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Sample records for repository rock material

  1. Rock support for nuclear waste repositories

    International Nuclear Information System (INIS)

    Abramson, L.W.; Schmidt, B.

    1984-01-01

    The design of rock support for underground nuclear waste repositories requires consideration of special construction and operation requirements, and of the adverse environmental conditions in which some of the support is placed. While repository layouts resemble mines, design, construction and operation are subject to quality assurance and public scrutiny similar to what is experienced for nuclear power plants. Exploration, design, construction and operation go through phases of review and licensing by government agencies as repositories evolve. This paper discusses (1) the various stages of repository development; (2) the environment that supports must be designed for; (3) the environmental effects on support materials; and (4) alternative types of repository rock support

  2. Rock fill in a KBS-3 repository. Rock material for filling of shafts and ramps in a KBS-3V repository in the closure phase

    International Nuclear Information System (INIS)

    Pusch, Roland

    2008-09-01

    The content of large blocks in blasted rock makes it impossible to fill and compact the material effectively unless those larger than about 500 mm are removed. Tunnel Boring Machine (TBM) muck gives flat chips, that are usually not longer than a couple of decimeters, and serves better as backfill. The granulometrical composition of both types can be more suitable for effective compaction by crushing, which is hence a preferable process. Use of unsorted, unprocessed blasted rock can only be accepted if the density and physical properties, like self-compaction, are not important. Crushing of blasted rock and TBM muck for backfilling can be made in one or two steps depending on the required gradation. Placement of rock fill is best made by use of tractors with blades that push the material forwards over already placed and compacted material. The dry density of well graded rock fill effectively compacted by very heavy vibratory rollers can be as high as 2,400 kg/m3. For road compaction by ordinary vibratory rollers common dry density values are in the interval 2,050 to 2,200 kg m 3 . Blasted rock dumped and moved on site by tractors can get an average dry density of 1,600-1,800 kg/m3 without compaction. Crushed, blasted rock and TBM muck placed by tractors in horizontal layers and compacted by 5-10 t vibrating rollers in the lower part of the rooms, and moved by tractors to form inclined layers compacted by vibrating plates in the upper part, would get a dry density of 1,900-2,000 kg/m 3 . Flushing water over the rock fill in conjunction with the compaction work gives more effective densification than dry compaction. Based on recorded settlement of Norwegian rock fill dams constructed with water flushing it is estimated that the self-compaction of a 5 m high backfill of crushed rock or TBM muck causes a settlement of the top of the backfill of about 8 mm while a 200 m high shaft fill would undergo compression by more than half a meter. Repeated, strong earthquakes may

  3. Rock fill in a KBS-3 repository. Rock material for filling of shafts and ramps in a KBS-3V repository in the closure phase

    Energy Technology Data Exchange (ETDEWEB)

    Pusch, Roland (Geodevelopment International AB/SWECO AB, Lund (Sweden))

    2008-09-15

    The content of large blocks in blasted rock makes it impossible to fill and compact the material effectively unless those larger than about 500 mm are removed. Tunnel Boring Machine (TBM) muck gives flat chips, that are usually not longer than a couple of decimeters, and serves better as backfill. The granulometrical composition of both types can be more suitable for effective compaction by crushing, which is hence a preferable process. Use of unsorted, unprocessed blasted rock can only be accepted if the density and physical properties, like self-compaction, are not important. Crushing of blasted rock and TBM muck for backfilling can be made in one or two steps depending on the required gradation. Placement of rock fill is best made by use of tractors with blades that push the material forwards over already placed and compacted material. The dry density of well graded rock fill effectively compacted by very heavy vibratory rollers can be as high as 2,400 kg/m3. For road compaction by ordinary vibratory rollers common dry density values are in the interval 2,050 to 2,200 kg m3. Blasted rock dumped and moved on site by tractors can get an average dry density of 1,600-1,800 kg/m3 without compaction. Crushed, blasted rock and TBM muck placed by tractors in horizontal layers and compacted by 5-10 t vibrating rollers in the lower part of the rooms, and moved by tractors to form inclined layers compacted by vibrating plates in the upper part, would get a dry density of 1,900-2,000 kg/m3. Flushing water over the rock fill in conjunction with the compaction work gives more effective densification than dry compaction. Based on recorded settlement of Norwegian rock fill dams constructed with water flushing it is estimated that the self-compaction of a 5 m high backfill of crushed rock or TBM muck causes a settlement of the top of the backfill of about 8 mm while a 200 m high shaft fill would undergo compression by more than half a meter. Repeated, strong earthquakes may

  4. Determination of soil mechanics of salt rock as a potential backfilling material in an underground repository

    International Nuclear Information System (INIS)

    Kappei, G.

    1987-09-01

    Within the framework of the research and development project 'Backfilling and sealing of boreholes, chambers and roadways in a final dump', the Institute for Underground Dumping chose - from the broad range of possible stowing materials - the material 'salt spoil' and investigated its soil-mechanical properties in detail. Besides the implementation of soil-mechanical standard analyses (determination of the grain size distribution, bulk density, limits of storage density, proctor density, permeabilities, and shear strength) of two selected salt spoils (heap salt and rock salt spoil), the studies concentrated on the determination of the compression behaviour of salt spoil. In order to obtain data on the compaction behaviour of this material in the case of increasing stress, compression tests with obstructed lateral expansion were carried out on a series of spoil samples differing mainly in the composition of grain sizes. In addition to this, for a small number of samples of rock salt spoil, the creep behaviour at constant stress was determined after the compaction phase. (orig./RB) [de

  5. The Computational Materials Repository

    DEFF Research Database (Denmark)

    Landis, David D.; Hummelshøj, Jens S.; Nestorov, Svetlozar

    2012-01-01

    The possibilities for designing new materials based on quantum physics calculations are rapidly growing, but these design efforts lead to a significant increase in the amount of computational data created. The Computational Materials Repository (CMR) addresses this data challenge and provides...

  6. Computational Materials Repository

    DEFF Research Database (Denmark)

    Landis, David

    , different abstraction levels and enables users to analyze their own results, and allows to share data with collaborators. The approach of the Computational Materials Repository (CMR) is to convert data to an internal format that maintains the original variable names without insisting on any semantics...

  7. Rock mechanics for hard rock nuclear waste repositories

    International Nuclear Information System (INIS)

    Heuze, F.E.

    1981-09-01

    The mined geologic burial of high level nuclear waste is now the favored option for disposal. The US National Waste Terminal Storage Program designed to achieve this disposal includes an extensive rock mechanics component related to the design of the wastes repositories. The plan currently considers five candidate rock types. This paper deals with the three hard rocks among them: basalt, granite, and tuff. Their behavior is governed by geological discontinuities. Salt and shale, which exhibit behavior closer to that of a continuum, are not considered here. This paper discusses both the generic rock mechanics R and D, which are required for repository design, as well as examples of projects related to hard rock waste storage. The examples include programs in basalt (Hanford/Washington), in granitic rocks (Climax/Nevada Test Site, Idaho Springs/Colorado, Pinawa/Canada, Oracle/Arizona, and Stripa/Sweden), and in tuff

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

    International Nuclear Information System (INIS)

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

    2005-12-01

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

  9. Repository for fissile materials

    International Nuclear Information System (INIS)

    Gablin, K.A.

    1976-01-01

    A repository for holding and storing fissile or other hazardous materials either under or above the ground is provided by enclosing one or more inner containers, such as standard steel drums, in a larger, corrosion-resistant outer shell, with a layer of foamed polyurethane occupying the space therebetween. The polyurethane foam is free of voids at its interfaces with the inner container and outer shell, and adheres to and reinforces same to provide a stress skin structure. Protection is afforded by the chemical and physical characteristics of the polyurethane foam against destructive influences such as water vapor intrusion, package leakage and damaging effects of the environment, such as freezing, electrolysis, chemical and bacterial action. The outer shell is shaped to conform generally to the shape of the inner container and is made of a tube of bituminized fiber material with endcaps of exterior grade plywood treated with wood preservative. A quantity of fluorescein dye is positioned within the inner container for monitoring each package for leakage

  10. Use of the mixture of clay and crushed rock as a backfill material for low and intermediate level radioactive waste repository. Appendix 10: Republic of Korea

    International Nuclear Information System (INIS)

    Cho, W.J.; Lee, J.O.; Hahn, P.S.; Chun, K.S.

