuclear Information System (INIS)

Lawroski, H.; Berreth, J.R.; Freeby, W.A.

1980-01-01

Storage of power reactor spent fuel is necessary at present because of the lack of reprocessing operations particularly in the U.S. By considering the above solidification and storage scenario, there is more than reasonable assurance that acceptable, stable, low heat generation rate, solidified waste can be produced, and safely disposed. The public perception of no waste disposal solutions is being exploited by detractors of nuclear power application. The inability to even point to one overall system demonstration lends credibility to the negative assertions. By delaying the gathering of on-line information to qualify repository sites, and to implement a demonstration, the actions of the nuclear power detractors are self serving in that they can continue to point out there is no demonstration of satisfactory high-level waste disposal. By maintaining the liquid and solidified high-level waste in secure above ground storage until acceptable decay heat generation rates are achieved, by producing a compatible, high integrity, solid waste form, by providing a second or even third barrier as a compound container and by inserting the enclosed waste form in a qualified repository with spacing to assure moderately low temperature disposal conditions, there appears to be no technical reason for not progressing further with the disposal of high-level wastes and needed implementation of the complete nuclear power fuel cycle

Solid waste disposal into salt mines

International Nuclear Information System (INIS)

Repke, W.

1981-01-01

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

Evaluation of conditioned high-level waste forms

International Nuclear Information System (INIS)

Mendel, J.E.; Turcotte, R.P.; Chikalla, T.D.; Hench, L.L.

1983-01-01

The evaluation of conditioned high-level waste forms requires an understanding of radiation and thermal effects, mechanical properties, volatility, and chemical durability. As a result of nuclear waste research and development programs in many countries, a good understanding of these factors is available for borosilicate glass containing high-level waste. The IAEA through its coordinated research program has contributed to this understanding. Methods used in the evaluation of conditioned high-level waste forms are reviewed. In the US, this evaluation has been facilitated by the definition of standard test methods by the Materials Characterization Center (MCC), which was established by the Department of Energy (DOE) in 1979. The DOE has also established a 20-member Materials Review Board to peer-review the activities of the MCC. In addition to comparing waste forms, testing must be done to evaluate the behavior of waste forms in geologic repositories. Such testing is complex; accelerated tests are required to predict expected behavior for thousands of years. The tests must be multicomponent tests to ensure that all potential interactions between waste form, canister/overpack and corrosion products, backfill, intruding ground water and the repository rock, are accounted for. An overview of the status of such multicomponent testing is presented

Deep geologic disposal of mixed waste in bedded salt: The Waste Isolation Pilot Plant

International Nuclear Information System (INIS)

Rempe, N.T.

1993-01-01

Mixed waste (i.e., waste that contains both chemically hazardous and radioactive components) poses a moral, political, and technical challenge to present and future generations. But an international consensus is emerging that harmful byproducts and residues can be permanently isolated from the biosphere in a safe and environmentally responsible manner by deep geologic disposal. To investigate and demonstrate such disposal for transuranic mixed waste, derived from defense-related activities, the US Department of Energy has prepared the Waste Isolation Pilot Plant (WIPP) near Carlsbad, New Mexico. This research and development facility was excavated approximately at the center of a 600 m thick sequence of salt (halite) beds, 655 m below the surface. Proof of the long-term tectonic and hydrological stability of the region is supplied by the fact that these salt beds have remained essentially undisturbed since they were deposited during the Late Permian age, approximately 225 million years ago. Plutonium-239, the main radioactive component of transuranic mixed waste, has a half-life of 24,500 years. Even ten half-lives of this isotope - amounting to about a quarter million years, the time during which its activity will decline to background level represent only 0.11 percent of the history of the repository medium. Therefore, deep geologic disposal of transuranic mixed waste in Permian bedded salt appears eminently feasible

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

International Nuclear Information System (INIS)

Hsu, P.C.

1997-01-01

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

Evaluation of solidified high-level waste forms

International Nuclear Information System (INIS)

1981-01-01

One of the objectives of the IAEA waste management programme is to coordinate and promote development of improved technology for the safe management of radioactive wastes. The Agency accomplished this objective specifically through sponsoring Coordinated Research Programmes on the ''Evaluation of Solidified High Level Waste Products'' in 1977. The primary objectives of this programme are to review and disseminate information on the properties of solidified high-level waste forms, to provide a mechanism for analysis and comparison of results from different institutes, and to help coordinate future plans and actions. This report is a summary compilation of the key information disseminated at the second meeting of this programme

Permanent Disposal of Nuclear Waste in Salt

Science.gov (United States)

Hansen, F. D.

2016-12-01

Salt formations hold promise for eternal removal of nuclear waste from our biosphere. Germany and the United States have ample salt formations for this purpose, ranging from flat-bedded formations to geologically mature dome structures. Both nations are revisiting nuclear waste disposal options, accompanied by extensive collaboration on applied salt repository research, design, and operation. Salt formations provide isolation while geotechnical barriers reestablish impermeability after waste is placed in the geology. Between excavation and closure, physical, mechanical, thermal, chemical, and hydrological processes ensue. Salt response over a range of stress and temperature has been characterized for decades. Research practices employ refined test techniques and controls, which improve parameter assessment for features of the constitutive models. Extraordinary computational capabilities require exacting understanding of laboratory measurements and objective interpretation of modeling results. A repository for heat-generative nuclear waste provides an engineering challenge beyond common experience. Long-term evolution of the underground setting is precluded from direct observation or measurement. Therefore, analogues and modeling predictions are necessary to establish enduring safety functions. A strong case for granular salt reconsolidation and a focused research agenda support salt repository concepts that include safety-by-design. Sandia National Laboratories is a multi-program laboratory managed and operated by Sandia Corporation, a wholly owned subsidiary of Lockheed Martin Corporation, for the U.S. Department of Energy's National Nuclear Security Administration under contract DE-AC04-94AL85000. Author: F. D. Hansen, Sandia National Laboratories

Partitioning of high level liquid waste: experiences in plant level adoption

International Nuclear Information System (INIS)

Manohar, Smitha; Kaushik, C.P.

2016-01-01

High Level Radioactive Wastes are presently vitrified in borosilicate matrices in all our back end facilities in our country. This is in accordance with internationally endorsed methodology for the safe management of high level radioactive wastes. Recent advancements in the field of partitioning technology in our group, has presented us with an opportunity to have a fresh perspective on management of high level liquid radioactive wastes streams, that emanate from reprocessing operations. This paper will highlight our experiences with respect to both partitioning studies and vitrification practices, with a focus on waste volume reduction for final disposal. Incorporation of this technique has led to the implementation of the concept of recovering wealth from waste, a marked decrease on the load of disposal in deep geological repositories and serve as a step towards the vision of transmutation of long lived radionuclides

Considerations of the Differences between Bedded and Domal Salt Pertaining to Disposal of Heat-Generating Nuclear Waste

Energy Technology Data Exchange (ETDEWEB)

Hansen, Francis D. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Kuhlman, Kristopher L. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sobolik, Steven R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2016-07-07

Salt formations hold promise for eternal removal of nuclear waste from our biosphere. Germany and the United States have ample salt formations for this purpose, ranging from flat-bedded formations to geologically mature dome structures. As both nations revisit nuclear waste disposal options, the choice between bedded, domal, or intermediate pillow formations is once again a contemporary issue. For decades, favorable attributes of salt as a disposal medium have been extoled and evaluated, carefully and thoroughly. Yet, a sense of discovery continues as science and engineering interrogate naturally heterogeneous systems. Salt formations are impermeable to fluids. Excavation-induced fractures heal as seal systems are placed or natural closure progresses toward equilibrium. Engineering required for nuclear waste disposal gains from mining and storage industries, as humans have been mining salt for millennia. This great intellectual warehouse has been honed and distilled, but not perfected, for all nuances of nuclear waste disposal. Nonetheless, nations are able and have already produced suitable license applications for radioactive waste disposal in salt. A remaining conundrum is site location. Salt formations provide isolation and geotechnical barriers reestablish impermeability after waste is placed in the geology. Between excavation and closure, physical, mechanical, thermal, chemical, and hydrological processes ensue. Positive attributes for isolation in salt have many commonalities independent of the geologic setting. In some cases, specific details of the environment will affect the disposal concept and thereby define interaction of features, events and processes, while simultaneously influencing scenario development. Here we identify and discuss high-level differences and similarities of bedded and domal salt formations. Positive geologic and engineering attributes for disposal purposes are more common among salt formations than are significant differences

Considerations of the Differences between Bedded and Domal Salt Pertaining to Disposal of Heat-Generating Nuclear Waste

Energy Technology Data Exchange (ETDEWEB)

Hansen, Francis D. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Kuhlman, Kristopher L. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sobolik, Steven R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2016-07-07

Salt formations hold promise for eternal removal of nuclear waste from our biosphere. Germany and the United States have ample salt formations for this purpose, ranging from flat-bedded formations to geologically mature dome structures. As both nations revisit nuclear waste disposal options, the choice between bedded, domal, or intermediate pillow formations is once again a contemporary issue. For decades, favorable attributes of salt as a disposal medium have been extoled and evaluated, carefully and thoroughly. Yet, a sense of discovery continues as science and engineering interrogate naturally heterogeneous systems. Salt formations are impermeable to fluids. Excavation-induced fractures heal as seal systems are placed or natural closure progresses toward equilibrium. Engineering required for nuclear waste disposal gains from mining and storage industries, as humans have been mining salt for millennia. This great intellectual warehouse has been honed and distilled, but not perfected, for all nuances of nuclear waste disposal. Nonetheless, nations are able and have already produced suitable license applications for radioactive waste disposal in salt. A remaining conundrum is site location. Salt formations provide isolation, and geotechnical barriers reestablish impermeability after waste is placed in the geology. Between excavation and closure, physical, mechanical, thermal, chemical, and hydrological processes ensue. Positive attributes for isolation in salt have many commonalities independent of the geologic setting. In some cases, specific details of the environment will affect the disposal concept and thereby define interaction of features, events and processes, while simultaneously influencing scenario development. Here we identify and discuss high-level differences and similarities of bedded and domal salt formations. Positive geologic and engineering attributes for disposal purposes are more common among salt formations than are significant differences

Considerations of the Differences between Bedded and Domal Salt Pertaining to Disposal of Heat-Generating Nuclear Waste

International Nuclear Information System (INIS)

Hansen, Francis D.; Kuhlman, Kristopher L.; Sobolik, Steven R.

2016-01-01

Salt formations hold promise for eternal removal of nuclear waste from our biosphere. Germany and the United States have ample salt formations for this purpose, ranging from flat-bedded formations to geologically mature dome structures. As both nations revisit nuclear waste disposal options, the choice between bedded, domal, or intermediate pillow formations is once again a contemporary issue. For decades, favorable attributes of salt as a disposal medium have been extoled and evaluated, carefully and thoroughly. Yet, a sense of discovery continues as science and engineering interrogate naturally heterogeneous systems. Salt formations are impermeable to fluids. Excavation-induced fractures heal as seal systems are placed or natural closure progresses toward equilibrium. Engineering required for nuclear waste disposal gains from mining and storage industries, as humans have been mining salt for millennia. This great intellectual warehouse has been honed and distilled, but not perfected, for all nuances of nuclear waste disposal. Nonetheless, nations are able and have already produced suitable license applications for radioactive waste disposal in salt. A remaining conundrum is site location. Salt formations provide isolation, and geotechnical barriers reestablish impermeability after waste is placed in the geology. Between excavation and closure, physical, mechanical, thermal, chemical, and hydrological processes ensue. Positive attributes for isolation in salt have many commonalities independent of the geologic setting. In some cases, specific details of the environment will affect the disposal concept and thereby define interaction of features, events and processes, while simultaneously influencing scenario development. Here we identify and discuss high-level differences and similarities of bedded and domal salt formations. Positive geologic and engineering attributes for disposal purposes are more common among salt formations than are significant differences

Evaluation of iron-base materials for waste package containers in a salt repository

International Nuclear Information System (INIS)

Westerman, R.E.; Nelson, J.L.; Kuhn, W.L.; Basham, S.G.; Moak, D.A.; Pitman, S.G.

1983-11-01

Design studies for high-level nuclear waste packages for salt repositories have identified low-carbon steel as a candidate material for containers. Among the requirements are strength, corrosion resistance, and fabricability. The studies of the corrosion resistance and structural stability of iron-base materials (particularly low-carbon steel) are treated in this paper. The materials have been exposed in brines that are characteristic of the potential sites for salt repositories. The effects of temperature, radiation level, oxygen level and other parameters are under investigation. The initial development of corrosion models for these environments is presented with discussion of the key mechanisms under consideration. 6 references, 5 figures

Managing the nation's commercial high-level radioactive waste

International Nuclear Information System (INIS)

1985-03-01

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

Granite disposal of U.S. high-level radioactive waste.

Energy Technology Data Exchange (ETDEWEB)

Freeze, Geoffrey A.; Mariner, Paul E.; Lee, Joon H.; Hardin, Ernest L.; Goldstein, Barry; Hansen, Francis D.; Price, Ronald H.; Lord, Anna Snider

2011-08-01

This report evaluates the feasibility of disposing U.S. high-level radioactive waste in granite several hundred meters below the surface of the earth. The U.S. has many granite formations with positive attributes for permanent disposal. Similar crystalline formations have been extensively studied by international programs, two of which, in Sweden and Finland, are the host rocks of submitted or imminent repository license applications. This report is enabled by the advanced work of the international community to establish functional and operational requirements for disposal of a range of waste forms in granite media. In this report we develop scoping performance analyses, based on the applicable features, events, and processes (FEPs) identified by international investigators, to support generic conclusions regarding post-closure safety. Unlike the safety analyses for disposal in salt, shale/clay, or deep boreholes, the safety analysis for a mined granite repository depends largely on waste package preservation. In crystalline rock, waste packages are preserved by the high mechanical stability of the excavations, the diffusive barrier of the buffer, and favorable chemical conditions. The buffer is preserved by low groundwater fluxes, favorable chemical conditions, backfill, and the rigid confines of the host rock. An added advantage of a mined granite repository is that waste packages would be fairly easy to retrieve, should retrievability be an important objective. The results of the safety analyses performed in this study are consistent with the results of comprehensive safety assessments performed for sites in Sweden, Finland, and Canada. They indicate that a granite repository would satisfy established safety criteria and suggest that a small number of FEPs would largely control the release and transport of radionuclides. In the event the U.S. decides to pursue a potential repository in granite, a detailed evaluation of these FEPs would be needed to inform site

Economics of defense high level waste management in the United States

International Nuclear Information System (INIS)

McDonell, W.R.

1987-01-01

Life-cycle costs of defense waste disposal, as presented in the foregoing sections, are summarized. Expressed as incremental costs per canister of waste deposited in a Federal geologic repository and per gallon of decontaminated salt solution immobilized in onsite concrete vaults, the tabulated values provide a measure of waste management costs relatively independent of the inventories of waste processed. Total values are about $350,000 per glass waste canister processed and $4.68 per gallon of decontaminated salt immobilized. These costs do not generally include contributions of fixed charges, such as capital costs, except in the case of transport and repository charges for which the quantities of waste handled determine allocation of fixed costs included in the fee assessments. 14 refs., 2 figs., 3 tabs

Technetium Chemistry in High-Level Waste

International Nuclear Information System (INIS)

Hess, Nancy J.

2006-01-01

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

Field experiments in salt formations

International Nuclear Information System (INIS)

Kuehn, K.