    2001-01-01

    At the time of the CRP, a repository for low and intermediate level radioactive wastes arising from nuclear power plant operation and radioisotope application in the Republic of Korea was to be constructed in the bedrock below ground surface. As the intermediate level waste cavern would contain the major part of radionuclide inventory in the cavern, the radionuclide release from the intermediate level waste cavern was therefore important from the viewpoint of disposal facility performance. The then current design concept suggested that the intermediate level waste would be emplaced into the compartment made of reinforced concrete, and the space between the concrete wall and cavern surface would be backfilled with a clay-based material. As compacted clay-based materials have a low hydraulic conductivity and the hydraulic gradient in a disposal cavern was expected to be relatively low, molecular diffusion was considered to be the principal mechanism by which radionuclides would migrate through the backfill. The mixture of calcium bentonite and crushed rock was being suggested as a candidate backfill material. This appendix summarises the KAERI research activities on the evaluation of hydraulic conductivity, radionuclide diffusion coefficient, and mechanical properties of the candidate clay-based backfill material for the intermediate level waste cavern

  11. Layout Optimization for the Repository within a discontinuous and saturated granitic rock mass

    International Nuclear Information System (INIS)

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

    2005-12-01

    The objective of the present study is a layout optimization of a single and double layer repositories within a repository site with special joint set arrangements. Single and double layer repository models, subjected to the variation of repository depth, cavern spacing, pitch, and layer spacing, are analyzed for the thermal, hydraulic, and mechanical interaction behavior during the period of 2000 years from waste emplacement. Material properties used for the granitic rock mass, rock joints, PWR spent fuel, disposal canister, compacted bentonite, backfill material, and groundwater are the data collected domestically, and foreign data are used for some of the data not available domestically. The repository model includes a saturated granitic rock mass with joints, PWR spent fuel in a disposal canister surrounded by compacted bentonite inside a deposition hole, and backfill material in the rest of the space within a repository cavern

  12. Gabbro as a host rock for a nuclear waste repository

    International Nuclear Information System (INIS)

    Ahlbom, K.; Leijon, B.; Smellie, J.; Liedholm, M.

    1992-09-01

    As an alternative to granitic rocks, gabbro and other basic rock types have been investigated with respect to their suitability to host a nuclear waste repository. The present report summarizes and examines existing geoscientific knowledge of relevance in assessing the potential merits of gabbro as a repository host rock. Implications in terms of site selection, repository construction and post-closure repository performance are also discussed. The objective of the study is to provide a basis for decisions as regards future consideration of the gabbro alternative. It is found that there are rather few gabbro bodies in Sweden, that are potentially of sufficient size to host a repository. Thus, gabbro offers little latitude as regards site selection. In comparison to siting a repository in granitic rocks, this is a major disadvantage, and it may in fact remove gabbro from further consideration. The potential advantages of gabbro refer to repository performance, and include low hydraulic conductivity and a chemical environment promoting efficient radionuclide retardation. However, results from field investigations show that groundwater flow in gabbro bodies is largely controlled by intersecting heterogeneities, in particular granitic dykes, that are significantly more conductive to water than the gabbro. In the far-field scale significant to repository performance, this may reduce or eliminate the potential effects of favourable hydraulic and chemical characteristics of the gabbro itself. In conclusion, there are apparent difficulties associated with siting a repository in gabbro, due to lack of sufficiently large gabbro bodies. On the basis of the present state of knowledge, no decisive differences can be demonstrated when comparing gabbro with granitic rocks, neither with respects to repository construction, nor as regards repository performance. (au)

  13. The rock mechanical stability of the VLJ repository

    International Nuclear Information System (INIS)

    Kuula, H.; Johansson, E.

    1991-03-01

    The aim of the study was to determine the rock mechanical stability around the VLJ repository based on the rock mechanical monitoring and rock mechanical modeling. Rock mechanical calculations were made in order to calculate the rock mass displacements and to analyze the stability around the VLJ repository The calculations were performed with three diiferent methods: continuum finite difference code FLAC, distinct element code UDEC and three dimensional distinct element code 3DEC. The first analyses were based on preliminary site investigations. The final modeling was based on investigations and rock mechanical monitoring done during the excavation. Some sensitive analyses were also performed. The modelled rock mass behaviour and the measured behaviour are generally close to each other. Both results show that the VLJ repository is rock mechanically stable. The modelled displacements and stresses were small enough to cause no instability around the rock caverns. The measured values do not indicate any discontinuous deformations like block movements or joint slip. The measured displacements in the extensometers during excavation indicates that the rock mass is even stiffer than anticipated

  14. Ground water movements around a repository. Rock mechanics analyses

    International Nuclear Information System (INIS)

    Ratigan, J.L.

    1977-09-01

    The determination and rational assessment of groundwater flow around a repository depends upon the accurate analysis of several interdependent and coupled phenomenological events occuring within the rock mass. In particular, the groundwater flow pathways (joints) are affected by the excavation and thermomechanical stresses developed within the rock mass, and the properties, of the groundwater are altered by the temperature perturbations in the rock mass. The objective of this report is to present the results of the rock mechanics analysis for the repository excavation and the thermally-induced loadings. Qualitative analysis of the significance of the rock mechanics results upon the groundwater flow is provided in this report whenever such an analysis can be performed. Non-linear rock mechanics calculations have been completed for the repository storage tunnels and the global repository domain. The rock mass has been assumed to possess orthoganol joint sets or planes of weakness with finite strength characteristics. In the local analyses of the repository storage tunnels the effects of jointorientation and repository ventilation have been examined. The local analyses indicated that storage room support requirements and regions of strength failure are highly dependent upon joint orientation. The addition of storage tunnel ventilation was noted to reduce regions of strength failure, particularly during the 30 year operational phase of the repository. Examination of the local stresses around the storage tunnels indicated the potential for perturbed hydraulic permeabilities. The permeabilities can be expected to be altered to a greater degree by the stresses resulting from excavation than from stresses which are thermally induced. The thermal loading provided by the instantaneous waste emplacement resulted in stress states and displacements quite similar to those provided by the linear waste emplacement sequence

  15. The Polar Rock Repository: Rescuing Polar Collections for New Research

    Science.gov (United States)

    Grunow, A.

    2016-12-01

    Geological field expeditions in polar regions are logistically difficult, financially expensive and can have a significant environmental impact on pristine regions. The scarcity of outcrop in Antarctica (98% ice-covered) makes previously collected rock samples very valuable to the science community. NSF recognized the need for preserving rock, dredge, and terrestrial core samples from polar areas and created the Polar Rock Repository (PRR). The PRR collection allows for full and open access to both samples and metadata via the PRR website. In addition to the physical samples and their basic metadata, the PRR archives supporting materials from the collector, field notebooks, images of the samples, field maps, air photos, thin sections and any associated bibliography/DOI's. Many of these supporting materials are unique. More than 40,000 samples are available from the PRR for scientific analysis to researchers around the globe. Most of the samples cataloged at the PRR were collected more than 30 years ago, some more than 100 years ago. The rock samples and metadata are made available online through an advanced search engine for the PRR website. This allows scientists to "drill down" into search results using categories and look-up object fields similar to websites like Amazon. Results can be viewed in a table, downloaded as a spreadsheet, or plotted on an interactive map that supports display of satellite imagery and bathymetry layers. Samples can be requested by placing them in the `shopping cart'. These old sample collections have been repeatedly used by scientists from around the world. One data request involved locating coal deposits in Antarctica for a global compilation and another for looking at the redox state of batholithic rocks from the Antarctic Peninsula using magnetic susceptibilities of PRR rocks. Sample usage has also included non-traditional geologic studies, such as a search for monopoles in Cenozoic volcanic samples, and remote sensing

  16. 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

  17. Proceedings of the scientific visit on crystalline rock repository development.

    Energy Technology Data Exchange (ETDEWEB)

    Mariner, Paul E.; Hardin, Ernest L.; Miksova, Jitka [RAWRA, Czech Republic

    2013-02-01

    A scientific visit on Crystalline Rock Repository Development was held in the Czech Republic on September 24-27, 2012. The visit was hosted by the Czech Radioactive Waste Repository Authority (RAWRA), co-hosted by Sandia National Laboratories (SNL), and supported by the International Atomic Energy Agency (IAEA). The purpose of the visit was to promote technical information exchange between participants from countries engaged in the investigation and exploration of crystalline rock for the eventual construction of nuclear waste repositories. The visit was designed especially for participants of countries that have recently commenced (or recommenced) national repository programmes in crystalline host rock formations. Discussion topics included repository programme development, site screening and selection, site characterization, disposal concepts in crystalline host rock, regulatory frameworks, and safety assessment methodology. Interest was surveyed in establishing a %E2%80%9Cclub,%E2%80%9D the mission of which would be to identify and address the various technical challenges that confront the disposal of radioactive waste in crystalline rock environments. The idea of a second scientific visit to be held one year later in another host country received popular support. The visit concluded with a trip to the countryside south of Prague where participants were treated to a tour of the laboratory and underground facilities of the Josef Regional Underground Research Centre.

  18. Thermal characteristics of rocks for high-level waste repository

    International Nuclear Information System (INIS)

    Shimooka, Kenji; Ishizaki, Kanjiro; Okamoto, Masamichi; Kumata, Masahiro; Araki, Kunio; Amano, Hiroshi

    1980-12-01

    Heat released by the radioactive decay of high-level waste in an underground repository causes a long term thermal disturbance in the surrounding rock mass. Several rocks constituting geological formations in Japan were gathered and specific heat, thermal conductivity, thermal expansion coefficient and compressive strength were measured. Thermal analysis and chemical analysis were also carried out. It was found that volcanic rocks, i.e. Andesite and Basalt had the most favorable thermal characteristics up to around 1000 0 C and plutonic rock, i.e. Granite had also favorable characteristics under 573 0 C, transition temperature of quartz. Other igneous rocks, i.e. Rhyolite and Propylite had a problem of decomposition at around 500 0 C. Sedimentary rocks, i.e. Zeolite, Tuff, Sandstone and Diatomite were less favorable because of their decomposition, low thermal conductivity and large thermal expansion coefficient. (author)

  19. PRINCIPLE ROCK TYPES FOR RADIOACTIVE WASTE REPOSITORIES

    Directory of Open Access Journals (Sweden)

    Sibila Borojević Šostarić

    2012-07-01

    Full Text Available Underground geological storage of high- and intermediate/low radioactive waste is aimed to represent a barrier between the surface environment and potentially hazardous radioactive elements. Permeability, behavior against external stresses, chemical reacatibility and absorption are the key geological parameters for the geological storage of radioactive waste. Three principal rock types were discussed and applied to the Dinarides: (1 evaporites in general, (2 shale, and (3 crystalline basement rocks. (1 Within the Dinarides, evaporite formations are located within the central part of a Carbonate platform and are inappropriate for storage. Offshore evaporites are located within diapiric structures of the central and southern part of the Adriatic Sea and are covered by thick Mesozoic to Cenozoic clastic sediment. Under very specific circumstances they can be considered as potential site locations for further investigation for the storage of low/intermediate level radioactive wast e. (2 Thick flysch type formation of shale to phyllite rocks are exposed at the basement units of the Petrova and Trgovska gora regions whereas (3 crystalline magmatic to metamorphic basement is exposed at the Moslavačka Gora and Slavonian Mts. regions. For high-level radioactive waste, basement phyllites and granites may represent the only realistic potential option in the NW Dinarides.

  20. Rock grouting. Current competence and development for the final repository

    International Nuclear Information System (INIS)

    Emmelin, Ann; Brantberger, Martin; Eriksson, Magnus; Gustafson, Gunnar; St ille, Haakan

    2007-06-01

    The report aims at presenting the overall state of grouting competence and development relating to the final repository and at motivating and giving detail to the grouting sections presented in the 2007 version of the overall SKB report 'Programme for research, development and demonstration of methods for the management and disposal of nuclear waste' that is presented to the government every three years. The report offers suggestions for principles for planning, design and execution of grouting and describes the further work thought to be necessary in order to meet the requirements of the final repository, that are currently given as working premises. This report does not aim to, and cannot, describe the grouting processes in detail. For details of current concepts, experience and development work, a list of references is provided. In Chapter 2, the task of sealing the underground repository is examined and an overall approach presented. Although the requirements related to this task are preliminary, it is made evident that they concern both the actual grouting results and the process leading to the achievement of these results. Chapter 3 is a conceptual description of grouting and the factors that govern the spreading of grout in the rock mass. It is intended as an introduction to Chapters 4-6, which describe the state of grouting competence and the tools available for the sealing of the final repository facility. Both common practice and cutting-edge research are dealt with in these chapters, mainly relying on references where available. Chapters 4 and 5 focus on the system consisting of the fundamental components the rock mass, the grout materials and the grouting technology, and how these system components interact whilst, in Chapter 6, the rock/grout technical system is viewed in a brief organizational context. Based on the requirements on results and the overall grouting process on the one hand and the current competence in grouting theory and practice on the

  1. Rock grouting. Current competence and development for the final repository

    Energy Technology Data Exchange (ETDEWEB)

    Emmelin, Ann (Swedish Nuclear Fuel and Waste Management Co., Stockholm (SE)); Brantberger, Martin (Ramboell (SE)); Eriksson, Magnus (Vattenfall Power Consultant (SE)); Gustafson, Gunnar (Chalmers Univ. of Technology, Goeteborg (SE)); Stille, Haakan (Royal Inst. of Technology, Stockholm (SE))

    2007-06-15

    The report aims at presenting the overall state of grouting competence and development relating to the final repository and at motivating and giving detail to the grouting sections presented in the 2007 version of the overall SKB report 'Programme for research, development and demonstration of methods for the management and disposal of nuclear waste' that is presented to the government every three years. The report offers suggestions for principles for planning, design and execution of grouting and describes the further work thought to be necessary in order to meet the requirements of the final repository, that are currently given as working premises. This report does not aim to, and cannot, describe the grouting processes in detail. For details of current concepts, experience and development work, a list of references is provided. In Chapter 2, the task of sealing the underground repository is examined and an overall approach presented. Although the requirements related to this task are preliminary, it is made evident that they concern both the actual grouting results and the process leading to the achievement of these results. Chapter 3 is a conceptual description of grouting and the factors that govern the spreading of grout in the rock mass. It is intended as an introduction to Chapters 4-6, which describe the state of grouting competence and the tools available for the sealing of the final repository facility. Both common practice and cutting-edge research are dealt with in these chapters, mainly relying on references where available. Chapters 4 and 5 focus on the system consisting of the fundamental components the rock mass, the grout materials and the grouting technology, and how these system components interact whilst, in Chapter 6, the rock/grout technical system is viewed in a brief organizational context. Based on the requirements on results and the overall grouting process on the one hand and the current competence in grouting theory and

  2. Foreign materials in the repository. Update of estimated quantities

    International Nuclear Information System (INIS)

    Karvonen, T.

    2011-06-01

    A variety of materials are used during the construction process and the operation of the repository for spent nuclear fuel at Olkiluoto in Eurajoki, Finland. In addition to materials necessary for the construction and operation, some materials may be transported into the repository with the ventilation air, as emissions from vehicles etc. Both of these two types of materials are considered here and both introduced quantities and the quantities that remain after the closure in the repository are estimated here based on the most recent information. This work is intended to update the previous estimations, and it takes advantage of the experience collected during the construction of the underground rock characterisation facility called ONKALO at Olkiluoto. The implemented quantities as well as designs and preliminary designs have been used in calculating the quantities of the foreign materials. The estimations made in this report are specific to a KBS-3V type repository. In some cases more generic information has been used, particularly when the relevant quantities have not been monitored in ONKALO. The estimations are based on the new repository layout produced in 2010 and consider the latest plans for grouting and rock support. As all of these plans are not final some quantities may change in the future. As the repository layout may still go through some changes this report also provides the foreign materials for a hundred meters of different deposition tunnels designed for the OL and LO type canisters1. The results have also been calculated for a space demanded by a deposition tunnel end plug and the tunnel lengths before and after one. The most significant foreign materials are certain accessory minerals of the clay materials followed by organic materials (including the organic carbon from the clay materials), cement, steel and silica. (orig.)

  3. Foreign materials in the repository - update of estimated quantities

    International Nuclear Information System (INIS)

    Hagros, A.

    2007-03-01

    In a repository for spent nuclear fuel, a variety of materials are used during the construction process and during the operation of the repository. In addition to materials necessary for the construction and operation, some materials may be transported into the repository through the ventilation air, as emissions from vehicles, as waste produced by the staff etc. Both of these two types of materials are considered here and their quantities - both the introduced quantities and the quantities that remain after closure - in the repository constructed at Olkiluoto in Eurajoki, Finland are estimated here based on new information. This work is intended to update the estimations that have been made previously, and it takes advantage of the experience collected during the construction of the underground rock characterisation facility ONKALO at Olkiluoto. During this construction process, the quantities of the different construction materials introduced into the underground openings have been monitored and they form a basis for estimating the quantities to be used in the future. The estimations made in this report are specific to a KBS-3V type repository and to the Olkiluoto site, although in some cases more generic information has been used, particularly when the relevant quantities have not been monitored in the ONKALO. The estimations are based on the new repository layout produced in 2006 and consider the latest plans for grouting and rock support. As these plans are generally not final yet, several different alternative plans are assumed when necessary. Also two different strategies for the backfilling of the tunnels are considered. The most significant differences with respect to the results of an earlier estimation are related to the materials used in grouting, shotcreting and in support bolts. In the cases where a mixture of bentonite and crushed rock is the used backfill alternative, gypsum and cement are the materials with the largest quantities remaining in the

  4. Waste-rock interactions in the immediate repository

    International Nuclear Information System (INIS)

    McCarthy, G.J.