1986-01-01

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

Techniques for the solidification of high-level wastes

International Nuclear Information System (INIS)

1977-01-01

The problem of the long-term management of the high-level wastes from the reprocessing of irradiated nuclear fuel is receiving world-wide attention. While the majority of the waste solutions from the reprocessing of commercial fuels are currently being stored in stainless-steel tanks, increasing effort is being devoted to developing technology for the conversion of these wastes into solids. A number of full-scale solidification facilities are expected to come into operation in the next decade. The object of this report is to survey and compare all the work currently in progress on the techniques available for the solidification of high-level wastes. It will examine the high-level liquid wastes arising from the various processes currently under development or in operation, the advantages and disadvantages of each process for different types and quantities of waste solutions, the stages of development, the scale-up potential and flexibility of the processes

Tetraphenylborate Solubility in High Ionic Strength Salt Solutions

International Nuclear Information System (INIS)

Serkiz, S.M.; Ginn, J.D.; Jurgensen, A.R.

1998-04-01

Solubility of sodium and potassium salts of the tetraphenylborate ion (TPB) in simulated Savannah River Site High Level Waste was investigated. Data generated from this study allow more accurate predictions of TPB solubility at the In-Tank Precipitation (ITP) facility. Because previous research showed large deviations in the observed solubility of TPB salts when compared with model predictions, additional data were generated to better understand the solubility of TPB in more complex systems of high ionic strength and those containing both potassium and sodium. These data allow evaluation of the ability of current models to accurately predict equilibrium TPB concentrations over the range of experimental conditions investigated in this study

Protein removal from waste brines generated during ham salting through acidification and centrifugation.

Science.gov (United States)

Gutiérrez-Martínez, Maria del Rosario; Muñoz-Guerrero, Hernán; Alcaína-Miranda, Maria Isabel; Barat, José Manuel

2014-03-01

The salting step in food processes implies the production of large quantities of waste brines, having high organic load, high conductivity, and other pollutants with high oxygen demand. Direct disposal of the residual brine implies salinization of soil and eutrophication of water. Since most of the organic load of the waste brines comes from proteins leaked from the salted product, precipitation of dissolved proteins by acidification and removal by centrifugation is an operation to be used in waste brine cleaning. The aim of this study is optimizing the conditions for carrying out the separation of proteins from waste brines generated in the pork ham salting operation, by studying the influence of pH, centrifugal force, and centrifugation time. Models for determining the removal of proteins depending on the pH, centrifugal force, and time were obtained. The results showed a high efficacy of the proposed treatment for removing proteins, suggesting that this method could be used for waste brine protein removal. The best pH value to be used in an industrial process seems to be 3, while the obtained results indicate that almost 90% of the proteins from the brine can be removed by acidification followed by centrifugation. A further protein removal from the brine should have to be achieved using filtrating techniques, which efficiency could be highly improved as a consequence of the previous treatment through acidification and centrifugation. Waste brines from meat salting have high organic load and electrical conductivity. Proteins can be removed from the waste brine by acidification and centrifugation. The total protein removal can be up to 90% of the initial content of the waste brine. Protein removal is highly dependent on pH, centrifugation rate, and time. © 2014 Institute of Food Technologists®

Roadmap for disposal of Electrorefiner Salt as Transuranic Waste.

Energy Technology Data Exchange (ETDEWEB)

Rechard, Robert P. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Trone, Janis R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Kalinina, Elena Arkadievna [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Wang, Yifeng [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Hadgu, Teklu [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sanchez, Lawrence C. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2017-12-01

The experimental breeder reactor (EBR-II) used fuel with a layer of sodium surrounding the uranium-zirconium fuel to improve heat transfer. Disposing of EBR-II fuel in a geologic repository without treatment is not prudent because of the potentially energetic reaction of the sodium with water. In 2000, the US Department of Energy (DOE) decided to treat the sodium-bonded fuel with an electrorefiner (ER), which produces metallic uranium product, a metallic waste, mostly from the cladding, and the salt waste in the ER, which contains most of the actinides and fission products. Two waste forms were proposed for disposal in a mined repository; the metallic waste, which was to be cast into ingots, and the ER salt waste, which was to be further treated to produce a ceramic waste form. However, alternative disposal pathways for metallic and salt waste streams may reduce the complexity. For example, performance assessments show that geologic repositories can easily accommodate the ER salt waste without treating it to form a ceramic waste form. Because EBR-II was used for atomic energy defense activities, the treated waste likely meets the definition of transuranic waste. Hence, disposal at the Waste Isolation Pilot Plant (WIPP) in southern New Mexico, may be feasible. This report reviews the direct disposal pathway for ER salt waste and describes eleven tasks necessary for implementing disposal at WIPP, provided space is available, DOE decides to use this alternative disposal pathway in an updated environmental impact statement, and the State of New Mexico grants permission.

Timing of High-level Waste Disposal

International Nuclear Information System (INIS)

2008-01-01

This study identifies key factors influencing the timing of high-level waste (HLW) disposal and examines how social acceptability, technical soundness, environmental responsibility and economic feasibility impact on national strategies for HLW management and disposal. Based on case study analyses, it also presents the strategic approaches adopted in a number of national policies to address public concerns and civil society requirements regarding long-term stewardship of high-level radioactive waste. The findings and conclusions of the study confirm the importance of informing all stakeholders and involving them in the decision-making process in order to implement HLW disposal strategies successfully. This study will be of considerable interest to nuclear energy policy makers and analysts as well as to experts in the area of radioactive waste management and disposal. (author)

Multipurpose optimization models for high level waste vitrification

International Nuclear Information System (INIS)

Hoza, M.

1994-08-01

Optimal Waste Loading (OWL) models have been developed as multipurpose tools for high-level waste studies for the Tank Waste Remediation Program at Hanford. Using nonlinear programming techniques, these models maximize the waste loading of the vitrified waste and optimize the glass formers composition such that the glass produced has the appropriate properties within the melter, and the resultant vitrified waste form meets the requirements for disposal. The OWL model can be used for a single waste stream or for blended streams. The models can determine optimal continuous blends or optimal discrete blends of a number of different wastes. The OWL models have been used to identify the most restrictive constraints, to evaluate prospective waste pretreatment methods, to formulate and evaluate blending strategies, and to determine the impacts of variability in the wastes. The OWL models will be used to aid in the design of frits and the maximize the waste in the glass for High-Level Waste (HLW) vitrification

Molten salt treatment to minimize and optimize waste

International Nuclear Information System (INIS)

Gat, U.; Crosley, S.M.; Gay, R.L.

1993-01-01

A combination molten salt oxidizer (MSO) and molten salt reactor (MSR) is described for treatment of waste. The MSO is proposed for contained oxidization of organic hazardous waste, for reduction of mass and volume of dilute waste by evaporation of the water. The NTSO residue is to be treated to optimize the waste in terms of its composition, chemical form, mixture, concentration, encapsulation, shape, size, and configuration. Accumulations and storage are minimized, shipments are sized for low risk. Actinides, fissile material, and long-lived isotopes are separated and completely burned or transmuted in an MSR. The MSR requires no fuel element fabrication, accepts the materials as salts in arbitrarily small quantities enhancing safety, security, and overall acceptability

Waste salt recovery, recycle, and destruction

International Nuclear Information System (INIS)

Hickman, R.G.

1992-12-01

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

Molten salt hazardous waste disposal process utilizing gas/liquid contact for salt recovery

International Nuclear Information System (INIS)

Grantham, L.F.; McKenzie, D.E.

1984-01-01

The products of a molten salt combustion of hazardous wastes are converted into a cooled gas, which can be filtered to remove hazardous particulate material, and a dry flowable mixture of salts, which can be recycled for use in the molten salt combustion, by means of gas/liquid contact between the gaseous products of combustion of the hazardous waste and a solution produced by quenching the spent melt from such molten salt combustion. The process results in maximizing the proportion of useful materials recovered from the molten salt combustion and minimizing the volume of material which must be discarded. In a preferred embodiment a spray dryer treatment is used to achieve the desired gas/liquid contact

Concept development for saltstone and low level waste disposal

International Nuclear Information System (INIS)

Wilhite, E.L.

1987-03-01

A low-level alkaline salt solution will be a byproduct in the processing of high-level waste at the Savannah River Plant (SRP). This solution will be incorporated into a cement wasteform, saltstone, and placed in surface vaults. Laboratory and field testing and mathematical modeling have demonstrated the predictability of contaminant release from cement wasteforms. Saltstone disposal in surface vaults will meet drinking water standards in shallow groundwater at the disposal area boundary. Planning for new Low-Level Waste (LLW) disposal could incorporate concepts developed for saltstone disposal

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

International Nuclear Information System (INIS)

Kerr, R.A.

1979-01-01

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

Test plan for immobilization of salt-containing surrogate mixed wastes using polyester resins

International Nuclear Information System (INIS)

Biyani, R.K.; Douglas, J.C.; Hendrickson, D.W.

1997-01-01

Past operations at many Department of Energy (DOE) sites have resulted in the generation of several waste streams with high salt content. These wastes contain listed and characteristic hazardous constituents and are radioactive. The salts contained in the wastes are primarily chloride, sulfate, nitrate, metal oxides, and hydroxides. DOE has placed these types of wastes under the purview of the Mixed Waste Focus Area (MWFA). The MWFA has been tasked with developing and facilitating the implementation of technologies to treat these wastes in support of customer needs and requirements. The MWFA has developed a Technology Development Requirements Document (TDRD), which specifies performance requirements for technology owners and developers to use as a framework in developing effective waste treatment solutions. This project will demonstrate the use of polyester resins in encapsulating and solidifying DOE's mixed wastes containing salts, as an alternative to conventional and other emerging immobilization technologies

Test plan for immobilization of salt-containing surrogate mixed wastes using polyester resins

Energy Technology Data Exchange (ETDEWEB)

Biyani, R.K.; Douglas, J.C.; Hendrickson, D.W.

1997-07-07

Past operations at many Department of Energy (DOE) sites have resulted in the generation of several waste streams with high salt content. These wastes contain listed and characteristic hazardous constituents and are radioactive. The salts contained in the wastes are primarily chloride, sulfate, nitrate, metal oxides, and hydroxides. DOE has placed these types of wastes under the purview of the Mixed Waste Focus Area (MWFA). The MWFA has been tasked with developing and facilitating the implementation of technologies to treat these wastes in support of customer needs and requirements. The MWFA has developed a Technology Development Requirements Document (TDRD), which specifies performance requirements for technology owners and developers to use as a framework in developing effective waste treatment solutions. This project will demonstrate the use of polyester resins in encapsulating and solidifying DOE`s mixed wastes containing salts, as an alternative to conventional and other emerging immobilization technologies.

Expected environment for waste packages in a salt repository

International Nuclear Information System (INIS)

Pederson, L.R.; Clark, D.E.; Hodges, F.N.; McVay, G.L.; Rai, D.

1983-01-01

This paper discusses results of recent efforts to define the very near-field (within approximately 2 m) environmental conditions to which waste packages will be exposed in a salt repository. These conditions must be considered in the experimental design for waste package materials testing, which includes corrosion of barrier materials and leaching of waste forms. Site-specific brine compositions have been determined, and standard brine compositions have been selected for testing purposes. Actual brine compositions will vary depending on origin, temperature, irradiation history, and contact with irradiated rock salt. Results of irradiating rock salt, synthetic brines, rock salt/brine mixtures, and reactions of irradiated rock salt with brine solutions are reported. 38 references, 3 figures, 2 tables

High-level waste-form-product performance evaluation

International Nuclear Information System (INIS)

Bernadzikowski, T.A.; Allender, J.S.; Stone, J.A.; Gordon, D.E.; Gould, T.H. Jr.; Westberry, C.F. III.

1982-01-01

Seven candidate waste forms were evaluated for immobilization and geologic disposal of high-level radioactive wastes. The waste forms were compared on the basis of leach resistance, mechanical stability, and waste loading. All forms performed well at leaching temperatures of 40, 90, and 150 0 C. Ceramic forms ranked highest, followed by glasses, a metal matrix form, and concrete. 11 tables

Molten salt processing of mixed wastes with offgas condensation

International Nuclear Information System (INIS)

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

1991-01-01

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

Properties and characteristics of high-level waste glass

International Nuclear Information System (INIS)

Ross, W.A.

1977-01-01

This paper has briefly reviewed many of the characteristics and properties of high-level waste glasses. From this review, it can be noted that glass has many desirable properties for solidification of high-level wastes. The most important of these include: (1) its low leach rate; (2) the ability to tolerate large changes in waste composition; (3) the tolerance of anticipated storage temperatures; (4) its low surface area even after thermal shock or impact

High-level waste management technology program plan

Energy Technology Data Exchange (ETDEWEB)

Harmon, H.D.

1995-01-01

The purpose of this plan is to document the integrated technology program plan for the Savannah River Site (SRS) High-Level Waste (HLW) Management System. The mission of the SRS HLW System is to receive and store SRS high-level wastes in a see and environmentally sound, and to convert these wastes into forms suitable for final disposal. These final disposal forms are borosilicate glass to be sent to the Federal Repository, Saltstone grout to be disposed of on site, and treated waste water to be released to the environment via a permitted outfall. Thus, the technology development activities described herein are those activities required to enable successful accomplishment of this mission. The technology program is based on specific needs of the SRS HLW System and organized following the systems engineering level 3 functions. Technology needs for each level 3 function are listed as reference, enhancements, and alternatives. Finally, FY-95 funding, deliverables, and schedules are s in Chapter IV with details on the specific tasks that are funded in FY-95 provided in Appendix A. The information in this report represents the vision of activities as defined at the beginning of the fiscal year. Depending on emergent issues, funding changes, and other factors, programs and milestones may be adjusted during the fiscal year. The FY-95 SRS HLW technology program strongly emphasizes startup support for the Defense Waste Processing Facility and In-Tank Precipitation. Closure of technical issues associated with these operations has been given highest priority. Consequently, efforts on longer term enhancements and alternatives are receiving minimal funding. However, High-Level Waste Management is committed to participation in the national Radioactive Waste Tank Remediation Technology Focus Area. 4 refs., 5 figs., 9 tabs.

High-level waste management technology program plan

International Nuclear Information System (INIS)

Harmon, H.D.

1995-01-01

The purpose of this plan is to document the integrated technology program plan for the Savannah River Site (SRS) High-Level Waste (HLW) Management System. The mission of the SRS HLW System is to receive and store SRS high-level wastes in a see and environmentally sound, and to convert these wastes into forms suitable for final disposal. These final disposal forms are borosilicate glass to be sent to the Federal Repository, Saltstone grout to be disposed of on site, and treated waste water to be released to the environment via a permitted outfall. Thus, the technology development activities described herein are those activities required to enable successful accomplishment of this mission. The technology program is based on specific needs of the SRS HLW System and organized following the systems engineering level 3 functions. Technology needs for each level 3 function are listed as reference, enhancements, and alternatives. Finally, FY-95 funding, deliverables, and schedules are s in Chapter IV with details on the specific tasks that are funded in FY-95 provided in Appendix A. The information in this report represents the vision of activities as defined at the beginning of the fiscal year. Depending on emergent issues, funding changes, and other factors, programs and milestones may be adjusted during the fiscal year. The FY-95 SRS HLW technology program strongly emphasizes startup support for the Defense Waste Processing Facility and In-Tank Precipitation. Closure of technical issues associated with these operations has been given highest priority. Consequently, efforts on longer term enhancements and alternatives are receiving minimal funding. However, High-Level Waste Management is committed to participation in the national Radioactive Waste Tank Remediation Technology Focus Area. 4 refs., 5 figs., 9 tabs

High-level radioactive wastes

International Nuclear Information System (INIS)

Grissom, M.C.

1982-10-01

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

Materials Science of High-Level Nuclear Waste Immobilization

International Nuclear Information System (INIS)

Weber, William J.; Navrotsky, Alexandra; Stefanovsky, S. V.; Vance, E. R.; Vernaz, Etienne Y.

2009-01-01

With the increasing demand for the development of more nuclear power comes the responsibility to address the technical challenges of immobilizing high-level nuclear wastes in stable solid forms for interim storage or disposition in geologic repositories. The immobilization of high-level nuclear wastes has been an active area of research and development for over 50 years. Borosilicate glasses and complex ceramic composites have been developed to meet many technical challenges and current needs, although regulatory issues, which vary widely from country to country, have yet to be resolved. Cooperative international programs to develop advanced proliferation-resistant nuclear technologies to close the nuclear fuel cycle and increase the efficiency of nuclear energy production might create new separation waste streams that could demand new concepts and materials for nuclear waste immobilization. This article reviews the current state-of-the-art understanding regarding the materials science of glasses and ceramics for the immobilization of high-level nuclear waste and excess nuclear materials and discusses approaches to address new waste streams

Performance Assessment of a Generic Repository in Bedded Salt for DOE-Managed Nuclear Waste

Science.gov (United States)

Stein, E. R.; Sevougian, S. D.; Hammond, G. E.; Frederick, J. M.; Mariner, P. E.

2016-12-01

A mined repository in salt is one of the concepts under consideration for disposal of DOE-managed defense-related spent nuclear fuel (SNF) and high level waste (HLW). Bedded salt is a favorable medium for disposal of nuclear waste due to its low permeability, high thermal conductivity, and ability to self-heal. Sandia's Generic Disposal System Analysis framework is used to assess the ability of a generic repository in bedded salt to isolate radionuclides from the biosphere. The performance assessment considers multiple waste types of varying thermal load and radionuclide inventory, the engineered barrier system comprising the waste packages, backfill, and emplacement drifts, and the natural barrier system formed by a bedded salt deposit and the overlying sedimentary sequence (including an aquifer). The model simulates disposal of nearly the entire inventory of DOE-managed, defense-related SNF (excluding Naval SNF) and HLW in a half-symmetry domain containing approximately 6 million grid cells. Grid refinement captures the detail of 25,200 individual waste packages in 180 disposal panels, associated access halls, and 4 shafts connecting the land surface to the repository. Equations describing coupled heat and fluid flow and reactive transport are solved numerically with PFLOTRAN, a massively parallel flow and transport code. Simulated processes include heat conduction and convection, waste package failure, waste form dissolution, radioactive decay and ingrowth, sorption, solubility limits, advection, dispersion, and diffusion. Simulations are run to 1 million years, and radionuclide concentrations are observed within an aquifer at a point approximately 4 kilometers downgradient of the repository. The software package DAKOTA is used to sample likely ranges of input parameters including waste form dissolution rates and properties of engineered and natural materials in order to quantify uncertainty in predicted concentrations and sensitivity to input parameters. Sandia

Demonstration of pyropartitioning process by using genuine high-level liquid waste. Reductive-extraction of actinide elements from chlorination product

International Nuclear Information System (INIS)

Uozumi, Koichi; Iizuka, Masatoshi; Kurata, Masaki; Ougier, Michel; Malmbeck, Rikard; Winckel, Stefaan van

2009-01-01

The pyropartitioning process separates the minor actinide elements (MAs) together with uranium and plutonium from the high-level liquid waste generated at the Purex reprocessing of spent LWR fuel and introduces them to metallic fuel cycle. For the demonstration of this technology, a series experiment using 520g of genuine high-level liquid waste was started and the conversion of actinide elements to their chlorides was already demonstrated by denitration and chlorination. In the present study, a reductive extraction experiment in molten salt/liquid cadmium system to recover actinide elements from the chlorination product of the genuine high-level liquid waste was performed. The results of the experiment are as following; 1) By the addition of the cadmium-lithium alloy reductant, almost all of plutonium and MAs in the initial high-level liquid waste were recovered in the cadmium phase. It means no mass loss during denitration, chlorination, and reductive-extraction. 2) The separation factor values of plutonium, MAs, and rare-earth fission product elements versus uranium agreed with the literature values. Therefore, actinide elements will be separated from fission product elements in the actual system. Hence, the pyropartitioning process was successfully demonstrated. (author)

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

International Nuclear Information System (INIS)

1987-05-01

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

Studies of corrosion in metallic container for storage of high level radioactive wastes

International Nuclear Information System (INIS)

Azkarate, I.; Madina, V.; Insausti, M.

1999-01-01

The metallic container is one of the most important barriers that, along with engineered and natural barriers, will isolate high level nuclear waste in saline and granite geological formations from the geosphere. However, general and localized corrosion modes such as stress corrosion cracking (SCC), pitting, crevice corrosion and hydrogen damage can be active under disposal conditions, so the corrosion behaviour of the metal container material must be carefully studied. Several metals and their alloys have been proposed for the fabrication of nuclear waste containers including carbon steels, stainless steels, titanium and titanium alloys and copper and copper-base alloys. Carbon steels and copper alloys are considered for the two rock formations, titanium is considered for salt environments and the stainless steel only in the case of a granite formation. (Author)

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

International Nuclear Information System (INIS)

Vinjamuri, K.

1993-06-01

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

Final report on cermet high-level waste forms

International Nuclear Information System (INIS)

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

1981-08-01

Cermets are being developed as an alternate method for the fixation of defense and commercial high level radioactive waste in a terminal disposal form. Following initial feasibility assessments of this waste form, consisting of ceramic particles dispersed in an iron-nickel base alloy, significantly improved processing methods were developed. The characterization of cermets has continued through property determinations on samples prepared by various methods from a variety of simulated and actual high-level wastes. This report describes the status of development of the cermet waste form as it has evolved since 1977. 6 tables, 18 figures

Methods of calculating the post-closure performance of high-level waste repositories

Energy Technology Data Exchange (ETDEWEB)

Ross, B. (ed.)

1989-02-01

This report is intended as an overview of post-closure performance assessment methods for high-level radioactive waste repositories and is designed to give the reader a broad sense of the state of the art of this technology. As described here, ''the state of the art'' includes only what has been reported in report, journal, and conference proceedings literature through August 1987. There is a very large literature on the performance of high-level waste repositories. In order to make a review of this breadth manageable, its scope must be carefully defined. The essential principle followed is that only methods of calculating the long-term performance of waste repositories are described. The report is organized to reflect, in a generalized way, the logical order to steps that would be taken in a typical performance assessment. Chapter 2 describes ways of identifying scenarios and estimating their probabilities. Chapter 3 presents models used to determine the physical and chemical environment of a repository, including models of heat transfer, radiation, geochemistry, rock mechanics, brine migration, radiation effects on chemistry, and coupled processes. The next two chapters address the performance of specific barriers to release of radioactivity. Chapter 4 treats engineered barriers, including containers, waste forms, backfills around waste packages, shaft and borehole seals, and repository design features. Chapter 5 discusses natural barriers, including ground water systems and stability of salt formations. The final chapters address optics of general applicability to performance assessment models. Methods of sensitivity and uncertainty analysis are described in Chapter 6, and natural analogues of repositories are treated in Chapter 7. 473 refs., 19 figs., 2 tabs.

Methods of calculating the post-closure performance of high-level waste repositories

International Nuclear Information System (INIS)

Ross, B.

1989-02-01

This report is intended as an overview of post-closure performance assessment methods for high-level radioactive waste repositories and is designed to give the reader a broad sense of the state of the art of this technology. As described here, ''the state of the art'' includes only what has been reported in report, journal, and conference proceedings literature through August 1987. There is a very large literature on the performance of high-level waste repositories. In order to make a review of this breadth manageable, its scope must be carefully defined. The essential principle followed is that only methods of calculating the long-term performance of waste repositories are described. The report is organized to reflect, in a generalized way, the logical order to steps that would be taken in a typical performance assessment. Chapter 2 describes ways of identifying scenarios and estimating their probabilities. Chapter 3 presents models used to determine the physical and chemical environment of a repository, including models of heat transfer, radiation, geochemistry, rock mechanics, brine migration, radiation effects on chemistry, and coupled processes. The next two chapters address the performance of specific barriers to release of radioactivity. Chapter 4 treats engineered barriers, including containers, waste forms, backfills around waste packages, shaft and borehole seals, and repository design features. Chapter 5 discusses natural barriers, including ground water systems and stability of salt formations. The final chapters address optics of general applicability to performance assessment models. Methods of sensitivity and uncertainty analysis are described in Chapter 6, and natural analogues of repositories are treated in Chapter 7. 473 refs., 19 figs., 2 tabs

High-Level Waste System Process Interface Description

International Nuclear Information System (INIS)

D'Entremont, P.D.

1999-01-01

The High-Level Waste System is a set of six different processes interconnected by pipelines. These processes function as one large treatment plant that receives, stores, and treats high-level wastes from various generators at SRS and converts them into forms suitable for final disposal. The three major forms are borosilicate glass, which will be eventually disposed of in a Federal Repository, Saltstone to be buried on site, and treated water effluent that is released to the environment

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

Cermet high level waste forms: a pregress report

International Nuclear Information System (INIS)

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

1978-06-01

The fixation of high level radioactive waste from both commercial and DOE defense sources as cermets is currently under study. This waste form consists of a continuous iron-nickel base metal matrix containing small particles of fission product oxides. Preliminary evaluations of cermets fabricated from a variety of simulated wastes indicate they possess properties providing advantages over other waste forms presently being considered, namely thermal conductivity, waste loading levels, and leach resistance. This report describes the progress of this effort, to date, since its initiation in 1977

Membrane Treatment of Liquid Salt Bearing Radioactive Wastes

International Nuclear Information System (INIS)

Dmitriev, S. A.; Adamovich, D. V.; Demkin, V. I.; Timofeev, E. M.

2003-01-01

The main fields of introduction and application of membrane methods for preliminary treatment and processing salt liquid radioactive waste (SLRW) can be nuclear power stations (NPP) and enterprises on atomic submarines (AS) utilization. Unlike the earlier developed technology for the liquid salt bearing radioactive waste decontamination and concentrating this report presents the new enhanced membrane technology for the liquid salt bearing radioactive waste processing based on the state-of-the-art membrane unit design, namely, the filtering units equipped with the metal-ceramic membranes of ''TruMem'' brand, as well as the electrodialysis and electroosmosis concentrators. Application of the above mentioned units in conjunction with the pulse pole changer will allow the marked increase of the radioactive waste concentrating factor and the significant reduction of the waste volume intended for conversion into monolith and disposal. Besides, the application of the electrodialysis units loaded with an ion exchange material at the end polishing stage of the radioactive waste decontamination process will allow the reagent-free radioactive waste treatment that meets the standards set for the release of the decontaminated liquid radioactive waste effluents into the natural reservoirs of fish-farming value

Treatment technologies for non-high-level wastes (USA)

International Nuclear Information System (INIS)

Cooley, C.R.; Clark, D.E.

1976-06-01

Non-high-level waste arising from operations at nuclear reactors, fuel fabrication facilities, and reprocessing facilities can be treated using one of several technical alternatives prior to storage. Each alternative and the associated experience and status of development are summarized. The technology for treating non-high-level wastes is generally available for industrial use. Improved techniques applicable to the commercial nuclear fuel cycle are being developed and demonstrated to reduce the volume of waste and to immobilize it for storage. 36 figures, 59 references

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

Impact of Alkali Source on Vitrification of SRS High Level Waste

International Nuclear Information System (INIS)

LAMBERT, D. P.; MILLER, D. H.; PEELER, D. K.; SMITH, M. E.; STONE, M. E.

2005-01-01

The Defense Waste Processing Facility (DWPF) Savannah River Site is currently immobilizing high level nuclear waste sludge by vitrification in borosilicate glass. The processing strategy involves blending a large batch of sludge into a feed tank, washing the sludge to reduce the amount of soluble species, then processing the large ''sludge batch'' through the DWPF. Each sludge batch is tested by the Savannah River National Laboratory (SRNL) using simulants and tests with samples of the radioactive waste to ''qualify'' the batch prior to processing in the DWPF. The DWPF pretreats the sludge by first acidifying the sludge with nitric and formic acid. The ratio of nitric to formic acid is adjusted as required to target a final glass composition that is slightly reducing (the target is for ∼20% of the iron to have a valence of two in the glass). The formic acid reduces the mercury in the feed to elemental mercury which is steam stripped from the feed. After a concentration step, the glass former (glass frit) is added as a 50 wt% slurry and the batch is concentrated to approximately 50 wt% solids. The feed slurry is then fed to a joule heated melter maintained at 1150 C. The glass must meet both processing (e.g., viscosity and liquidus temperature) and product performance (e.g., durability) constraints The alkali content of the final waste glass is a critical parameter that affects key glass properties (such as durability) as well as the processing characteristics of the waste sludge during the pretreatment and vitrification processes. Increasing the alkali content of the glass has been shown to improve the production rate of the DWPF, but the total alkali in the final glass is limited by constraints on glass durability and viscosity. Two sources of alkali contribute to the final alkali content of the glass: sodium salts in the waste supernate and sodium and lithium oxides in the glass frit added during pretreatment processes. Sodium salts in the waste supernate can

Third report, development of site suitability and design performance data base for a high level nuclear waste repository

International Nuclear Information System (INIS)

1978-05-01

This report presents the results of mining and geotechnical studies provided as support for the development of a technical data base suitable for the generation of a regulatory framework governing high-level nuclear waste disposal: flow path models for basalts, salt domes, and crystalline rocks; simplistic model evaluations; loss of administrative control; repository conceptual designs; repository design considerations; and design performance criteria

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

International Nuclear Information System (INIS)

Decamps, F.

1993-01-01

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

Design and construction of the low-level liquid waste treatment system

International Nuclear Information System (INIS)

Baker, M.N.; Mateer, W.E.; Metzler, G.H.; Reeves, S.R.; Rickettson, D.J.

1989-03-01

This report describes the design and construction of the Low-Level Liquid Waste Treatment System (LWTS). The LWTS is part of a system that will prepare High-Level Radioactive Waste for solidification in glass. This preparation includes removal of water and salts from the stored waste. The topics addressed are: the design objective to reuse the Process Building to contain LWTS, the special considerations that arise when building a new system inside a decontaminated facility, interface to existing plant systems, phased construction, and construction testing. 8 refs., 24 figs

Prostaglandin-E2 Mediated Increase in Calcium and Phosphate Excretion in a Mouse Model of Distal Nephron Salt Wasting.