    1977-01-01

    The high level wastes (HLW's) to be placed underground in rock formations will contain significant amounts of radioactive decay heat for the first hundred-or-so years of isolation. Several physical-chemical changes analogous to natural geochemical processes can occur during this ''thermal period.'' The waste canister can act as a heat source and cause changes in the mineralogy and properties of the surrounding rocks. Geochemically, this is ''contact metamorphism.'' In the event that the canister is corroded and breached, chemical reactions can occur between the HLW, the surrounding rock and possibly the remains of the canister. In a dry repository which has not been backfilled (and thus pressurized) these interactions could be slow at best and with rates decreasing rapidly as the HLW cools. However, significant interactions can occur in years, months or even days under hydrothermal conditions. These conditions could be created by the combination of HLW heat, overburden pressure and water mobilized from the rocks or derived from groundwater intrusion. At the end of the thermal period these interaction products would constitute the actual HLW form (or ''source term'') subject to the low temperature leaching and migration processes under investigation in other laboratories. It is quite possible that these interaction product waste forms will have superior properties compared to the original HLW. Experimental programs initiated at Penn State during the last year aim at determining the nature of any chemical or mineralogical changes in, or interactions between, HLW solids and host rocks under various repository ambients. The accompanying figures describe the simulated HLW forms and the experimental approach and techniques. Studies with basalts as the repository rock are supported by Rockwell Hanford Operations and with shales by the Office of Waste Isolation

  5. Lithophysal Rock Mass Mechanical Properties of the Repository Host Horizon

    International Nuclear Information System (INIS)

    D. Rigby

    2004-01-01

    The purpose of this calculation is to develop estimates of key mechanical properties for the lithophysal rock masses of the Topopah Spring Tuff (Tpt) within the repository host horizon, including their uncertainties and spatial variability. The mechanical properties to be characterized include an elastic parameter, Young's modulus, and a strength parameter, uniaxial compressive strength. Since lithophysal porosity is used as a surrogate property to develop the distributions of the mechanical properties, an estimate of the distribution of lithophysal porosity is also developed. The resulting characterizations of rock parameters are important for supporting the subsurface design, developing the preclosure safety analysis, and assessing the postclosure performance of the repository (e.g., drift degradation and modeling of rockfall impacts on engineered barrier system components)

  6. Radioactive waste repositories in hard rock aquifers--hydrodynamic aspects

    International Nuclear Information System (INIS)

    Thunvik, R.; Braester, C.

    1984-01-01

    A mathematical model for mass and heat flow and a computer program have been developed to demonstrate the effect of heat released from a hypothetical radioactive waste repository on the groundwater flow regime. The model, based on the continuum approach, conceptualizes the fracture pattern and the solid blocks as two overlapping continua and consists of a set of coupled nonlinear partial differential equations. The general form of the model is three-dimensional and can treat the fluid and rock either as two separate media with a quasi-steady exchange of heat between them or as a single equivalent medium with instantaneous thermal equilibrium. Numerical solutions have been obtained by the Galerkin finite element method. Examples have been presented for topographically different locations of the repository: below a horizontal ground surface, below a hill crest, below a hillside, and close to major fractures. The effects of constant permeability and porosity or downward decreasing with depth as well as the effect of anisotropic permeability have been investigated. Solutions include the velocity field, path lines, and traveling times of water particles passing the repository and the temperature distribution. The examples have been worked out for a two-dimensional flow domain, assuming that instantaneous thermal equilibrium takes place. This assumption was found to be justified by the relatively low flow velocities that occurred in the examples. Except for the location close to a major draining fracture, heat released from the radioactive waste repository may have a significant influence on the flow regime around the repository

  7. Analysis on one underground nuclear waste repository rock mass in USA

    International Nuclear Information System (INIS)

    Ha Qiuling; Zhang Tiantian

    2012-01-01

    When analyzing the rock mass of a underground nuclear waste repository, the current studies are all based on the loading mechanical condition, and the unloading damage of rock mass is unconsidered. According to the different mechanical condition of actual engineering rock mass of loading and unloading, this paper implements a comprehensive analysis on the rock mass deformation of underground nuclear waste repository through the combination of present loading and unloading rock mass mechanics. It is found that the results of comprehensive analysis and actual measured data on the rock mass deformation of underground nuclear waste repository are basically the same, which provide supporting data for the underground nuclear waste repository. (authors)

  8. Characterization of nearfield rock - A basis for comparison of repository concepts

    International Nuclear Information System (INIS)

    Pusch, R.; Hoekmark, H.

    1991-12-01

    The hydraulic conductivity of the nearfield rock controls the rate of wetting of adjacent buffer material as well as the rate of degradation of its smectite content and of the transport of radionuclides from the buffer/rock interface. Comparison of different repository concepts with respect to the function of the nearfield rock requires a common rock structure model, which is suggested in the report. Applying this model and 2D and 3D numerical calculations for evaluation of stress-induced structural changes, major differences between the three concepts VDH, KBS3 and VLH concerning the hydraulic conductivity of the nearfield have been identified. The importance of the orientation of the excavations turns out to be particularly obvious. Further development of the rock structure model is concluded to offer ways of quantifying more accurately the damaging effects of blasting and TBM-drilling. (au)

  9. Thermal Analysis of a Nuclear Waste Repository in Argillite Host Rock

    Science.gov (United States)

    Hadgu, T.; Gomez, S. P.; Matteo, E. N.

    2017-12-01

    Disposal of high-level nuclear waste in a geological repository requires analysis of heat distribution as a result of decay heat. Such an analysis supports design of repository layout to define repository footprint as well as provide information of importance to overall design. The analysis is also used in the study of potential migration of radionuclides to the accessible environment. In this study, thermal analysis for high-level waste and spent nuclear fuel in a generic repository in argillite host rock is presented. The thermal analysis utilized both semi-analytical and numerical modeling in the near field of a repository. The semi-analytical method looks at heat transport by conduction in the repository and surroundings. The results of the simulation method are temperature histories at selected radial distances from the waste package. A 3-D thermal-hydrologic numerical model was also conducted to study fluid and heat distribution in the near field. The thermal analysis assumed a generic geological repository at 500 m depth. For the semi-analytical method, a backfilled closed repository was assumed with basic design and material properties. For the thermal-hydrologic numerical method, a repository layout with disposal in horizontal boreholes was assumed. The 3-D modeling domain covers a limited portion of the repository footprint to enable a detailed thermal analysis. A highly refined unstructured mesh was used with increased discretization near heat sources and at intersections of different materials. All simulations considered different parameter values for properties of components of the engineered barrier system (i.e. buffer, disturbed rock zone and the host rock), and different surface storage times. Results of the different modeling cases are presented and include temperature and fluid flow profiles in the near field at different simulation times. Sandia National Laboratories is a multimission laboratory managed and operated by National Technology and

  10. Evaluation of Five Sedimentary Rocks Other Than Salt for Geologic Repository Siting Purposes

    Energy Technology Data Exchange (ETDEWEB)

    Croff, A.G.; Lomenick, T.F.; Lowrie, R.S.; Stow, S.H.

    2003-11-15

    The US Department of Energy (DOE), in order to increase the diversity of rock types under consideration by the geologic disposal program, initiated the Sedimary ROck Program (SERP), whose immediate objectiv eis to evaluate five types of secimdnary rock - sandstone, chalk, carbonate rocks (limestone and dolostone), anhydrock, and shale - to determine the potential for siting a geologic repository. The evaluation of these five rock types, together with the ongoing salt studies, effectively results in the consideration of all types of relatively impermeable sedimentary rock for repository purposes. The results of this evaluation are expressed in terms of a ranking of the five rock types with respect to their potential to serve as a geologic repository host rock. This comparative evaluation was conducted on a non-site-specific basis, by use of generic information together with rock evaluation criteria (RECs) derived from the DOE siting guidelines for geologic repositories (CFR 1984). An information base relevant to rock evaluation using these RECs was developed in hydrology, geochemistry, rock characteristics (rock occurrences, thermal response, rock mechanics), natural resources, and rock dissolution. Evaluation against postclosure and preclosure RECs yielded a ranking of the five subject rocks with respect to their potential as repository host rocks. Shale was determined to be the most preferred of the five rock types, with sandstone a distant second, the carbonate rocks and anhydrock a more distant third, and chalk a relatively close fourth.

  11. Evaluation of Five Sedimentary Rocks Other Than Salt for Geologic Repository Siting Purposes

    International Nuclear Information System (INIS)

    Croff, A.G.; Lomenick, T.F.; Lowrie, R.S.; Stow, S.H.

    2003-01-01

    The US Department of Energy (DOE), in order to increase the diversity of rock types under consideration by the geologic disposal program, initiated the Sedimary ROck Program (SERP), whose immediate objectiv eis to evaluate five types of secimdnary rock - sandstone, chalk, carbonate rocks (limestone and dolostone), anhydrock, and shale - to determine the potential for siting a geologic repository. The evaluation of these five rock types, together with the ongoing salt studies, effectively results in the consideration of all types of relatively impermeable sedimentary rock for repository purposes. The results of this evaluation are expressed in terms of a ranking of the five rock types with respect to their potential to serve as a geologic repository host rock. This comparative evaluation was conducted on a non-site-specific basis, by use of generic information together with rock evaluation criteria (RECs) derived from the DOE siting guidelines for geologic repositories (CFR 1984). An information base relevant to rock evaluation using these RECs was developed in hydrology, geochemistry, rock characteristics (rock occurrences, thermal response, rock mechanics), natural resources, and rock dissolution. Evaluation against postclosure and preclosure RECs yielded a ranking of the five subject rocks with respect to their potential as repository host rocks. Shale was determined to be the most preferred of the five rock types, with sandstone a distant second, the carbonate rocks and anhydrock a more distant third, and chalk a relatively close fourth.