Directory of Open Access Journals (Sweden)

Manoocher Soleimani

Full Text Available Contribution of salt wasting and volume depletion to the pathogenesis of hypercalciuria and hyperphosphaturia is poorly understood. Pendrin/NCC double KO (pendrin/NCC-dKO mice display severe salt wasting under basal conditions and develop profound volume depletion, prerenal renal failure, and metabolic alkalosis and are growth retarded. Microscopic examination of the kidneys of pendrin/NCC-dKO mice revealed the presence of calcium phosphate deposits in the medullary collecting ducts, along with increased urinary calcium and phosphate excretion. Confirmatory studies revealed decreases in the expression levels of sodium phosphate transporter-2 isoforms a and c, increases in the expression of cytochrome p450 family 4a isotypes 12 a and b, as well as prostaglandin E synthase 1, and cyclooxygenases 1 and 2. Pendrin/NCC-dKO animals also had a significant increase in urinary prostaglandin E2 (PGE-2 and renal content of 20-hydroxyeicosatetraenoic acid (20-HETE levels. Pendrin/NCC-dKO animals exhibit reduced expression levels of the sodium/potassium/2chloride co-transporter 2 (NKCC2 in their medullary thick ascending limb. Further assessment of the renal expression of NKCC2 isoforms by quantitative real time PCR (qRT-PCR reveled that compared to WT mice, the expression of NKCC2 isotype F was significantly reduced in pendrin/NCC-dKO mice. Provision of a high salt diet to rectify volume depletion or inhibition of PGE-2 synthesis by indomethacin, but not inhibition of 20-HETE generation by HET0016, significantly improved hypercalciuria and salt wasting in pendrin/NCC dKO mice. Both high salt diet and indomethacin treatment also corrected the alterations in NKCC2 isotype expression in pendrin/NCC-dKO mice. We propose that severe salt wasting and volume depletion, irrespective of the primary originating nephron segment, can secondarily impair the reabsorption of salt and calcium in the thick ascending limb of Henle and/or proximal tubule, and reabsorption of

Treatment of waste salts by oxygen sparging and vacuum distillation

International Nuclear Information System (INIS)

Cho, Y.J.; Yang, H.C.; Kim, E.H.; Kin, I.T.; Eun, H.C.

2007-01-01

Full text of publication follows. During the electrorefining process of the oxide spent fuel from LWR, amounts of waste salts containing some metal chloride species such as rare earths and actinide chlorides are generated, where the reuse of the waste salts is very important from the standpoint of an economical as well as an environmental aspect. In order to reuse the waste salts, a salt vacuum distillation method can be used. For the best separation by a vacuum distillation, the metal chloride species involved in the waste salts must be converted into their oxide(or oxychloride) forms due to the their low volatility compared to that of LiCl-KCl. In this study, an oxygen sparging process was adopted for the oxidation (or precipitation) of rare earth chlorides. The effects of oxygen flow rate and molten salt temperature on the conversion of rare earth chlorides to the precipitate phase (i.e. oxide or oxychloride) were investigated. In addition, distillation characteristics of LiCl-KCl molten salt with system pressure and temperature were studied. (authors)

US program for the immobilization of high-level nuclear wastes

International Nuclear Information System (INIS)

Crandall, J.L.

1979-01-01

A program has been developed for long-term management of high-level nuclear waste. The Savannah River Operations Office of the US Department of Energy is acting as the lead office for this program with technical advice from the E.I. du Pont de Nemours and Company. The purpose of the long-term program is to immobilize the DOE high-level waste in forms that act as highly efficient barriers against radionuclide release to the disposal site and to provide technology for similar treatment of commercial high-level waste in case reprocessing of commercial nuclear fuels is ever resumed. Descriptions of existing DOE and commercial wastes, program strategy, program expenditures, development of waste forms, evaluation and selection of waste forms, regulatory aspects of waste form selection, project schedules, and cost estimates for immobilization facilities are discussed

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.

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

Engineering study of the potential uses of salts from selective crystallization of Hanford tank wastes

International Nuclear Information System (INIS)

Hendrickson, D.W.

1996-01-01

The Clean Salt Process (CSP) is the fractional crystallization of nitrate salts from tank waste stored on the Hanford Site. This study reviews disposition options for a CSP product made from Hanford Site tank waste. These options range from public release to onsite low-level waste disposal to no action. Process, production, safety, environment, cost, schedule, and the amount of CSP material which may be used are factors considered in each option. The preferred alternative is offsite release of clean salt. Savings all be generated by excluding the material from low-level waste stabilization. Income would be received from sales of salt products. Savings and income from this alternative amount to $1,027 million, excluding the cost of CSP operations. Unless public sale of CSP products is approved, the material should be calcined. The carbonate form of the CSP could then be used as ballast in tank closure and stabilization efforts. Not including the cost of CSP operations, savings of $632 million would be realized. These savings would result from excluding the material from low-level waste stabilization and reducing purchases of chemicals for caustic recycle and stabilization and closure. Dose considerations for either alternative are favorable. No other cost-effective alternatives that were considered had the capacity to handle significant quantities of the CSP products. If CSP occurs, full-scale tank-waste stabilization could be done without building additional treatment facilities after Phase 1 (DOE 1996). Savings in capital and operating cost from this reduction in waste stabilization would be in addition to the other gains described

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

International Nuclear Information System (INIS)

1978-04-01

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

Treatment of waste salt from the advanced spent fuel conditioning process (II) : optimum immobilization condition

International Nuclear Information System (INIS)

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

2004-01-01

Since zeolite is known to be stable at a high temperature, it has been reported as a promising immobilization matrix for waste salt. The crystal structure of dehydrated zeolite A breaks down above 1060 K, resulting in the formation of an amorphous solid and re-crystallization to beta-Cristobalite. This structural degradation depends on the existence of chlorides. When contacted to HCl, zeolite 4A is not stable even at 473 K. The optimum consolidation condition for LiCl salt waste from the oxide fuel reduction process based on the electrochemical method (Advanced spent fuel Conditioning Process; ACP) has been studied using zeolite A since 2001. Actually the constituents of waste salt are water-soluble. And, alkali halides are known to be readily radiolyzed to yield interstitial halogens and metal colloids. For disposal in a geological repository, the waste salt must meet the acceptance criteria. For a waste form containing chloride salt, two of the more important criteria are leach resistance and waste form durability. In this work, we prepared some samples with different mixing ratios of LiCl salt to zeolite A, and then compared some characteristics such as thermal stability, salt occlusion, free chloride content, leach resistance, mixing effect, etc

Supplemental technical information in support of Y/OWI/TM--44. Volume 17. Drawings for repository preconceptual design studies: BPNL waste forms in salt

International Nuclear Information System (INIS)

1978-04-01

Volume 17 contains drawings of a preconceptual design for a nuclear waste storage facility in salt. Three full cycles are considered: full recycle, throwaway cycle, and uranium recycle with plutonium in high-level waste

Design of off-gas cleaning systems for high-level waste vitrification

International Nuclear Information System (INIS)

Hanson, M.S.; Kaser, J.D.

1976-01-01

High-level wastes are generally nitric acid solutions. Vitrification converts the nitrate salts to oxides, forming nitrogen oxides (NO/sub x/) as a by-product. These NO/sub x/ releases can be controlled by nitric acid recovery or by conversion of the NO/sub x/ to an acceptable species for release, such as N 2 O or N 2 . The off-gas system must also be capable of controlling any fission products which may be voltatilized in appreciable quantities and may be controlled in the off-gas system by absorption or adsorption. Whichever method is used, the recovered fission products must somehow be converted to a safe disposal form. Proposed off-gas systems are described, and areas requiring research and development are discussed

Determination of total cyanide in Hanford Site high-level wastes

Energy Technology Data Exchange (ETDEWEB)

Winters, W.I. [Westinghouse Hanford Co., Richland, WA (United States); Pool, K.H. [Pacific Northwest Lab., Richland, WA (United States)

1994-05-01

Nickel ferrocyanide compounds (Na{sub 2-x}Cs{sub x}NiFe (CN){sub 6}) were produced in a scavenging process to remove {sup 137}Cs from Hanford Site single-shell tank waste supernates. Methods for determining total cyanide in Hanford Site high-level wastes are needed for the evaluation of potential exothermic reactions between cyanide and oxidizers such as nitrate and for safe storage, processing, and management of the wastes in compliance with regulatory requirements. Hanford Site laboratory experience in determining cyanide in high-level wastes is summarized. Modifications were made to standard cyanide methods to permit improved handling of high-level waste samples and to eliminate interferences found in Hanford Site waste matrices. Interferences and associated procedure modifications caused by high nitrates/nitrite concentrations, insoluble nickel ferrocyanides, and organic complexants are described.

Determination of total cyanide in Hanford Site high-level wastes

International Nuclear Information System (INIS)

Winters, W.I.; Pool, K.H.

1994-05-01

Nickel ferrocyanide compounds (Na 2-x Cs x NiFe (CN) 6 ) were produced in a scavenging process to remove 137 Cs from Hanford Site single-shell tank waste supernates. Methods for determining total cyanide in Hanford Site high-level wastes are needed for the evaluation of potential exothermic reactions between cyanide and oxidizers such as nitrate and for safe storage, processing, and management of the wastes in compliance with regulatory requirements. Hanford Site laboratory experience in determining cyanide in high-level wastes is summarized. Modifications were made to standard cyanide methods to permit improved handling of high-level waste samples and to eliminate interferences found in Hanford Site waste matrices. Interferences and associated procedure modifications caused by high nitrates/nitrite concentrations, insoluble nickel ferrocyanides, and organic complexants are described

Waste package materials testing for a salt repository: 1983 status summary report

International Nuclear Information System (INIS)

Moak, D.P.

1986-09-01

The United States plans to safely dispose of nuclear waste in deep, stable geologic formations. As part of these plans, the US Department of Energy is sponsoring research on the designing and testing of waste packages and waste package materials. This fiscal year 1983 status report summarizes recent results of waste package materials testing in a salt environment. The results from these tests will be used by waste package designers and performance assessment experts. Release characteristics data are available on two waste forms (spent fuel and waste-containing glass) that were exposed to leaching tests at various radiation levels, temperatures, pH, glass surface area to solution volume ratios, and brine solutions simulating expected salt repository conditions. Candidate materials tested for corrosion resistance and other properties include iron alloys; TI-CODE 12, the most promising titanium alloy for containment; and nickel alloys. In component interaction testing, synergistic effects have not ruled out any candidate material. 21 refs., 37 figs., 15 tabs

Assessment of lead tellurite glass for immobilizing electrochemical salt wastes from used nuclear fuel reprocessing

Energy Technology Data Exchange (ETDEWEB)

Riley, Brian J.; Kroll, Jared O.; Peterson, Jacob A.; Pierce, David A.; Ebert, William L.; Williams, Benjamin D.; Snyder, Michelle M. V.; Frank, Steven M.; George, Jaime L.; Kruska, Karen

2017-11-01

This paper provides an overview of research evaluating the use of tellurite glass as a waste form for salt wastes from electrochemical processing. The capacities to immobilize different salts were evaluated including: a LiCl-Li2O oxide reduction salt (for oxide fuel) containing fission products, a LiCl-KCl eutectic salt (for metallic fuel) containing fission products, and SrCl2. Physical and chemical properties of the glasses were characterized by using X-ray diffraction, bulk density measurements, chemical durability tests, scanning electron microscopy, and energy dispersive X-ray emission spectroscopy. These glasses were found to accommodate high concentrations of halide salts and have high densities. However, improvements are needed to meet chemical durability requirements.

Waste package reference conceptual designs for a repository in salt

International Nuclear Information System (INIS)

1986-02-01

This report provides the reference conceptual waste package designs for the Office of Nuclear Waste Isolation to baseline these designs, thereby establishing the configuration and interface controls necessary, within the Civilian Radioactive Waste Management Program, formerly the National Waste Terminal Storage Program, to proceed in an orderly manner with preliminary design. Included are designs for the current reference defense high-level waste form from the Savannah River Plant, an optimized commercial high-level waste form, and spent fuel which has been disassembled and compacted into a circular bundle containing either 12 pressurized-water reactor or 30 boiling-water reactor assemblies. For compacted spent fuel, it appears economically attractive to standardize the waste package diameter for all fuel types. The reference waste packages consist of the containerized waste form, a low carbon steel overpack, and, after emplacement, a cover of salt. The overpack is a hollow cylinder with a flat head welded to each end. Its design thickness is the sum of the structural thickness required to resist the 15.4-MPa lithostatic pressure plus the corrosion allowance necessary to assure the required structural thickness will exist through the 1000-year containment period. Based on available data and completed analyses, the reference concepts described in this report satisfy all requirements of the US Department of Energy and the US Nuclear Regulatory Commission with reasonable assurance. In addition, sufficient design maturity exists to form a basis for preliminary design; these concepts can be brought under configuration control to serve as reference package designs. Development programs are identified that will be required to support these designs during the licensing process. 19 refs., 37 figs., 31 tabs

Solidification of Savannah River Plant high level waste

International Nuclear Information System (INIS)

Maher, R.; Shafranek, L.F.; Kelley, J.A.; Zeyfang, R.W.

1981-11-01

Authorization for construction of the Defense Waste Processing Facility (DWPF) is expected in FY 83. The optimum time for stage 2 authorization is about three years later. Detailed design and construction will require approximately five years for stage 1, with stage 2 construction completed about two to three years later. Production of canisters of waste glass would begin in 1988, and the existing backlog of high level waste sludge stored at SRP would be worked off by about the year 2000. Stage 2 operation could begin in 1990. The technology and engineering are ready for construction and eventual operation of the DWPF for immobilizing high level radioactive waste at Savannah River Plant (SRP). Proceeding with this project will provide the public, and the leadership of this country, with a crucial demonstration that a major quantity of existing high level nuclear wastes can be safely and permanently immobilized. Early demonstration will both expedite and facilitate rational decision making on this aspect of the nuclear program. Delay in providing these facilities will result in significant DOE expenditures at SRP for new tanks just for continued temporary storage of wastes, and would probably result in dissipation of the intellectual and planning momentum that has built up in developing the project

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

High level waste at Hanford: Potential for waste loading maximization

International Nuclear Information System (INIS)

Hrma, P.R.; Bailey, A.W.

1995-09-01

The loading of Hanford nuclear waste in borosilicate glass is limited by phase-related phenomena, such as crystallization or formation of immiscible liquids, and by breakdown of the glass structure because of an excessive concentration of modifiers. The phase-related phenomena cause both processing and product quality problems. The deterioration of product durability determines the ultimate waste loading limit if all processing problems are resolved. Concrete examples and mass-balance based calculations show that a substantial potential exists for increasing waste loading of high-level wastes that contain a large fraction of refractory components

An HMS/TRAC analysis of a high-level radioactive waste tank

International Nuclear Information System (INIS)

Travis, J.R.; Nichols, B.D.; Spore, J.W.; Wilson, T.L.

1991-01-01

It has been observed that a high-level radioactive waste tank generates quantities of hydrogen and nitrous oxide mixtures that are potentially well within flammability limits. These gases are produced from chemical and nuclear decay reactions in a slurry of radioactive waste material. The slurry is covered by a thick crust composed of sodium nitrate and nitrite salts. Significant amounts of the combustible and reactant gases are produced over a 3- to 4-month period before the crust ruptures and the gases are vented into the air cover gas space above the crust. Postulating an ignition of the hydrogen/nitrous oxide/air mixture after this venting into the cover gas, we have calculated the pressure and temperature loading on the double-walled waste tank with the three-dimensional, time-dependent fluid dynamics coupled with chemical kinetics HMS (Hydrogen Mixing Studies) computer code. The waste tank has a ventilation system designed to maintain a slight negative gage pressure during steady-state operation. We have modeled the ventilation system with TRAC (the Transient Reactor Analysis Code), and we have coupled these two best-estimate accident analysis tools to provide the ventilation response to pressure and temperatures generated by the hydrogen burn. Significant pressures are produced by this event, and the threat to the tank's integrity currently is being evaluated. 3 refs., 4 figs

Decision Document for Heat Removal from High-Level Waste Tanks

International Nuclear Information System (INIS)

WILLIS, W.L.