  12. Interim rock mass properties and conditions for analyses of a repository in crystalline rock

    International Nuclear Information System (INIS)

    Tammemagi, H.Y.; Chieslar, J.D.

    1985-03-01

    A summary of rock properties for generic crystalline rock is compiled from literature sources to provide the input data for analyses of a conceptual repository in crystalline rock. Frequency histograms, mean values and ranges of physical, mechanical, thermal, and thermomechanical properties, and the dependence of these properties on temperature are described. A description of the hydrogeologic properties of a crystalline rock mass and their dependence on depth is provided. In addition, the temperature gradients, mean annual surface temperature, and in situ stress conditions are summarized for the three regions of the United States currently under consideration to host a crystalline repository; i.e., the North Central, Northeastern, and Southeastern. Brief descriptions of the regional geology are also presented. Large-scale underground experiments in crystalline rock at Stripa, Sweden, and in Climax Stock in Nevada, are reviewed to assess whether the rock properties presented in this report are representative of in situ conditions. The suitability of each rock property and the sufficiency of its data base are described. 110 refs., 27 figs., 4 tabs

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

    International Nuclear Information System (INIS)

    Neretnieks, I.

    1990-12-01

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

  14. 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

  15. Calculations of the Temperature Evolution of a Repository for Spent Fuel in Crystalline and Sedimentary Rocks

    International Nuclear Information System (INIS)

    Sato, R.; Sasaki, T.; Ando, K.; Smith, P.A.; Schneider, J.W.

    1998-08-01

    Thermal evolution is a factor influencing repository design, and must be considered in safety assessment, since many of the processes that affect the long-term safety are temperature dependent. This report presents calculations of the thermal evolution of a repository for spent nuclear fuel. The calculations are based on a provisional repository near-field design in which spent fuel is encapsulated in composite copper-steel canisters, which are emplaced centrally along the horizontal axes of repository tunnels, with the space around the canisters backfilled with bentonite. The temperature of these near-field components varies with time, due to the radiogenic heat produced by the spent fuel. The rate of heat production per canister depends on the initial composition of the fuel, its reactor history, the period of intermediate storage before final disposal and the loading of the canisters. The rate decreases with time, as shorter-lived radionuclides decay. The base-case calculation considers spent fuel that is assumed to generate 1000 W per canister, 40 years after unloading of the fuel from the reactor. The results of the base case calculation indicate that the temperatures at the bentonite/host rock interface, at the centre of the bentonite and at the bentonite/canister interface rise to 98 o C, 103 o C and 126 o C, respectively, before declining towards the ambient temperature of the host rock which, in the base case, is taken to be the crystalline basement of Northern Switzerland. In addition to the base case, parameter variations are examined that investigate the sensitivity of thermal evolution to alternative heat output, design specifications and to uncertainties in material properties. Key findings include (i), that an increase in heat generation to 1500 W per canister 40 years after unloading results in a significant increase of repository temperatures (e.g. at the bentonite/host rock interface, an increase of 22 o C is observed), (ii), that a decrease in

  16. A Rock Mechanics and Coupled Hydro mechanical Analysis of Geological Repository of High Level Nuclear Waste in Fractured Rocks

    International Nuclear Information System (INIS)

    Min, Kibok

    2011-01-01

    This paper introduces a few case studies on fractured hard rock based on geological data from Sweden, Korea is one of a few countries where crystalline rock is the most promising rock formation as a candidate site of geological repository of high level nuclear waste. Despite the progress made in the area of rock mechanics and coupled hydro mechanics, extensive site specific study on multiple candidate sites is essential in order to choose the optimal site. For many countries concerned about the safe isolation of nuclear wastes from the biosphere, disposal in a deep geological formation is considered an attractive option. In geological repository, thermal loading continuously disturbs the repository system in addition to disturbances a recent development in rock mechanics and coupled hydro mechanical study using DFN(Discrete Fracture Network) - DEM(Discrete Element Method) approach mainly applied in hard, crystalline rock containing numerous fracture which are main sources of deformation and groundwater flow

  17. The United States Polar Rock Repository: A geological resource for the Earth science community

    Science.gov (United States)

    Grunow, Annie M.; Elliot, David H.; Codispoti, Julie E.

    2007-01-01

    The United States Polar Rock Repository (USPRR) is a U. S. national facility designed for the permanent curatorial preservation of rock samples, along with associated materials such as field notes, annotated air photos and maps, raw analytic data, paleomagnetic cores, ground rock and mineral residues, thin sections, and microfossil mounts, microslides and residues from Polar areas. This facility was established by the Office of Polar Programs at the U. S. National Science Foundation (NSF) to minimize redundant sample collecting, and also because the extreme cold and hazardous field conditions make fieldwork costly and difficult. The repository provides, along with an on-line database of sample information, an essential resource for proposal preparation, pilot studies and other sample based research that should make fieldwork more efficient and effective. This latter aspect should reduce the environmental impact of conducting research in sensitive Polar Regions. The USPRR also provides samples for educational outreach. Rock samples may be borrowed for research or educational purposes as well as for museum exhibits.

  18. Experiments on thermal conductivity in buffer materials for geologic repository

    International Nuclear Information System (INIS)

    Kanno, T.; Yano, T.; Wakamatsu, H.; Matsushima, E.

    1989-01-01

    Engineered barriers for geologic disposal for HLW are planned to consist of canister, overpack and buffer elements. One of important physical characteristics of buffer materials is determining temperature profiles within the near field in a repository. Buffer materials require high thermal conductivity to disperse radiogenic heat away to the host rock. As the buffer materials, compacted blocks of the mixture of sodium bentonite and sand have been the most promising candidate in some countries, e.g. Sweden, Switzerland and Japan. The authors have been carrying out a series of thermal dispersion experiments to evaluate thermal conductivity of bentonite/quartz sand blocks. In this study, the following two factors considered to affect thermal properties of the near field were examined: effective thermal conductivities of buffer materials, and heat transfer characteristics of the gap between overpack and buffer materials

  19. Principal organic materials in a repository for spent nuclear fuel

    International Nuclear Information System (INIS)

    Hallbeck, Lotta

    2010-01-01

    The largest pool of organic material in a repository at closure is the organic material in the bentonite in buffer and backfill. It is impossible to make any assumptions as to how much of this material will be available for biodegradation, since the character of the material is unknown. However, it is unlikely that this organic material can dissolve in groundwater unless the bentonite loses its swelling capacity. The second largest pool will be the biofilms formed on the rock surfaces. This assumption presupposes that no cleaning is undertaken before repository closure. The third largest pool is the organic material produced by microorganisms using hydrogen from the anaerobic corrosion of iron in steel as an energy source. The following provides summary descriptions of the different pools of organic material that will remain in the repository: 1. Microorganisms. Their effect would mainly be to reduce the redox potential soon after repository closure. They may contribute to the depletion of the oxygen entrapped during repository construction, an effect that would not jeopardise repository stability. If the dominant microorganisms in the anaerobic environment are sulphate-reducing bacteria, oxidation of organic material would lead to the formation of HS - . The produced sulphide could corrode the copper canisters under anaerobic conditions if it reaches them. Another effect of microorganisms would be to increase the complexing capacity of the groundwater due to excreted metabolites. The impact of these compounds is not yet clear, although it will surely not be very important, due to the small amounts of such substances. 2. Materials in the ventilation air. Their effect will probably be to help maintain reducing conditions in the area, although this effect will likely be minimal or negligible. 3. Construction materials. Among these materials, we emphasise the organic materials present in concrete, asphalt, bentonite, and wood. Hydrocarbons from asphalt may help reduce

  20. Principal organic materials in a repository for spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Hallbeck, Lotta (Microbial Analytics Sweden AB, Moelnlycke (Sweden))

    2010-01-15

    The largest pool of organic material in a repository at closure is the organic material in the bentonite in buffer and backfill. It is impossible to make any assumptions as to how much of this material will be available for biodegradation, since the character of the material is unknown. However, it is unlikely that this organic material can dissolve in groundwater unless the bentonite loses its swelling capacity. The second largest pool will be the biofilms formed on the rock surfaces. This assumption presupposes that no cleaning is undertaken before repository closure. The third largest pool is the organic material produced by microorganisms using hydrogen from the anaerobic corrosion of iron in steel as an energy source. The following provides summary descriptions of the different pools of organic material that will remain in the repository: 1. Microorganisms. Their effect would mainly be to reduce the redox potential soon after repository closure. They may contribute to the depletion of the oxygen entrapped during repository construction, an effect that would not jeopardise repository stability. If the dominant microorganisms in the anaerobic environment are sulphate-reducing bacteria, oxidation of organic material would lead to the formation of HS-. The produced sulphide could corrode the copper canisters under anaerobic conditions if it reaches them. Another effect of microorganisms would be to increase the complexing capacity of the groundwater due to excreted metabolites. The impact of these compounds is not yet clear, although it will surely not be very important, due to the small amounts of such substances. 2. Materials in the ventilation air. Their effect will probably be to help maintain reducing conditions in the area, although this effect will likely be minimal or negligible. 3. Construction materials. Among these materials, we emphasise the organic materials present in concrete, asphalt, bentonite, and wood. Hydrocarbons from asphalt may help reduce

  1. Sealing of rock fractures around HLW repositories, 2

    International Nuclear Information System (INIS)

    Chigira, Masahiro

    1993-01-01

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

  2. Investigations of possibilities to dispose of spent nuclear fuel in Lithuania: a model case. Volume 2, Concept of Repository in Crystalline Rocks

    International Nuclear Information System (INIS)

    Motiejunas, S.; Poskas, P.