2000-01-01

This document establishes the combination of design and operational configurations that will be used to provide heat removal from high-level waste tanks during Phase 1 waste feed delivery to prevent the waste temperature from exceeding tank safety requirement limits. The chosen method--to use the primary and annulus ventilation systems to remove heat from the high-level waste tanks--is documented herein

Review of high-level waste form properties

International Nuclear Information System (INIS)

Rusin, J.M.

1980-12-01

This report is a review of waste form options for the immobilization of high-level-liquid wastes from the nuclear fuel cycle. This review covers the status of international research and development on waste forms as of May 1979. Although the emphasis in this report is on waste form properties, process parameters are discussed where they may affect final waste form properties. A summary table is provided listing properties of various nuclear waste form options. It is concluded that proposed waste forms have properties falling within a relatively narrow range. In regard to crystalline versus glass waste forms, the conclusion is that either glass of crystalline materials can be shown to have some advantage when a single property is considered; however, at this date no single waste form offers optimum properties over the entire range of characteristics investigated. A long-term effort has been applied to the development of glass and calcine waste forms. Several additional waste forms have enough promise to warrant continued research and development to bring their state of development up to that of glass and calcine. Synthetic minerals, the multibarrier approach with coated particles in a metal matrix, and high pressure-high temperature ceramics offer potential advantages and need further study. Although this report discusses waste form properties, the total waste management system should be considered in the final selection of a waste form option. Canister design, canister materials, overpacks, engineered barriers, and repository characteristics, as well as the waste form, affect the overall performance of a waste management system. These parameters were not considered in this comparison

Thermo-mechanical analysis of high level nuclear wastes in granite

International Nuclear Information System (INIS)

Millard, A.; Guri, G.; Raimbault, M.

1991-01-01

In order to appraise the safety of a storage of high level nuclear wastes in rock masses, it is necessary to assess, among other features, the thermo-mechanical behaviour of the host rock for long periods (thousands of years). In France, four different media are considered as potential host rocks: granite, shale, salt, clay. The present paper is devoted to some analysis of a generic storage configuration in granite. The case of a rock mass without any major fault has been considered. The granite is modelled by means of an elastic fracturing model (no tension type). The results obtained show that some fissures, induced by the heat generation, develop mainly above the repository. The opening of the fissures, within the frame of the adopted hypothesis, have not a strong influence on the rock mass, as a geological barrier for the radionuclides. (author)

Proposed classification scheme for high-level and other radioactive wastes

International Nuclear Information System (INIS)

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

1986-01-01

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

Initial performance assessment of the disposal of spent nuclear fuel and high-level waste stored at Idaho National Engineering Laboratory. Volume 1, Methodology and results

Energy Technology Data Exchange (ETDEWEB)

Rechard, R.P. [ed.

1993-12-01

This performance assessment characterized plausible treatment options conceived by the Idaho National Engineering Laboratory (INEL) for its spent fuel and high-level radioactive waste and then modeled the performance of the resulting waste forms in two hypothetical, deep, geologic repositories: one in bedded salt and the other in granite. The results of the performance assessment are intended to help guide INEL in its study of how to prepare wastes and spent fuel for eventual permanent disposal. This assessment was part of the Waste Management Technology Development Program designed to help the US Department of Energy develop and demonstrate the capability to dispose of its nuclear waste. Although numerous caveats must be placed on the results, the general findings were as follows: Though the waste form behavior depended upon the repository type, all current and proposed waste forms provided acceptable behavior in the salt and granite repositories.

Release rates from waste packages in a salt repository

International Nuclear Information System (INIS)

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

1987-06-01

In this report we present estimates of radionuclide release rates from waste packages into salt. This conservative and bounding analysis shows that release rates from waste packages in salt are well below the US Nuclear Regulatory Commission's performance objectives for the engineered barrier system. 2 refs., 2 figs

Spent fuel and high-level radioactive waste storage

International Nuclear Information System (INIS)

Trigerman, S.

1988-06-01

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

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

International Nuclear Information System (INIS)

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

2013-01-01

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

Areal thermal loading recommendations for nuclear waste repositories in salt

International Nuclear Information System (INIS)

Russell, J.E.

1979-06-01

This document gives a wider understanding of the history of the recommended thermal loadings in salt for both high-level waste (HLW) from fresh UO 2 -fueled, light-water reactors (LWR) with no recycle and spent unreprocessed fuel (SURF) from LWRs. Aspects of the current recommendations that need further study are identified. Finally, an interim set of design thermal-loading recommendations are given that have a common rationale of satisfying performance limits within our current state of knowledge. These recommendations are made on a generic rather than a site-specific basis. 11 figures, 5 tables

Development of high-level waste solidification technology 1

Energy Technology Data Exchange (ETDEWEB)

Kim, Joon Hyung; Kim, Hwan Young; Kim, In Tae [and others

1999-02-01

Spent nuclear fuel contains useful nuclides as valuable resource materials for energy, heat and catalyst. High-level wastes (HLW) are expected to be generated from the R and D activities and reuse processes. It is necessary to develop vitrification or advanced solidification technologies for the safe long-term management of high level wastes. As a first step to establish HLW vitrification technology, characterization of HLWs that would arise at KAERI site, glass melting experiments with a lab-scale high frequency induction melter, and fabrication and property evaluation of base-glass made of used HEPA filter media and additives were performed. Basic study on the fabrication and characterization of candidate ceramic waste form (Synroc) was also carried out. These HLW solidification technologies would be directly useful for carrying out the R and Ds on the nuclear fuel cycle and waste management. (author). 70 refs., 29 tabs., 35 figs.

Alternative Electrochemical Salt Waste Forms, Summary of FY11-FY12 Results

Energy Technology Data Exchange (ETDEWEB)

Riley, Brian J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Mccloy, John S. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Crum, Jarrod V. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Lepry, William C. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Rodriguez, Carmen P. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Windisch, Charles F. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Matyas, Josef [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Westman, Matthew P. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Rieck, Bennett T. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Lang, Jesse B. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Olszta, Matthew J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Pierce, David A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2014-01-17

The Fuel Cycle Research and Development Program, sponsored by the U.S. Department of Energy Office of Nuclear Energy, is currently investigating alternative waste forms for wastes generated from nuclear fuel processing. One such waste results from an electrochemical separations process, called the “Echem” process. The Echem process utilizes a molten KCl-LiCl salt to dissolve the fuel. This process results in a spent salt containing alkali, alkaline earth, lanthanide halides and small quantities of actinide halides, where the primary halide is chloride with a minor iodide fraction. Pacific Northwest National Laboratory (PNNL) is concurrently investigating two candidate waste forms for the Echem spent-salt: high-halide minerals (i.e., sodalite and cancrinite) and tellurite (TeO2)-based glasses. Both of these candidates showed promise in fiscal year (FY) 2009 and FY2010 with a simplified nonradioactive simulant of the Echem waste. Further testing was performed on these waste forms in FY2011 and FY2012 to assess the possibility of their use in a sustainable fuel cycle. This report summarizes the combined results from FY2011 and FY2012 efforts.

Proposed classification scheme for high-level and other radioactive wastes

International Nuclear Information System (INIS)

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

1986-01-01

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

Future directions of defense programs high-level waste technology programs

International Nuclear Information System (INIS)

Chee, T.C.; Shupe, M.W.; Turner, D.A.; Campbell, M.H.

1987-01-01

The Department of Energy has been managing high-level waste from the production of nuclear materials for defense activities over the last forty years. An objective for the Defense Waste and Transportation Management program is to develop technology which ensures the safe, permanent disposal of all defense radioactive wastes. Technology programs are underway to address the long-term strategy for permanent disposal of high-level waste generated at each Department of Energy site. Technology is being developed for assessing the hazards, environmental impacts, and costs of each long-term disposal alternative for selection and implementation. This paper addresses key technology development areas, and consideration of recent regulatory requirements associated with the long-term management of defense radioactive high-level waste

Deployment of quasi-digital sensor for high temperature molten salt level measurement in pyroprocessing plants

Science.gov (United States)

Sanga, Ramesh; Agarwal, Sourabh; Sivaramakrishna, M.; Rao, G. Prabhakara

2018-04-01

Development of a liquid molten salt level sensor device that can detect the level of liquid molten salt in the process vessels of pyrochemical reprocessing of spent metallic fuels is detailed. It is proposed to apply a resistive-type pulsating sensor-based level measurement approach. There are no commercially available sensors due to limitations of high temperature, radiation, and physical dimensions. A compact, simple, rugged, low power, and high precise pulsating sensor-based level probe and simple instrumentation for the molten salt liquid level sensor to work in the extreme conditions has been indigenously developed, with high precision and accuracy. The working principle, design concept, and results have been discussed. This level probe is mainly composed of the variable resistor made up of ceramic rods. This resistor constitutes the part of resistance-capacitance-type Logic Gate Oscillator (LGO). A change in the molten salt level inside the tank causes a small change in the resistance which in turn changes the pulse frequency of the LGO. Thus the frequency, the output of the instrument that is displayed on the LCD of an embedded system, is a function of molten salt level. In the present design, the range of level measurement is about 10 mm. The sensitivity in position measurement up to 10 mm is ˜2.5 kHz/mm.

Study on application of molten salt oxidation technology (MSO) for PVC wastes treatment

International Nuclear Information System (INIS)

Tran Thu Ha; Nguyen Hong Quy; Pham Quoc Ky; Nguyen Quang Long; Vuong Thu Bac; Dang Duc Nhan

2007-01-01

The project 'Study on application of molten salt oxidation (MSO) for PVC plastic wastes treatment' aims at three followings: 1) Installation of lab-scale MSO unit with essential compositions builds up foundation for the 2) estimation of waste destruction efficiency of the technology. 3) Based on the results of testing PVC - the chlorinated organic wastes on the lab-scale unit, the ability of the technology application at pilot-scale level will be primary estimated. The adjustment and correction of some compositions in the lab-scale unit theoretically designed during experiment overcame the shortages by design and fabrication such as heat distribution regime, feeding wastes and draining spent salt. These solutions adapt to the technical requirement of operation as well as scientific requirement of the research on MSO process. PVC waste treatment was tested on the MSO lab-scale unit in different conditions of operation temperature, superficial air velocity related to air/oxygen feeding rate, waste feeding rate. The testing results showed that destruction efficiency of chlorine in MSO technology was almost absolute. HCl and Cl 2 emission were insignificant in different operation conditions. HCl and Cl 2 emission depend on resident time and nature of molten salt. However, with inherent attributes of MSO technology emission of CO is not avoided in processing waste treatment. Therefore, finding active solutions for reduction CO emission is essential to complete the technology. The experiments also were carried in conditions of single molten salt (Na 2 CO 3 ) and molten (Na 2 CO 3 - K 2 CO 3 ) eutectic. The comparison of efficiency of these tests gives idea of using molten salt eutectic to reduce operation cost in MSO technology. Based on operation parameters and scientific verification results during experiments, the introductory procedure of waste treatment by MSO process was built up. Thereby, primary estimation of development of the technology in pilot-scale is given

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

Leaching of vitrified high-level-active-waste in a near reality simulated repository system

International Nuclear Information System (INIS)

Froeschen, W.; Wolf, G.K.

1987-01-01

In the FRG it is planned to vitrify the high level waste from spent fuel reprocessing and to dispose of in a salt-mine. If water penetrates into the repository a highly corrosive brine (Q-brine) will be formed and radioactive material may be leached from the glasses and transported to human environment. The corrosion system of brine, corroded steel containers of the vitrified waste, and waste-glasses was investigated under near reality conditions. Experiments in hydrothermal environment were carried out including gamma radiation of the waste-glasses and ceramic In Can Lining between glasses and metallic containments. Screening experiments by application of external cobalt-gamma-radiation showed no principal changes in leaching behaviour of simulate glasses compared to leaching without radiation. Radiation effects result in pH changes mainly which are diminished by buffer capacity of Q-brine. Lining of steel containments with ceramic fleece does not reduce leaching but retards solution of Mo and Sr into brine. Decreasing of elements Sr, Cs and Mo in the near surface area of the glass and increasing of Zr and Ti has been found to be enhanced considerably in presence of canister corrosion products in Q-brine as well as in NaCl-leaching solution. (orig.) With 13 refs., 22 figs [de

High level waste canister emplacement and retrieval concepts study

International Nuclear Information System (INIS)

1975-09-01

Several concepts are described for the interim (20 to 30 years) storage of canisters containing high level waste, cladding waste, and intermediate level-TRU wastes. It includes requirements, ground rules and assumptions for the entire storage pilot plant. Concepts are generally evaluated and the most promising are selected for additional work. Follow-on recommendations are made

Assessment of lead tellurite glass for immobilizing electrochemical salt wastes from used nuclear fuel reprocessing

Energy Technology Data Exchange (ETDEWEB)

Riley, Brian J.; Kroll, Jared O.; Peterson, Jacob A.; Pierce, David A.; Ebert, William L.; Williams, Benjamin D.; Snyder, Michelle M. V.; Frank, Steven M.; George, Jaime L.; Kruska, Karen

2017-11-01

This paper provides an overview of research evaluating the use of lead tellurite glass as a waste form for salt wastes from electrochemical reprocessing of used nuclear fuel. The efficacy of using lead tellurite glass to immobilize three different salt compositions was evaluated: a LiCl-Li2O oxide reduction salt containing fission products from oxide fuel, a LiCl-KCl eutectic salt containing fission products from metallic fuel, and SrCl2. Physical and chemical properties of glasses made with these salts were characterized with X-ray diffraction, bulk density measurements, differential thermal analysis, chemical durability tests, scanning and transmission electron microscopies, and energy-dispersive X-ray spectroscopy. These glasses were found to accommodate high salt concentrations and have high densities, but further development is needed to improve chemical durability. (C) 2017 Published by Elsevier B.V.

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

High level waste fixation in cermet form

International Nuclear Information System (INIS)

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

1981-01-01

Commercial and defense high level waste fixation in cermet form is being studied by personnel of the Isotopes Research Materials Laboratory, Solid State Division (ORNL). As a corollary to earlier research and development in forming high density ceramic and cermet rods, disks, and other shapes using separated isotopes, similar chemical and physical processing methods have been applied to synthetic and real waste fixation. Generally, experimental products resulting from this approach have shown physical and chemical characteristics which are deemed suitable for long-term storage, shipping, corrosive environments, high temperature environments, high waste loading, decay heat dissipation, and radiation damage. Although leach tests are not conclusive, what little comparative data are available show cermet to withstand hydrothermal conditions in water and brine solutions. The Soxhlet leach test, using radioactive cesium as a tracer, showed that leaching of cermet was about X100 less than that of 78 to 68 glass. Using essentially uncooled, untreated waste, cermet fixation was found to accommodate up to 75% waste loading and yet, because of its high thermal conductivity, a monolith of 0.6 m diameter and 3.3 m-length would have only a maximum centerline temperature of 29 K above the ambient value

Immobilization of LiCl-Li 2 O pyroprocessing salt wastes in chlorosodalite using glass-bonded hydrothermal and salt-occlusion methods

Energy Technology Data Exchange (ETDEWEB)

Riley, Brian J.; Peterson, Jacob A.; Kroll, Jared O.; Frank, Steven M.

2018-04-01

In this study, salt occlusion and hydrothermal processes were used to make chlorosodalite through reaction with a high-LiCl salt simulating a waste stream following pyrochemical treatment of oxide-based used nuclear fuel. Some products were reacted with glass binders to increase chlorosodalite yield through alkali ion exchange and aide in densification. Hydrothermal processes included reaction of the salt simulant in an acid digestion vessel with either zeolite 4A or sodium aluminate and colloidal silica. Chlorosodalite yields in the crystalline products were nearly complete in the glass-bonded materials at values of 100 mass% for the salt-occlusion method, up to 99.0 mass% for the hydrothermal synthesis with zeolite 4A, and up to 96 mass% for the hydrothermal synthesis with sodium aluminate and colloidal silica. These results show promise for using chemically stable chlorosodalite to immobilize oxide reduction salt wastes.

Immobilization of LiCl-Li2O pyroprocessing salt wastes in chlorosodalite using glass-bonded hydrothermal and salt-occlusion methods

Science.gov (United States)

Riley, Brian J.; Peterson, Jacob A.; Kroll, Jared O.; Frank, Steven M.

2018-04-01

In this study, hydrothermal and salt-occlusion processes were used to make chlorosodalite through reactions with a high-LiCl salt simulating a waste stream generated from pyrochemical treatment of oxide-based used nuclear fuel. Some products were reacted with glass binders to increase chlorosodalite yield through alkali ion exchange and to aid in densification. Hydrothermal processes included reaction of the salt simulant in an autoclave with either zeolite 4A or sodium aluminate and colloidal silica. Chlorosodalite yields in the crystalline products were nearly complete in the glass-bonded materials at values of 100 mass% for the salt-occlusion method, up to 99.0 mass% for the hydrothermal synthesis with zeolite 4A, and up to 96 mass% for the hydrothermal synthesis with sodium aluminate and colloidal silica. These results show promise for using chemically stable chlorosodalite to immobilize oxide reduction salt wastes.

Alternative Electrochemical Salt Waste Forms, Summary of FY/CY2011 Results

International Nuclear Information System (INIS)

Riley, Brian J.; McCloy, John S.; Crum, Jarrod V.; Rodriguez, Carmen P.; Windisch, Charles F.; Lepry, William C.; Matyas, Josef; Westman, Matthew P.; Rieck, Bennett T.; Lang, Jesse B.; Pierce, David A.

2011-01-01

This report summarizes the 2011 fiscal+calendar year efforts for developing waste forms for a spent salt generated in reprocessing nuclear fuel with an electrochemical separations process. The two waste forms are tellurite (TeO2-based) glasses and sol-gel-derived high-halide mineral analogs to stable minerals found in nature.

Alternative Electrochemical Salt Waste Forms, Summary of FY/CY2011 Results

Energy Technology Data Exchange (ETDEWEB)

Riley, Brian J.; McCloy, John S.; Crum, Jarrod V.; Rodriguez, Carmen P.; Windisch, Charles F.; Lepry, William C.; Matyas, Josef; Westman, Matthew P.; Rieck, Bennett T.; Lang, Jesse B.; Pierce, David A.

2011-12-01

This report summarizes the 2011 fiscal+calendar year efforts for developing waste forms for a spent salt generated in reprocessing nuclear fuel with an electrochemical separations process. The two waste forms are tellurite (TeO2-based) glasses and sol-gel-derived high-halide mineral analogs to stable minerals found in nature.

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

International Nuclear Information System (INIS)

Cowan, S.P.

1985-01-01

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

Near-surface storage facilities for vitrified high-level wastes

International Nuclear Information System (INIS)

Kondrat'ev, A.N.; Kulichenko, V.V.; Kryukov, I.I.; Krylova, N.V.; Paramoshkin, V.I.; Strakhov, M.V.

1980-01-01

Concurrently with the development of methods for solidifying liquid radioactive wastes, reliable and safe methods for the storage and disposal of solidified wastes are being devised in the USSR and other countries. One of the main factors affecting the choice of storage conditions for solidified wastes originating from the vitrification of high-level liquid wastes from fuel reprocessing plants is the problem of removing the heat produced by radioactive decay. In order to prevent the temperature of solidified wastes from exceeding the maximum permissible level for the material concerned, it is necessary to limit either the capacity of waste containers or the specific heat release of the wastes themselves. In order that disposal of high-level wastes in geological formations should be reliable and economic, solidified wastes undergo interim storage in near-surface storage facilities with engineered cooling systems. The paper demonstrates the relative influences of specific heat release, of the maximum permissible storage temperature for vitrified wastes and of the methods chosen for cooling wastes in order for the dimensions of waste containers to be reduced to the extent required. The effect of concentrating wastes to a given level in the vitrification process on the cost of storage in different types of storage facility is also examined. Calculations were performed for the amount of vitrified wastes produced by a reprocessing plant with a capacity of five tonnes of uranium per 24 hours. Fuel elements from reactors of the water-cooled, water-moderated type are sent for reprocessing after having been held for about two years. The dimensions of the storage facility are calculated on the assumption that it will take five years to fill

DESIGN ANALYSIS FOR THE DEFENSE HIGH-LEVEL WASTE DISPOSAL CONTAINER

International Nuclear Information System (INIS)

Radulesscu, G.; Tang, J.S.

2000-01-01

The purpose of ''Design Analysis for the Defense High-Level Waste Disposal Container'' analysis is to technically define the defense high-level waste (DHLW) disposal container/waste package using the Waste Package Department's (WPD) design methods, as documented in ''Waste Package Design Methodology Report'' (CRWMS M andO [Civilian Radioactive Waste Management System Management and Operating Contractor] 2000a). The DHLW disposal container is intended for disposal of commercial high-level waste (HLW) and DHLW (including immobilized plutonium waste forms), placed within disposable canisters. The U.S. Department of Energy (DOE)-managed spent nuclear fuel (SNF) in disposable canisters may also be placed in a DHLW disposal container along with HLW forms. The objective of this analysis is to demonstrate that the DHLW disposal container/waste package satisfies the project requirements, as embodied in Defense High Level Waste Disposal Container System Description Document (SDD) (CRWMS M andO 1999a), and additional criteria, as identified in Waste Package Design Sensitivity Report (CRWMS M andQ 2000b, Table 4). The analysis briefly describes the analytical methods appropriate for the design of the DHLW disposal contained waste package, and summarizes the results of the calculations that illustrate the analytical methods. However, the analysis is limited to the calculations selected for the DHLW disposal container in support of the Site Recommendation (SR) (CRWMS M andO 2000b, Section 7). The scope of this analysis is restricted to the design of the codisposal waste package of the Savannah River Site (SRS) DHLW glass canisters and the Training, Research, Isotopes General Atomics (TRIGA) SNF loaded in a short 18-in.-outer diameter (OD) DOE standardized SNF canister. This waste package is representative of the waste packages that consist of the DHLW disposal container, the DHLW/HLW glass canisters, and the DOE-managed SNF in disposable canisters. The intended use of this

DESIGN ANALYSIS FOR THE DEFENSE HIGH-LEVEL WASTE DISPOSAL CONTAINER

Energy Technology Data Exchange (ETDEWEB)

G. Radulesscu; J.S. Tang

2000-06-07

The purpose of ''Design Analysis for the Defense High-Level Waste Disposal Container'' analysis is to technically define the defense high-level waste (DHLW) disposal container/waste package using the Waste Package Department's (WPD) design methods, as documented in ''Waste Package Design Methodology Report'' (CRWMS M&O [Civilian Radioactive Waste Management System Management and Operating Contractor] 2000a). The DHLW disposal container is intended for disposal of commercial high-level waste (HLW) and DHLW (including immobilized plutonium waste forms), placed within disposable canisters. The U.S. Department of Energy (DOE)-managed spent nuclear fuel (SNF) in disposable canisters may also be placed in a DHLW disposal container along with HLW forms. The objective of this analysis is to demonstrate that the DHLW disposal container/waste package satisfies the project requirements, as embodied in Defense High Level Waste Disposal Container System Description Document (SDD) (CRWMS M&O 1999a), and additional criteria, as identified in Waste Package Design Sensitivity Report (CRWMS M&Q 2000b, Table 4). The analysis briefly describes the analytical methods appropriate for the design of the DHLW disposal contained waste package, and summarizes the results of the calculations that illustrate the analytical methods. However, the analysis is limited to the calculations selected for the DHLW disposal container in support of the Site Recommendation (SR) (CRWMS M&O 2000b, Section 7). The scope of this analysis is restricted to the design of the codisposal waste package of the Savannah River Site (SRS) DHLW glass canisters and the Training, Research, Isotopes General Atomics (TRIGA) SNF loaded in a short 18-in.-outer diameter (OD) DOE standardized SNF canister. This waste package is representative of the waste packages that consist of the DHLW disposal container, the DHLW/HLW glass canisters, and the DOE-managed SNF in disposable

Evaluation and selection of candidate high-level waste forms

International Nuclear Information System (INIS)

1982-03-01

Seven candidate waste forms being developed under the direction of the Department of Energy's National High-Level Waste (HLW) Technology Program, were evaluated as potential media for the immobilization and geologic disposal of high-level nuclear wastes. The evaluation combined preliminary waste form evaluations conducted at DOE defense waste-sites and independent laboratories, peer review assessments, a product performance evaluation, and a processability analysis. Based on the combined results of these four inputs, two of the seven forms, borosilicate glass and a titanate based ceramic, SYNROC, were selected as the reference and alternative forms for continued development and evaluation in the National HLW Program. Both the glass and ceramic forms are viable candidates for use at each of the DOE defense waste-sites; they are also potential candidates for immobilization of commercial reprocessing wastes. This report describes the waste form screening process, and discusses each of the four major inputs considered in the selection of the two forms

Mixing processes in high-level waste tanks. 1998 annual progress report

International Nuclear Information System (INIS)

Peterson, P.F.

1998-01-01

Flammable gases can be generated in DOE high-level waste tanks, including radiolytic hydrogen, and during cesium precipitation from salt solutions, benzene. Under normal operating conditions the potential for deflagration or detonation from these gases is precluded by purging and ventilation systems, which remove the flammable gases and maintain a well-mixed condition in the tanks. Upon failure of the ventilation system, due to seismic or other events, however, it has proven more difficult to make strong arguments for well-mixed conditions, due to the potential for density-induced stratification which can potentially sequester fuel or oxidizer at concentrations significantly higher than average. This has complicated the task of defining the safety basis for tank operation. Waste-tank mixing processes have considerable overlap with similar large-enclosure mixing processes that occur in enclosure fires and nuclear reactor containments. Significant differences also exist, so that modeling techniques that have been developed previously can not be directly applied to waste tanks. In particular, mixing of air introduced through tank roof penetrations by buoyancy and pressure driven exchange flows, mixed convection induced by an injected high-velocity purge jet interacting with buoyancy driven flow, and onset and breakdown of stable stratification under the influence of an injected jet have not been adequately studied but are important in assessing the potential for accumulation of high-concentration pockets of fuel and oxygen. Treating these phenomena requires a combination of experiments and the development of new, more general computational models than those that have been developed for enclosure fires. U.C. Berkeley is now completing the second year of its three-year project that started in September, 1996. Excellent progress has been made in several important areas related to waste-tank ventilation and mixing processes.'

International high-level radioactive waste repositories

International Nuclear Information System (INIS)

Lin, W.

1996-01-01

Although nuclear technologies benefit everyone, the associated nuclear wastes are a widespread and rapidly growing problem. Nuclear power plants are in operation in 25 countries, and are under construction in others. Developing countries are hungry for electricity to promote economic growth; industrialized countries are eager to export nuclear technologies and equipment. These two ingredients, combined with the rapid shrinkage of worldwide fossil fuel reserves, will increase the utilization of nuclear power. All countries utilizing nuclear power produce at least a few tens of tons of spent fuel per year. That spent fuel (and reprocessing products, if any) constitutes high-level nuclear waste. Toxicity, long half-life, and immunity to chemical degradation make such waste an almost permanent threat to human beings. This report discusses the advantages of utilizing repositories for disposal of nuclear wastes

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

International Nuclear Information System (INIS)

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

1987-06-01

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

The HAW project. Test storage of high-level radiation sources in the Asse salt mine. Documentation and assessment of the storage system

International Nuclear Information System (INIS)

Mueller, K.; Rothfuchs, T.

1994-01-01

The HAW project aimed primarily at studying the interaction between high-level radioactive waste moulds and rock salt as the respository medium. Another priority was the prototype development and testing of a technical system for the emplacement of high-level radioactive moulds in deep storage boreholes. To simulate real high-level radioactive wastes, special high-level radiation sources (Cs-137, Sr-90) were produced in the United States of America under a German-American cooperation contract, for carrying out the tests at the Asse salt mine. The components of the storage system are described, their position and task within the entire handling procedure explained. Questions of radiation protection and accident protection, of functioning and operating reliability, of quality assurance and examination of documents, materials, of manufacture and functioning, and of documentation are dealt with in detail. With a view to the planning of storage techniques for a mine respository, the experience of development and operation is recorded, and recommendation of further developments are given. Problems which arose during work on the HAW project were partly due to test-specific reasons and will not or not in this form occur in a mine respository. It was planned to start the test emplacement in 1987, and it could have been executed in 1993 after appropriate preparation and approval of the storage system by the mining authority and the Hanover TUEV in 1991. In December 1992, however, the Federal Government decided to give up to the project due to the uncertain licensing situation, and to immediately stop all preparatory work. (orig./HP) [de

High-level radioactive waste in Canada. Background paper

International Nuclear Information System (INIS)

Fawcett, R.

1993-11-01

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

High-level radioactive waste in Canada. Background paper

Energy Technology Data Exchange (ETDEWEB)

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

1993-11-01

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

Alternative methods of salt disposal at the seven salt sites for a nuclear waste repository

International Nuclear Information System (INIS)

1987-02-01

This study discusses the various alternative salt management techniques for the disposal of excess mined salt at seven potentially acceptable nuclear waste repository sites: Deaf Smith and Swisher Counties, Texas; Richton and Cypress Creek Domes, Mississippi; Vacherie Dome, Louisiana; and Davis and Lavender Canyons, Utah. Because the repository development involves the underground excavation of corridors and waste emplacement rooms, in either bedded or domed salt formations, excess salt will be mined and must be disposed of offsite. The salt disposal alternatives examined for all the sites include commercial use, ocean disposal, deep well injection, landfill disposal, and underground mine disposal. These alternatives (and other site-specific disposal methods) are reviewed, using estimated amounts of excavated, backfilled, and excess salt. Methods of transporting the excess salt are discussed, along with possible impacts of each disposal method and potential regulatory requirements. A preferred method of disposal is recommended for each potentially acceptable repository site. 14 refs., 5 tabs

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

International Nuclear Information System (INIS)

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

1989-01-01

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

Waste management analysis for the nuclear fuel cycle. I. Actinide recovery from aqueous salt wastes

International Nuclear Information System (INIS)

Martella, L.L.; Navratil, J.D.

1979-01-01

A preliminary feasibility study of solvent extraction methods has been completed for removing actinides from selected salt wastes likely to be produced during reactor fuel fabrication and reprocessing. The use of a two-step solvent extraction system, tributyl phosphate (TBP) followed by a bidentate organophosphorus extractant (DHDECMP), appears most efficient for removing actinides from salt waste. The TBP step would remove most of the plutonium and >99.99% of the uranium. The second step, using DHDECMP, would remove >99.91% of the americium, the remaining plutonium (>99.98%), and other actinides from the acidified salt waste

High-temperature vitrification of low-level radioactive and hazardous wastes

International Nuclear Information System (INIS)

Schumacher, R.F.; Kielpinski, A.L.; Bickford, D.F.; Cicero, C.A.; Applewhite-Ramsey, A.; Spatz, T.L.; Marra, J.C.