    2005-01-01

    The aim is to present the generic repository concept in crystalline rocks in Lithuania and cost assessment of the disposal of spent nuclear fuel and long-lived intermediate level waste in this repository. Due to limited budget of this project the repository concept development for Lithuania was based mostly on the experience of foreign countries. In this Volume a review of the existing information on disposal concept in crystalline rocks from various countries is presented. Described repository concept for crystalline rocks in Lithuania covers repository layout, backfill, canister, construction materials and auxiliary buildings. Costs calculations for disposal of spent nuclear fuel and long-lived intermediate-level wastes from Ignalina NPP are presented too. Thermal, criticality and other important disposal evaluations for RBMK-1500 spent nuclear fuel emplaced in copper canister were performed and described

  3. Release consequence analysis for a hypothetical geologic radioactive waste repository in hard rock

    International Nuclear Information System (INIS)

    1979-12-01

    This report makes an evaluation of the long-term behaviour of the wastes placed in a hard rock repository. Impacts were analyzed for the seven reference fuel cycles of WG 7. The reference repository for this study is for granitic rock or gneiss as the host rock. The descriptions of waste packages and repository facilities used in this study represent only one of many possible designs based on the multiple barriers concept. The repository's size is based on a nuclear economy producing 100 gigawatts of electricity per year for 1 year. The objective of the modeling efforts presented in this study is to predict the rate of transport of radioactive contaminants from the repository through the geosphere to the biosphere and thus determine an estimate of the potential dose to humans so that the release consequence impacts of the various fuel cycles can be compared. Currently available hydrologic, leach, transport, and dose models were used in this study

  4. Rock mass modification around a nuclear waste repository in welded tuff

    International Nuclear Information System (INIS)

    Mack, M.G.; Brandshaug, T.; Brady, B.H.

    1989-08-01

    This report presents the results of numerical analyses to estimate the extent of rock mass modification resulting from the presence of a High Level Waste (HLW) repository. Changes in rock mass considered are stresses and joint deformations resulting from disposal room excavation and thermal efffects induced by the heat generated by nuclear waste. rock properties and site conditions are taken from the Site Characterization Plan Conceptual Design Report for the potential repository site at Yucca Mountain, Nevada. Analyses were conducted using boundary element and distinct element methods. Room-scale models and repository-scale models were investigated for up to 500 years after waste emplacement. Results of room-scale analyses based on the thermoelastic boundary element model indicate that a zone of modified rock develops around the disposal rooms for both vertical and horizontal waste emplacement. This zone is estimated to extend a distance of roughly two room diameters from the room surface. Results from the repository-scale model, which are based on the thermoelastic boundary element model and the distinct element model, indicate a zone with modified rock mass properties starting approximately 100 m above and below the repository, with a thickness of approximately 200 m above and 150 m below the repository. Slip-prone subhorizontal features are shown to have a substantial effect on rock mass response. The estimates of rock mass modification reflect uncertainties and simplifying assumptions in the models. 32 refs., 57 figs., 1 tab

  5. Characterization of cement-based ancient building materials in support of repository seal materials studies

    International Nuclear Information System (INIS)

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

    1983-12-01

    Ancient mortars and plasters collected from Greek and Cypriot structures dating to about 5500 BC have been investigated because of their remarkable durability. The characteristics and performance of these and other ancient cementitious materials have been considered in the light of providing information on longevity of concrete materials for sealing nuclear waste geological repositories. The matrices of these composite materials have been characterized and classified into four categories: (1) gypsum cements; (2) hydraulic hydrated lime and hydrated-lime cements; (3) hydraulic aluminous and ferruginous hydrated-lime cements (+- siliceous components); and (4) pozzolana/hydrated-lime cements. Most of the materials investigated, including linings of ore-washing basins and cisterns used to hold water, are in categories (2) and (3). The aggregates used included carbonates, sandstones, shales, schists, volcanic and pyroclastic rocks, and ore minerals, many of which represent host rock types of stratigraphic components of a salt repository. Numerous methods were used to characterize the materials chemically, mineralogically, and microstructurally and to elucidate aspects of both the technology that produced them and their response to the environmental exposure throughout their centuries of existence. Their remarkable properties are the result of a combination of chemical (mineralogical) and microstructural factors. Durability was found to be affected by matrix mineralogy, particle size and porosity, and aggregate type, grading, and proportioning, as well as method of placement and exposure conditions. Similar factors govern the potential for durability of modern portland cement-containing materials, which are candidates for repository sealing. 29 references, 29 figures, 6 tables

  6. Microbiologically mediated processes in a repository sited in a clay host rock

    International Nuclear Information System (INIS)

    Schwyn, B.; Leupin, O. X.; Bagnoud, A.; Bernier-Latmani, R.

    2012-01-01

    Document available in extended abstract form only. Because of their favourable retention properties for radionuclides, clay-rich sediments are being considered in Switzerland as host rocks for the geological disposal of high, intermediate- and low-level radioactive waste. Compacted bentonite is foreseen as backfill material in the high level waste repository whereas for intermediate- and low-level waste the near field will mainly consist of cementitious material. The evolution of both types of repositories, which includes re-saturation, heat generation (only high level waste), near field degradation, gas production and radionuclide release may be impacted by microbial activity and vice versa. In this respect questions arise such as: - Are microorganisms present in a repository and its host rock? - Under which condition are microorganisms active in and around a repository? - In which processes are microorganisms involved? Various in situ experiments in a wide range of geological environments have evidenced the presence of microorganisms. Whether the microorganisms found in these in situ experiments are indigenous or introduced by drilling or/and excavation activities is still controversial. However, recent findings suggest the presence of indigenous microorganisms in Opalinus Clay. To conclusively answer the question about the origin of microorganisms, an international investigation programme has been launched to probe rock samples from the Underground Rock Laboratory at Mont Terri. So far, no metabolic activity has been observed in undisturbed clay rocks. Such activity may have ceased during diagenetic compaction of the sediment as suggested by the pore water composition measured in the Callovian-Oxfordian clayey formation of Bure (France). For the safety case of a repository the origin of microorganisms is of minor importance compared to the understanding of the conditions under which they might be metabolically active. Pore size distribution and connectivity can

  7. Evaluation of the damages in rocks caused by the construction of a repository

    International Nuclear Information System (INIS)

    Devillers, C.; Escalier des Orres, P.

    1988-12-01

    The Commissariat a l'Energie Atomique (French Atomic Energy Commission) has conducted a bibliographic study of the damages in the rock caused by the construction of a repository, and several hydraulic simulations, to appreciate the influence of these damages on the safety of the repository. These studies have led to the proposal of construction techniques in accordance safety requirements and industrial feasibility [fr

  8. Importance of creep failure of hard rock in the near field of a nuclear waste repository

    Energy Technology Data Exchange (ETDEWEB)

    Blacic, J D [Los Alamos National Laboratory, NM, (USA)

    1982-12-31

    Potential damage resulting from slow creep deformation intuitively seems unlikely for a high-level nuclear waste repository excavated in hard rock. However, recent experimental and modeling results indicate that the processes of time-dependent microcracking and water-induced stress corrosion can lead to significant reductions in strength and alteration of other key rock properties in the near-field region of a repository. We review the small data base supporting these conclusions and stress the need for an extensive laboratory program to obtain the new data that will be required for design of a repository.

  9. Geological disposal of high-level radioactive waste. Conceptual repository design in hard rock

    International Nuclear Information System (INIS)

    Beale, H.; Griffin, J.R.; Davies, J.W.; Burton, W.R.

    1980-01-01

    The paper gives an interim report on UK studies on possible designs for a repository for vitrified high-level radioactive waste in crystalline rock. The properties of the waste are described and general technical considerations of consequences of disposal in the rock. As an illustration, two basic designs are described associated with pre-cooling in an intermediate store. Firstly, a 'wet repository' is outlined wherein canisters are sealed up closely in boreholes in the rock in regions of low groundwater movement. Secondly, a 'dry repository' above sea level is described where emplacement in tunnels is followed by a loose backfill containing activity absorbers. A connection to deep permeable strata maintains water levels below emplacement positions. Variants on the two basic schemes (tunnel emplacement in a wet repository and in situ cooling) are also assessed. It is concluded that all designs discussed produce a size of repository feasible for construction in the UK. Further, (1) a working figure of 100 0 C per maximum rock temperature is not exceeded, (2) no insuperable engineering problems have so far been found, though rock mechanics studies are at an early stage; (3) it is not possible to discount the escape of a few long-lived 'man-made' isotopes. A minute increment to natural activity in the biosphere may occur from traces of uranium and its decay chains; (4) at this stage, all the designs are still possible candidates for the construction of a UK repository. (author)

  10. Information base for waste repository design. Volume 3. Waste/rock interactions

    International Nuclear Information System (INIS)

    Koplick, C.M.; Pentz, D.L.; Oston, S.G.; Talbot, R.