1995-01-01

The US Department of Energy (DOE) weapons complex has numerous radioactive waste streams which cannot be easily treated with joule-heated vitrification systems. However, it appears these streams could be treated With certain robust, high-temperature, melter technologies. These technologies are based on the use of plasma torch, graphite arc, and induction heating sources. The Savannah River Technology Center (SRTC), with financial support from the Department of Energy, Office of Technology Development (OTD) and in conjunction with the sites within the DOE weapons complex, has been investigating high-temperature vitrification technologies for several years. This program has been a cooperative effort between a number of nearby Universities, specific sites within the DOE complex, commercial equipment suppliers and the All-Russian Research Institute of Chemical Technology. These robust vitrification systems appear to have advantages for the waste streams containing inorganic materials in combination with significant quantities of metals, organics, salts, or high temperature materials. Several high-temperature technologies were selected and will be evaluated and employed to develop supporting technology. A general overview of the SRTC ''High-Temperature Program'' will be provided

High-temperature vitrification of low-level radioactive and hazardous wastes

International Nuclear Information System (INIS)

Schumacher, R.F.; Kielpinski, A.L.; Bickford, D.F.; Cicero, C.A.; Applewhite-Ramsey, A.; Spatz, T.L.; Marra, J.C.

1995-01-01

The US Department of Energy (DOE) weapons complex has numerous radioactive waste streams which cannot be easily treated with joule-heated vitrification systems. However, it appears that these streams could be treated with certain robust, high-temperature, melter technologies. These technologies are based on the use of plasma torch, graphite arc, and induction heating sources. The Savannah River Technology Center (SRTC), with financial support from the Department of Energy, Office of Technology Development (OTD) and in conjunction with the sites within the DOE weapons complex, has been investigating high-temperature vitrification technologies for several years. This program has been a cooperative effort between a number of nearby Universities, specific sites within the DOE complex, commercial equipment suppliers and the All-Russian Research Institute of Chemical Technology. These robust vitrification systems appear to have advantages for the waste streams containing inorganic materials in combination with significant quantities of metals, organics, salts, or high temperature materials. Several high-temperature technologies were selected and will be evaluated and employed to develop supporting technology. A general overview of the SRTC ''High-Temperature Program'' will be provided

Methods of Disposing Of High-Level Radioactive Waste: A Review

International Nuclear Information System (INIS)

Abumurade, K.

2002-01-01

High level nuclear waste from both commercial reactors and defense industry presents a difficult problem to the scientific community as well as the public. The solutions to this problem is still debatable both technically and ethically. There are few methods proposed for disposing of high level waste. Each method has its own advantages and disadvantages. However, the very deep underground geologic repository is the best choice for disposing of high-level radioactive wastes. The cost benefit equation of nuclear power production and its waste is discussed. However, the public should be educated about this matter to minimize the gap between them and the nuclear power community including scientists industry, and governments. (Author) 15 refs., 4 tabs., 1 fig

High level radioactive wastes: Considerations on final disposal

International Nuclear Information System (INIS)

Ciallella, Norberto R.

2000-01-01

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

Design concepts of definitive disposal for high level radioactive wastes

International Nuclear Information System (INIS)

Badillo A, V.E.; Alonso V, G.

2007-01-01

It is excessively known the importance about finding a solution for the handling and disposition of radioactive waste of all level. However, the polemic is centered in the administration of high level radioactive waste and the worn out fuel, forgetting that the more important volumes of waste its are generated in the categories of low level wastes or of very low level. Depending on the waste that will be confined and of the costs, several technological modalities of definitive disposition exist, in function of the depth of the confinement. The concept of deep geologic storage, technological option proposed more than 40 years ago, it is a concept of isolation of waste of long half life placed in a deep underground installation dug in geologic formations that are characterized by their high stability and their low flow of underground water. In the last decades, they have registered countless progresses in technical and scientific aspects of the geologic storage, making it a reliable technical solution supported with many years of scientific work carried out by numerous institutions in the entire world. In this work the design concepts that apply some countries for the high level waste disposal that its liberate heat are revised and the different geologic formations that have been considered for the storage of this type of wastes. (Author)

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).

High-level radioactive waste disposal type and theoretical analyses

International Nuclear Information System (INIS)

Lu Yingfa; Wu Yanchun; Luo Xianqi; Cui Yujun

2006-01-01

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

Expedited demonstration of molten salt mixed waste treatment technology. Final report

International Nuclear Information System (INIS)

1995-01-01

This final report discusses the molten salt mixed waste project in terms of the various subtasks established. Subtask 1: Carbon monoxide emissions; Establish a salt recycle schedule and/or a strategy for off-gas control for MWMF that keeps carbon monoxide emission below 100 ppm on an hourly averaged basis. Subtask 2: Salt melt viscosity; Experiments are conducted to determine salt viscosity as a function of ash composition, ash concentration, temperature, and time. Subtask 3: Determine that the amount of sodium carbonate entrained in the off-gas is minimal, and that any deposited salt can easily be removed form the piping using a soot blower or other means. Subtask 4: The provision of at least one final waste form that meets the waste acceptance criteria of a landfill that will take the waste. This report discusses the progress made in each of these areas

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

Comparison of the intermediate storage periods and areas required for final storage of high-level radioactive waste and spent fuel in various types of host rock

International Nuclear Information System (INIS)

Mueller-Hoeppe, N.; Lerch, C.; Jobmann, M.; Filbert, W.

2005-01-01

The present new version of the German concept for radioactive waste and spent fuel management is based on the assumption that a repository for high-level waste and spent fuel will not be required until 2030. One reason frequently given for this date is the intermediate storage period of at least forty years to allow the very high initial heat generation to decay. However, calculations performed by the authors have shown that the minimum intermediate storage period for a repository in rock salt is only between four and nineteen years, depending on the final storage concept and the load of the waste package. In clay as a host rock, the minimum intermediate storage times were calculated to be between 31 and 142 years; the same time spans are expected to apply to final storage in magmatic rock, such as granite. The maximum permissible loads of a container holding spent fuel in salt are many times those in clay and granite, respectively. It was also seen that the area requirement for final storage of the same waste structures is roughly a factor of ten higher in clay than in salt. The differences between granite and salt are similar. The reasons for these grave differences, on the one hand, are the better thermal conductivity of salt and, on the other hand, the better heat tolerance of the crushed salt used as backfill material compared to that of bentonite used in the clay and granite concepts. While salt will allow temperatures of up to 200 C, the maximum temperature in bentonite is limited to 100 C. (orig.)

Nuclear waste. DOE's program to prepare high-level radioactive waste for final disposal

International Nuclear Information System (INIS)

Bannerman, Carl J.; Owens, Ronald M.; Dowd, Leonard L.; Herndobler, Christopher S.; Purvine, Nancy R.; Stenersen, Stanley G.

1989-11-01

In summary, as of December 1988, the four sites collectively stored about 95 million gallons of high-level waste in underground tanks and bins. Approximately 57 million gallons are stored at Hanford, 34 million gallons at Savannah River, 3 million gallons at INEL, and 6 million gallons at West Valley. The waste is in several forms, including liquid, sludge, and dry granular materials, that make it unsuitable for permanent storage in its current state at these locations. Leaks from the tanks, designed for temporary storage, can pose an environmental hazard to surrounding land and water for thousands of years. DOE expects that when its waste processes at Savannah River, West Valley, and Hanford become operational, the high-level radioactive waste stored at these sites will be blended with other materials to immobilize it by forming a glass-like substance. The glass form will minimize the risk of environmental damage and make the waste more acceptable for permanent disposal in a geologic repository. At INEL, DOE is still considering various other immobilization and permanent disposal approaches. In July 1989, DOE estimated that it would cost about $13 billion (in fiscal year 1988 dollars) to retrieve, process, immobilize, and store the high-level waste until it can be moved to a permanent disposal site: about $5.3 billion is expected to be spent at Savannah River, $0.9 billion at West Valley, $2.8 billion at Hanford, and $4.0 billion at INEL. DOE has started construction at Savannah River and West Valley for facilities that will be used to transform the waste into glass (a process known as vitrification). These sites have each encountered schedule delays, and one has encountered a significant cost increase over earlier estimates. More specifically, the Savannah River facility is scheduled to begin high-level waste vitrification in 1992; the West Valley project, based on a January 1989 estimate, is scheduled to begin high-level waste vitrification in 1996, about 8

Chemical modeling of nuclear waste repositories in the salt repository project

International Nuclear Information System (INIS)

Jansen, G.; Raines, G.E.; Kircher, J.F.; Hubbard, N.

1985-01-01

Salt deposits contain small amounts of water as brine in fluid inclusions in halite and in hydrous minerals, e.g., clays, kieserite (MgSO 4 . H 2 O) and carnallite (KMgCl 3 . 6H 2 O). For the candidate salt deposits, the total amounts of water as volume % brine are: Palo Duro Basin, Texas, approximately 1.8; Paradox Basin, Utah, approximately 5.0 for the carnallite-marker zone, and less than approximately 0.5 below this zone; Gulf Coast salt domes, less than 0.15. For the Palo Duro and Paradox salt, the brines are Mg-rich (approximately 20,000 mg/L to approximately 100,000 mg/L) and sometimes Ca-rich (up to about 20,000 mg/L) NaCl brines. Brine migration calculations have been made using calculations of the time-variant thermal gradient around the waste packages and conservatively high brine volumes in the salt (5.0 volume % for the Texas and Utah sites and 0.5 volume % for the Gulf Coast) as input data. The maximum amounts of brine that eventually migrate to each waste package are about 1.0m 3 (for 5.0 volume % brine) and 0.2m 3 (for 0.5 volume % brine). With current conceptual designs for waste package overpacks (10 to 15 cm thick low-carbon steel), the waste package is not breached by uniform corrosion within 10,000 years. In brines this material thus far shows only uniform corrosion. For the expected conditions, where the brine is provided solely by brine migration, the brine is consumed by reaction with the iron of the overpack nearly as fast as it migrates to the waste package. Therefore, for the expected conditions, data about corrosion rates, radiolysis, etc., are not important. However, it is essential that accurate volumes of in-migrating brine can be calculated

Department of Energy pretreatment of high-level and low-level wastes

International Nuclear Information System (INIS)

McGinnis, C.P.; Hunt, R.D.

1995-01-01

The remediation of the 1 x 10 8 gal of highly radioactive waste in the underground storage tanks (USTs) at five US Department of Energy (DOE) sites is one of DOE's greatest challenges. Therefore, the DOE Office of Environmental Management has created the Tank Focus Area (TFA) to manage an integrated technology development program that results in the safe and efficient remediation of UST waste. The TFA has divided its efforts into five areas, which are safety, characterization, retrieval/closure, pretreatment, and immobilization. All DOE pretreatment activities are integrated by the Pretreatment Technical Integration Manager of the TFA. For FY 1996, the 14 pretreatment tasks are divided into 3 systems: supernate separations, sludge treatment, and solid/liquid separation. The plans and recent results of these TFA tasks, which include two 25,000-gal demonstrations and two former TFA tasks on Cs removal, are presented. The pretreatment goals are to minimize the volume of high-level waste and the radioactivity in low-level waste

High-level waste description, inventory and hazard

International Nuclear Information System (INIS)

Crandall, J.; Hennelly, E.J.; McElroy, J.L.

1983-01-01

High-level nuclear waste (HLW), including its origin, is described and the current differences in definitions discussed. Quantities of defense and commercial radioactive HLW, both volume and curie content, are given. Current waste handling, which is interimin nature, is described for the several sites. The HLW hazard is defined by the times during which various radionuclides are the dominant contributors. The hazard is also compared to that of the ore. Using ICRP-2, which is the legal reference in the US, the hazard of the waste reduces to a level equal to the ore in about 300 years. The disposal plans are summarized and it is shown that regulatory requirements will probably govern disposal operations in such a conservative manner that the risk (product of hazard times probability of release) may well be lower than for any other wastes in existence or perhaps lower than those for any other human endeavor

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

Energy Technology Data Exchange (ETDEWEB)

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

1980-01-01

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

Risk analyses for disposing nonhazardous oil field wastes in salt caverns

Energy Technology Data Exchange (ETDEWEB)

Tomasko, D.; Elcock, D.; Veil, J.; Caudle, D.

1997-12-01

Salt caverns have been used for several decades to store various hydrocarbon products. In the past few years, four facilities in the US have been permitted to dispose nonhazardous oil field wastes in salt caverns. Several other disposal caverns have been permitted in Canada and Europe. This report evaluates the possibility that adverse human health effects could result from exposure to contaminants released from the caverns in domal salt formations used for nonhazardous oil field waste disposal. The evaluation assumes normal operations but considers the possibility of leaks in cavern seals and cavern walls during the post-closure phase of operation. In this assessment, several steps were followed to identify possible human health risks. At the broadest level, these steps include identifying a reasonable set of contaminants of possible concern, identifying how humans could be exposed to these contaminants, assessing the toxicities of these contaminants, estimating their intakes, and characterizing their associated human health risks. The contaminants of concern for the assessment are benzene, cadmium, arsenic, and chromium. These were selected as being components of oil field waste and having a likelihood to remain in solution for a long enough time to reach a human receptor.

Molten salt destruction process for mixed wastes

International Nuclear Information System (INIS)

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

1993-04-01

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

Preconceptual design study for solidifying high-level waste: West Valley Demonstration Project

International Nuclear Information System (INIS)

Hill, O.F.

1981-04-01

This report presents a preconceptual design study for processing radioactive high-level liquid waste presently stored in underground tanks at Western New York Nuclear Service Center (WNYNSC) near West Valley, New York, and for incorporating the radionculides in that waste into a solid. The high-level liquid waste accumulated from the operation of a chemical reprocessing plant by the Nuclear Fuel Services, Inc. from 1966 to 1972. The high-level liquid waste consists of approximately 560,000 gallons of alkaline waste from Purex process operations and 12,000 gallons of acidic (nitric acid) waste from one campaign of processing thoria fuels by a modified Thorex process (during this campaign thorium was left in the waste). The alkaline waste contains approximately 30 million curies and the acidic waste contains approximately 2.5 million curies. The reference process described in this report is concerned only with chemically processing the high-level liquid waste to remove radionuclides from the alkaline supernate and converting the radionuclide-containing nonsalt components in the waste into a borosilicate glass

Status of commercial nuclear high-level waste disposal. Special report

International Nuclear Information System (INIS)

Dau, G.J.; Williams, R.F.

1976-09-01

The results of this review, presented in the form of a functional description of high level waste management system, shows that technology is available to dispose of nuclear waste safely by several different processes. The most attractive alternative in terms of available technology and shortness of time to demonstrate it at commercial scale is a system that converts the waste to a solid by immobilizing the radioactive elements in a glass matrix. Brief comments are also given on international efforts in high level waste management and advanced disposal concepts

Risk assessment of nonhazardous oil-field waste disposal in salt caverns.

Energy Technology Data Exchange (ETDEWEB)

Elcock, D.