    1979-01-01

    This report describes the important effects resulting from interaction between radioactive waste and the rock in a nuclear waste repository. The state of the art in predicting waste/rock interactions is summarized. Where possible, independent numerical calculations have been performed. Recommendations are made pointing out areas which require additional research

  11. Fluid geochemistry associated associated to rocks: preliminary tests om minerals of granite rocks potentially hostess of radioactive waste repository

    International Nuclear Information System (INIS)

    Amorim, Lucas E.D.; Rios, Francisco J.; Oliveira, Lucilia A.R. de; Alves, James V.; Fuzikawa, Kazuo; Garcia, Luiz; Ribeiro, Yuri; Matos, Evandro C. de

    2009-01-01

    Fluid inclusions (FI) are micro cavities present on crystals and imprison the mineralizer fluids, and are formed during or posterior to the mineral formation. Those kind of studies are very important for orientation of the engineer barrier projects for this purpose, in order to avoid that the solutions present in the mineral FI can affect the repository walls. This work proposes the development of FI micro compositional studies in the the hostess minerals viewing the contribution for a better understanding of the solution composition present in the metamorphosis granitoid rocks. So, micro thermometric, microchemical and characterization of the material confined in the FI, and the hostess minerals. Great part of the found FI are present in the quartz and plagioclase crystals. The obtained data on the mineral compositions and their inclusions will allow to formulate hypothesis on the process which could occurs at the repository walls, decurrens from of the corrosive character (or not) of the fluids present in the FI, and propose measurements to avoid them

  12. Appraisal of hard rock for potential underground repositories of radioactive wastes

    International Nuclear Information System (INIS)

    Cook, N.G.W.

    1977-10-01

    The mechanical safety and stability of such an underground repository depends largely on the virgin state of stress in the rock, groundwater pressures, the strengths of the rocks, heating by the decay of the radioactive wastes, and the layout of the excavations and the disposition of waste cannisters within them. A large body of pertinent data exists in the literature, and each of these factors has been analyzed in the light of this information. The results indicate that there are no fundamental geological nor mechanical reasons why repositories capable of storing radioactive wastes should not be excavated at suitable sites in hard rock. However, specific tests to determine the mechanical and thermal properties of the rocks at a site would be needed to provide the data for the engineering design of a repository. Also, little experience exists of the effects on underground excavations of thermal loads, so that this aspect requires theoretical study and experimental validation. The depths of these potential repositories would lie in the range from 0.5 to 2.0 km below surface, depending upon the strength of the rock. Virgin states of stress have been measured at such depths which would retard the ingress of groundwater and obviate the incidence of faulting. A typical repository comprising three horizons each with a total area of 5 km 2 would have the capacity to store wastes with thermal output of 240 MW

  13. Nuclear waste. DOE has terminated research evaluating crystalline rock for a repository

    International Nuclear Information System (INIS)

    Fultz, Keith O.; Sprague, John W.; Weigel, Dwayne E.; Price, Vincent P.

    1989-05-01

    We found that DOE terminated funding of research projects specifically designed to evaluate the suitability of crystalline rock for a repository. DOE continued other research efforts involving crystalline rock because they will provide information that it considers useful for evaluating the suitability of Yucca Mountain, Nevada, for a potential repository. Such research activities are not prohibited by the amendments. In January 1988, DOE began evaluating both its domestic and international research programs to ensure their compliance with the 1987 amendments. Several DOE offices and contractors were involved in the evaluation. DOE officials believe that the evaluation effectively brought the Office of Civilian Radioactive Waste Management activities into compliance with the amendments while maintaining useful international relations of continuing benefit to the nuclear waste program in general and to DOE's investigation of the Yucca Mountain site in particular. (The 1987 amendments designated Yucca Mountain as the only site that DOE is to investigate for a potential repository.) The approach and results of DOE's evaluation are discussed. Our review of DOE documents indicates that, by June 22, 1988, DOE completed its evaluation of ongoing crystalline rock research projects to ensure compliance with the 1987 amendments, terminated those research activities it identified as being specifically designed to evaluate the suitability of crystalline rock for a repository, continued some research activities involving crystalline rock because these activities would benefit the investigation and development of the Yucca Mountain repository site, and redirected some research activities so that they would contribute to investigating and developing the Yucca Mountain site

  14. Iron-nickel alloys as canister material for radioactive waste disposal in underground repositories

    International Nuclear Information System (INIS)

    Apps, J.A.

    1982-01-01

    Canisters containing high-level radioactive waste must retain their integrity in an underground waste repository for at least one thousand years after burial (Nuclear Regulatory Commission, 1981). Since no direct means of verifying canister integrity is plausible over such a long period, indirect methods must be chosen. A persuasive approach is to examine the natural environment and find a suitable material which is thermodynamically compatible with the host rock under the environmental conditions with the host rock under the environmental conditions expected in a waste repository. Several candidates have been proposed, among them being iron-nickel alloys that are known to occur naturally in altered ultramafic rocks. The following review of stability relations among iron-nickel alloys below 350 0 C is the initial phase of a more detailed evaluation of these alloys as suitable canister materials

  15. Review of important rock mechanics studies required for underground high level nuclear waste repository program

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, S.; Cho, W. J

    2007-01-15

    Disposal concept adapting room and pillar method, which is a confirmed technique in mining and tunnel construction for long time, has advantages at cost, safety, technical feasibility, flexibility, and international cooperation point of views. Then the important rock mechanics principals and in situ and laboratory tests for understanding the behavior of rock, buffer, and backfill as well as their interactions will be reviewed. The accurate understanding of them is important for developing a safe disposal concept and successful operation of underground repository for permanent disposal of radioactive wastes. First of all, In this study, current status of rock mechanics studies for HLW disposal in foreign countries such as Sweden, USA, Canada, Finland, Japan, and France were reviewed. After then the in situ and laboratory tests for site characterization were summarized. Furthermore, rock mechanics studies required during the whole procedure for the disposal project from repository design to the final closure will be reviewed systematically. This study will help for developing a disposal system including site selection, repository design, operation, maintenance, and closure of a repository in deep underground rock. By introducing the required rock mechanics tests at different stages, it would be helpful from the planning stage to the operation stage of a radioactive waste disposal project.

  16. Review of important rock mechanics studies required for underground high level nuclear waste repository program

    International Nuclear Information System (INIS)

    Kwon, S.; Cho, W. J.

    2007-01-01

    Disposal concept adapting room and pillar method, which is a confirmed technique in mining and tunnel construction for long time, has advantages at cost, safety, technical feasibility, flexibility, and international cooperation point of views. Then the important rock mechanics principals and in situ and laboratory tests for understanding the behavior of rock, buffer, and backfill as well as their interactions will be reviewed. The accurate understanding of them is important for developing a safe disposal concept and successful operation of underground repository for permanent disposal of radioactive wastes. First of all, In this study, current status of rock mechanics studies for HLW disposal in foreign countries such as Sweden, USA, Canada, Finland, Japan, and France were reviewed. After then the in situ and laboratory tests for site characterization were summarized. Furthermore, rock mechanics studies required during the whole procedure for the disposal project from repository design to the final closure will be reviewed systematically. This study will help for developing a disposal system including site selection, repository design, operation, maintenance, and closure of a repository in deep underground rock. By introducing the required rock mechanics tests at different stages, it would be helpful from the planning stage to the operation stage of a radioactive waste disposal project

  17. Workshop on rock mechanics issues in repository design and performance assessment

    International Nuclear Information System (INIS)

    1996-04-01

    The Center for Nuclear Waste Regulatory Analyses organized and hosted a workshop on ''Rock Mechanics Issues in Repository Design and Performance Assessment'' on behalf its sponsor the U.S. Nuclear Regulatory Commission (NRC). This workshop was held on September 19- 20, 1994 at the Holiday Inn Crowne Plaza, Rockville, Maryland. The objectives of the workshop were to stimulate exchange of technical information among parties actively investigating rock mechanics issues relevant to the proposed high-level waste repository at Yucca Mountain and identify/confirm rock mechanics issues important to repository design and performance assessment The workshop contained three technical sessions and two panel discussions. The participants included technical and research staffs representing the NRC and the Department of Energy and their contractors, as well as researchers from the academic, commercial, and international technical communities. These proceedings include most of the technical papers presented in the technical sessions and the transcripts for the two panel discussions

  18. Surrounding rock stress analysis of underground high level waste repository

    International Nuclear Information System (INIS)