1998-03-05

In 1996, Argonne National Laboratory (ANL) conducted a preliminary technical and legal evaluation of disposing of nonhazardous oil-field wastes (NOW) into salt caverns. Argonne determined that if caverns are sited and designed well, operated carefully, closed properly, and monitored routinely, they could be suitable for disposing of oil-field wastes. On the basis of these findings, Argonne subsequently conducted a preliminary evaluation of the possibility that adverse human health effects (carcinogenic and noncarcinogenic) could result from exposure to contaminants released from the NOW disposed of in domal salt caverns. Steps used in this evaluation included the following: identifying potential contaminants of concern, determining how humans could be exposed to these contaminants, assessing contaminant toxicities, estimating contaminant intakes, and calculating human cancer and noncancer risk estimates. Five postclosure cavern release scenarios were assessed. These were inadvertent cavern intrusion, failure of the cavern seal, failure of the cavern through cracks, failure of the cavern through leaky interbeds, and a partial collapse of the cavern roof. Assuming a single, generic, salt cavern and generic oil-field wastes, potential human health effects associated with constituent hazardous substances (arsenic, benzene, cadmium, and chromium) were assessed under each of these scenarios. Preliminary results provided excess cancer risk and hazard index (referring to noncancer health effects) estimates that were well within the US Environmental Protection Agency (EPA) target range for acceptable exposure risk levels. These results led to the preliminary conclusion that from a human health perspective, salt caverns can provide an acceptable disposal method for nonhazardous oil-field wastes.

Risk assessment of nonhazardous oil-field waste disposal in salt caverns

International Nuclear Information System (INIS)

Elcock, D.

1998-01-01

In 1996, Argonne National Laboratory (ANL) conducted a preliminary technical and legal evaluation of disposing of nonhazardous oil-field wastes (NOW) into salt caverns. Argonne determined that if caverns are sited and designed well, operated carefully, closed properly, and monitored routinely, they could be suitable for disposing of oil-field wastes. On the basis of these findings, Argonne subsequently conducted a preliminary evaluation of the possibility that adverse human health effects (carcinogenic and noncarcinogenic) could result from exposure to contaminants released from the NOW disposed of in domal salt caverns. Steps used in this evaluation included the following: identifying potential contaminants of concern, determining how humans could be exposed to these contaminants, assessing contaminant toxicities, estimating contaminant intakes, and calculating human cancer and noncancer risk estimates. Five postclosure cavern release scenarios were assessed. These were inadvertent cavern intrusion, failure of the cavern seal, failure of the cavern through cracks, failure of the cavern through leaky interbeds, and a partial collapse of the cavern roof. Assuming a single, generic, salt cavern and generic oil-field wastes, potential human health effects associated with constituent hazardous substances (arsenic, benzene, cadmium, and chromium) were assessed under each of these scenarios. Preliminary results provided excess cancer risk and hazard index (referring to noncancer health effects) estimates that were well within the US Environmental Protection Agency (EPA) target range for acceptable exposure risk levels. These results led to the preliminary conclusion that from a human health perspective, salt caverns can provide an acceptable disposal method for nonhazardous oil-field wastes

Containment of solidified liquid hazardous waste in domal salt

International Nuclear Information System (INIS)

Domenico, P.A.; Lerman, A.

1992-01-01

In recent years, the solidification of hazardous liquid waste has become a viable option in waste management. The solidification process results in an increased volume but more stable waste form that must be disposed of or stored in a dry environment. An environment of choice in south central Texas is domal salt. The salt dome currently under investigation has a water content of 0.002 percent by weight and a permeability less than one nanodarcy. A question that must be addressed is whether a salt dome has a particular set of attributes that will prevent the release of contaminants to the environment. From a regulatory perspective, a ''no migration'' petition must be approved by the U.S.E.P.A. for the containment facility. By ''no migration'' it is implied that the waste must be contained for 10,000 years. A demonstration that this condition will be met will require model calculations and such models must be based on the physical and chemical characteristics of the waste form and the geologic environment. In particular, the models must address the rate of brine infiltration into the caverns, providing information on how fast an immobile solid waste form could convert to a more mobile liquid state. Additionally, the potential for migration by both diffusion and advection is of concern. Lastly, given a partially saturated cavern, the question of how far gaseous waste will be transported over the 10,000 year containment period must also be addressed. Results indicate that the containment capabilities of domal salt are exceptional. A nominal volume of brine will seep into the cavern and most voids between the injected solidified waste pellets will remain unsaturated. Very small quantities of hazardous constituents will be leached from the waste pellets

UK-Nuclear decommissioning authority and US Salt-stone waste management issues

International Nuclear Information System (INIS)

Lawless, William; Whitton, John

2007-01-01

Available in abstract form only. Full text of publication follows: We update two case studies of stakeholder issues in the UK and US. Earlier versions were reported at Waste Management 2006 and 2007 and at ICEM 2005. UK: The UK nuclear industry has begun to consult stakeholders more widely in recent years. Historically, methods of engagement within the industry have varied, however, recent discussions have generally been carried out with the explicit understanding that engagement with stakeholders will be 'dialogue based' and will 'inform' the final decision made by the decision maker. Engagement is currently being carried out at several levels within the industry; at the national level (via the Nuclear Decommissioning Authority's (NDA) National Stakeholder Group (NSG)); at a local site level (via Site Stakeholder Groups) and at a project level (usually via the Best Practicable Environmental Option process (BPEO)). This paper updates earlier results by the co-author with findings from a second questionnaire issued to the NSG in Phase 2 of the engagement process. An assessment is made regarding the development of stakeholder perceptions since Phase 1 towards the NDA process. US: The US case study reviews the resolution of issues on salt-stone by Department of Energy's (DOE) Savannah River Site (SRS) Citizens Advisory Board (CAB), in Aiken, SC. Recently, SRS-CAB encouraged DOE and South Carolina's regulatory Department of Health and Environmental Control (SC-DHEC) to resolve a conflict preventing SC-DHEC from releasing a draft permit to allow SRS to restart salt-stone operations. It arose with a letter sent from DOE blaming the Governor of South Carolina for delay in restarting salt processing. In reply, the Governor blamed DOE for failing to assure that Salt Waste Processing Facility (SWPF) would be built. SWPF is designed to remove most of the radioactivity from HLW prior to vitrification, the remaining fraction destined for salt-stone. (authors)

Engineering materials for high level radioactive waste repository

International Nuclear Information System (INIS)

Wen Zhijian

2009-01-01

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

Waste Isolation Pilot Plant Salt Decontamination Testing

Energy Technology Data Exchange (ETDEWEB)

Rick Demmer; Stephen Reese

2014-09-01

On February 14, 2014, americium and plutonium contamination was released in the Waste Isolation Pilot Plant (WIPP) salt caverns. At the request of WIPP’s operations contractor, Idaho National Laboratory (INL) personnel developed several methods of decontaminating WIPP salt, using surrogate contaminants and also americium (241Am). The effectiveness of the methods is evaluated qualitatively, and to the extent possible, quantitatively. One of the requirements of this effort was delivering initial results and recommendations within a few weeks. That requirement, in combination with the limited scope of the project, made in-depth analysis impractical in some instances. Of the methods tested (dry brushing, vacuum cleaning, water washing, strippable coatings, and mechanical grinding), the most practical seems to be water washing. Effectiveness is very high, and it is very easy and rapid to deploy. The amount of wastewater produced (2 L/m2) would be substantial and may not be easy to manage, but the method is the clear winner from a usability perspective. Removable surface contamination levels (smear results) from the strippable coating and water washing coupons found no residual removable contamination. Thus, whatever is left is likely adhered to (or trapped within) the salt. The other option that shows promise is the use of a fixative barrier. Bartlett Nuclear, Inc.’s Polymeric Barrier System (PBS) proved the most durable of the coatings tested. The coatings were not tested for contaminant entrapment, only for coating integrity and durability.

Some legal aspects on high level radioactive waste disposal in Japan

International Nuclear Information System (INIS)

Tanabe, Tomoyuki

1997-01-01

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

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

Fluidized-bed calcination of simulated commercial high-level radioactive wastes

International Nuclear Information System (INIS)

Freeby, W.A.

1975-11-01

Work is in progress at the Idaho Chemical Processing Plant to verify process flowsheets for converting simulated commercial high-level liquid wastes to granular solids using the fluidized-bed calcination process. Primary emphasis in the series of runs reported was to define flowsheets for calcining simulated Allied-General Nuclear Services (AGNS) waste and to evaluate product properties significant to calcination, solids storage, or post treatment. Pilot-plant studies using simulated high-level acid wastes representative of those to be produced by Nuclear Fuel Services, Inc. (NFS) are also included. Combined AGNS high-level and intermediate-level waste (0.26 M Na in blend) was successfully calcined when powdered iron was added (to result in a Na/Fe mole ratio of 1.0) to the feed to prevent particle agglomeration due to sodium nitrate. Long-term runs (approximately 100 hours) showed that calcination of the combined waste is practical. Concentrated AGNS waste containing sodium at concentrations less than 0.2 M were calcined successfully; concentrated waste containing 1.13 M Na calcined successfully when powdered iron was added to the feed to suppress sodium nitrate formation. Calcination of dilute AGNS waste by conventional fluid-bed techniques was unsuccessful due to the inability to control bed particle size--both particle size and bed level decreased. Fluid-bed solidification of AGNS dilute waste at conditions in which most of the calcined solids left the calciner vessel with the off-gas was successful. In such a concept, the steady-state composition of the bed material would be approximately 22 wt percent calcined solids deposited on inert particles. Calcination of simulated NFS acid waste indicated that solidification by the fluid-bed process is feasible

Modified sulfur cement solidification of low-level wastes

Energy Technology Data Exchange (ETDEWEB)

1985-10-01

This topical report describes the results of an investigation on the solidification of low-level radioactive wastes in modified sulfur cement. The work was performed as part of the Waste Form Evaluation Program, sponsored by the US Department of Energy's Low-Level Waste Management Program. Modified sulfur cement is a thermoplastic material developed by the US Bureau of Mines. Processing of waste and binder was accomplished by means of both a single-screw extruder and a dual-action mixing vessel. Waste types selected for this study included those resulting from advanced volume reduction technologies (dry evaporator concentrate salts and incinerator ash) and those which remain problematic for solidification using contemporary agents (ion exchange resins). Process development studies were conducted to ascertain optimal process control parameters for successful solidification. Maximum waste loadings were determined for each waste type and method of processing. Property evaluation testing was carried out on laboratory scale specimens in order to compare with waste form performance for other potential matrix materials. Waste form property testing included compressive strength, water immersion, thermal cycling and radionuclide leachability. Recommended waste loadings of 40 wt. % sodium sulfate and boric acid salts and 43 wt. % incinerator ash, which are based on processing and performance considerations, are reported. Solidification efficiencies for these waste types represent significant improvements over those of hydraulic cements. Due to poor waste form performance, incorporation of ion exchange resin waste in modified sulfur cement is not recommended.

Modified sulfur cement solidification of low-level wastes

International Nuclear Information System (INIS)

1985-10-01

This topical report describes the results of an investigation on the solidification of low-level radioactive wastes in modified sulfur cement. The work was performed as part of the Waste Form Evaluation Program, sponsored by the US Department of Energy's Low-Level Waste Management Program. Modified sulfur cement is a thermoplastic material developed by the US Bureau of Mines. Processing of waste and binder was accomplished by means of both a single-screw extruder and a dual-action mixing vessel. Waste types selected for this study included those resulting from advanced volume reduction technologies (dry evaporator concentrate salts and incinerator ash) and those which remain problematic for solidification using contemporary agents (ion exchange resins). Process development studies were conducted to ascertain optimal process control parameters for successful solidification. Maximum waste loadings were determined for each waste type and method of processing. Property evaluation testing was carried out on laboratory scale specimens in order to compare with waste form performance for other potential matrix materials. Waste form property testing included compressive strength, water immersion, thermal cycling and radionuclide leachability. Recommended waste loadings of 40 wt. % sodium sulfate and boric acid salts and 43 wt. % incinerator ash, which are based on processing and performance considerations, are reported. Solidification efficiencies for these waste types represent significant improvements over those of hydraulic cements. Due to poor waste form performance, incorporation of ion exchange resin waste in modified sulfur cement is not recommended

Development of a test system for high level liquid waste partitioning

Directory of Open Access Journals (Sweden)

Duan Wu H.

2015-01-01

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

High-level nuclear waste disposal

International Nuclear Information System (INIS)

Burkholder, H.C.

1985-01-01

The meeting was timely because many countries had begun their site selection processes and their engineering designs were becoming well-defined. The technology of nuclear waste disposal was maturing, and the institutional issues arising from the implementation of that technology were being confronted. Accordingly, the program was structured to consider both the technical and institutional aspects of the subject. The meeting started with a review of the status of the disposal programs in eight countries and three international nuclear waste management organizations. These invited presentations allowed listeners to understand the similarities and differences among the various national approaches to solving this very international problem. Then seven invited presentations describing nuclear waste disposal from different perspectives were made. These included: legal and judicial, electric utility, state governor, ethical, and technical perspectives. These invited presentations uncovered several issues that may need to be resolved before high-level nuclear wastes can be emplaced in a geologic repository in the United States. Finally, there were sixty-six contributed technical presentations organized in ten sessions around six general topics: site characterization and selection, repository design and in-situ testing, package design and testing, disposal system performance, disposal and storage system cost, and disposal in the overall waste management system context. These contributed presentations provided listeners with the results of recent applied RandD in each of the subject areas

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

International Nuclear Information System (INIS)

Koyama, Tadafumi.

1994-01-01

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

An Approach for the Analysis of Regulatory Analytes in High Level Radioactive Waste Stored at Hanford, Richland, Washington

International Nuclear Information System (INIS)

Wiemers, K.D.; Miller, M.; Lerchen, M.E.

1999-01-01

Radiation levels, salt concentration, and the oxidizing nature of the waste dictates modifications to the SW-846 methods. Modified methods will be used to meet target EQLs and QC currently in SW-846. Method modifications will be validated per SW-846 and HASQARD and will be documented consistent with WAC 173-303-910. The affect of modifications to holding times and storage conditions will be evaluated using techniques developed by Maskarinec and Bayne (1996). After validating the methods and performing the holding time study on a minimum of two Phase 1 candidate feed source tank wastes, DOE and Ecology will assess: whether different methods are needed, whether holding time/storage conditions should be altered, whether the high priority analyte list should be refined, and which additional tank waste needs to be characterized

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

International Nuclear Information System (INIS)

1978-04-01

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

Design features of a full-scale high-level waste vitrification system

International Nuclear Information System (INIS)

Siemens, D.H.; Bonner, W.F.

1976-08-01

A system has been designed and is currently under construction for vitrification of commercial high-level waste. The process consists of a spray calciner coupled to an in-can melter. Due to the high radiation levels expected, this equipment is designed for totally remote operation and maintenance. The in-cell arrangement of this equipment has been developed cooperatively with a nuclear fuel reprocessor. The system will be demonstrated both full scale with nonradioactive simulated waste and pilot scale with actual high-level waste

Permitting plan for the high-level waste interim storage

International Nuclear Information System (INIS)

Deffenbaugh, M.L.

1997-01-01

This document addresses the environmental permitting requirements for the transportation and interim storage of solidified high-level waste (HLW) produced during Phase 1 of the Hanford Site privatization effort. Solidified HLW consists of canisters containing vitrified HLW (glass) and containers that hold cesium separated during low-level waste pretreatment. The glass canisters and cesium containers will be transported to the Canister Storage Building (CSB) in a U.S. Department of Energy (DOE)-provided transportation cask via diesel-powered tractor trailer. Tri-Party Agreement (TPA) Milestone M-90 establishes a new major milestone, and associated interim milestones and target dates, governing acquisition and/or modification of facilities necessary for: (1) interim storage of Tank Waste Remediation Systems (TWRS) immobilized HLW (IHLW) and other canistered high-level waste forms; and (2) interim storage and disposal of TWRS immobilized low-activity tank waste (ILAW). An environmental requirements checklist and narrative was developed to identify the permitting path forward for the HLW interim storage (HLWIS) project (See Appendix B). This permitting plan will follow the permitting logic developed in that checklist

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