    Liu Wengang; Wang Ju; Wang Guangdi

    2006-01-01

    During decay of nuclear waste, enormous energy was released, which results in temperature change of surrounding rock of depository. Thermal stress was produced because thermal expansion of rock was controlled. Internal structure of surrounding rock was damaged and strength of rock was weakened. So, variation of stress was a dynamic process with the variation of temperature. BeiShan region of Gansu province was determined to be the depository field in the future, it is essential to make research on granite in this region. In the process of experiment, basic physical parameters of granite were analyzed preliminary with MTS. Long range temperature and stress filed was simulated considering the damage effect of surrounding rock, and rules of temperature and stress was achieved. (authors)

  19. Evaluation of radiological safety assessment of a repository in a clay rock formation

    International Nuclear Information System (INIS)

    1999-01-01

    This report presents a comprehensive description of the post-closure radiological safety assessment of a repository for the spent fuel arisings resulting from the Spanish nuclear program excavated in a clay host rock formation. In this report three scenarios have been analysed in detail. The first scenario represents the normal in detail. The first scenario represents the normal evolution of the repository (Reference Scenario); and includes a set of variants to investigate the relative importance of the various repository components and examine the sensitivity of the performance to parameters variations. Two altered scenarios have also been considered: deep well construction and poor sealing of the repository. This document contains a detailed description of the repository system, the methodology adopted for the scenarios generation, the process modelling approach and the results of the consequences analysis. (Author)

  20. Selection of Corrosion Resistant Materials for Nuclear Waste Repositories

    International Nuclear Information System (INIS)

    R.B. Rebak

    2006-01-01

    Several countries are considering geological repositories to dispose of nuclear waste. The environment of most of the currently considered repositories will be reducing in nature, except for the repository in the US, which is going to be oxidizing. For the reducing repositories, alloys such as carbon steel, stainless steels and titanium are being evaluated. For the repository in the US, some of the most corrosion resistant commercially available alloys are being investigated. This paper presents a summary of the behavior of the different materials under consideration for the repositories and the current understanding of the degradation modes of the proposed alloys in ground water environments from the point of view of general corrosion, localized corrosion and environmentally assisted cracking

  1. Estimated quantities of residual materials in a KBS-3H repository at Olkiluoto

    International Nuclear Information System (INIS)

    Hagros, Annika

    2008-12-01

    The quantities of residual materials in a KBS-3H type repository have been estimated in this report. The repository is assumed to be constructed at Olkiluoto in Eurajoki, Western Finland. Both the total quantities of the materials introduced into the repository and the quantities of materials that remain in the repository after closure have been calculated. The calculations are largely based on a similar work regarding the material quantities in the Finnish KBS-3V repository and the main goal has been to identify the differences between the KBS-3H and KBS-3V repositories with respect to the type and quantities of residual materials. As the design of the KBS-3H repository is not final yet, the results are only preliminary. Several alternative designs were assumed in the calculations, resulting in different total quantities of materials. The design alternatives that had the greatest effect on the total material quantities were the two different tunnel backfill options, bentonite-crushed rock and Friedland clay. If Friedland clay is used instead of a bentonite-crushed rock mixture, the total quantity of pyrite remaining in the repository is 20 times larger and the quantities of organic materials and gypsum are also increased significantly. The other design alternatives did not have a substantial effect on the total material quantities. The remaining quantity of cement can be reduced by some 20% by selecting the silica grouting alternative in the sealing of the rock mass and low-pH cement in the shotcreting of the repository, instead of using the ordinary cement alternatives. If the total quantity of steel should be minimised, the use of the DAWE design alternative would be better than the Basic Design, although the total reduction would be less than 10%. The main difference between the different drift end plug alternatives is related to the total remaining quantity of silica, which is some 80% smaller if the rock plug is used instead of the LHHP (Low Heat High

  2. Estimated quantities of residual materials in a KBS-3H repository at Olkiluoto

    Energy Technology Data Exchange (ETDEWEB)

    Hagros, Annika (Sannio and Riekkola OY (Finland))

    2008-12-15

    The quantities of residual materials in a KBS-3H type repository have been estimated in this report. The repository is assumed to be constructed at Olkiluoto in Eurajoki, Western Finland. Both the total quantities of the materials introduced into the repository and the quantities of materials that remain in the repository after closure have been calculated. The calculations are largely based on a similar work regarding the material quantities in the Finnish KBS-3V repository and the main goal has been to identify the differences between the KBS-3H and KBS-3V repositories with respect to the type and quantities of residual materials. As the design of the KBS-3H repository is not final yet, the results are only preliminary. Several alternative designs were assumed in the calculations, resulting in different total quantities of materials. The design alternatives that had the greatest effect on the total material quantities were the two different tunnel backfill options, bentonite-crushed rock and Friedland clay. If Friedland clay is used instead of a bentonite-crushed rock mixture, the total quantity of pyrite remaining in the repository is 20 times larger and the quantities of organic materials and gypsum are also increased significantly. The other design alternatives did not have a substantial effect on the total material quantities. The remaining quantity of cement can be reduced by some 20% by selecting the silica grouting alternative in the sealing of the rock mass and low-pH cement in the shotcreting of the repository, instead of using the ordinary cement alternatives. If the total quantity of steel should be minimised, the use of the DAWE design alternative would be better than the Basic Design, although the total reduction would be less than 10%. The main difference between the different drift end plug alternatives is related to the total remaining quantity of silica, which is some 80% smaller if the rock plug is used instead of the LHHP (Low Heat High

  3. Geotechnical materials considerations for conceptual repository design in the Palo Duro Basin, Texas

    International Nuclear Information System (INIS)

    Versluis, W.S.; Balderman, M.A.

    1984-01-01

    The Palo Duro Basin is only one of numerous potential repository locations for placement of a nuclear waste repository. Conceptual designs in the Palo Duro Basin involve considerations of the character and properties of the geologic materials found on several sites throughout the Basin. The first consideration presented includes current basin exploration results and interpretations of engineering properties for the basin geologic sequences. The next consideration presented includes identification of the characteristics of rock taken from the geologic sequence of interest through laboratory and field testing. Values for materials properties of representative samples are obtained for input into modeling of the material response to repository placement. Conceptual designs which respond to these geotechnical considerations are discussed. 4 references, 4 figures, 4 tables

  4. Coupled modelling of convergence, steel corrosion, gas production and brine flow in a rock salt repository

    International Nuclear Information System (INIS)

    Becker, D.A.; Hirsekorn, R.P.

    2013-01-01

    This poster presents the global simulation of the behaviour of thick-walled steel containers piled up in a borehole in a rock salt repository. The simulation takes into account: the convergence by the creeping of rock salt, the backfill and waste compaction, the porosity dependent flow resistance, the anaerobic corrosion (iron to magnetite transformation, gas production, brine consumption, water consumption and salt precipitation) and pressure development. Mechanical influence of corrosion has not yet been taken into account in the integrated code LOPOS

  5. Selection and durability of seal materials for a bedded salt repository: preliminary studies

    International Nuclear Information System (INIS)

    Roy, D.M.; Grutzeck, M.W.; Wakeley, L.D.

    1983-11-01

    This report details preliminary results of both experimental and theoretical studies of cementitious seal materials for use in a proposed nuclear waste repository in bedded salt. Effects of changes in bulk composition and environment upon phase stability and physical/mechanical properties have been evaluated for more than 25 formulations. Bonding and interfacial characteristics of the region between host rock and seal material or concrete aggregate and cementitious matrix for selected formulations have been studied. Compatibilities of clays and zeolites in brines typical of the SE New Mexico region have been investigated, and their stabilities reviewed. Results of these studies have led to the conclusion that cementitious materials can be formulated which are compatible with the major rock types in a bedded salt repository environment. Strengths are more than adequate, permeabilities are consistently very low, and elastic moduli generally increase only very slightly with time. Seal formulation guidelines and recommendations for present and future work are presented. 73 references, 25 figures, 61 tables

  6. Appraisal of hard rock for potential underground repositories of radioactive wastes. LBL-7004

    International Nuclear Information System (INIS)

    Cook, N.G.W.

    1978-01-01

    Underground burial of radioactive wastes in hard rock may be an effective and safe means of isolating them from the environment and from man. The mechanical safety and stability of such an underground repository depends largely on the virgin state of stress in the rock, groundwater pressures, the strengths of the rocks, heating by the decay of the radioactive wastes, and the layout of the excavations and the disposition of waste cannisters within them. A large body of pertinent data exists in the literature, and each of these factors has been analyzed in the light of this information. The results indicate that there are no fundamental geological nor mechanical reasons why repositories capable of storing radioactive wastes should not be excavated at suitable sites in hard rock. However, specific tests to determine the mechanical and thermal properties of the rocks at a site would be needed to provide the data for the engineering design of a repository. Also, little experience exists of the effects on underground excavations of thermal loads, so that this aspect requires theoretical study and experimental validation. The depths of these potential repositories would lie in the range from 0.5 km to 2.0 km below surface, depending upon the strength of the rock. Virgin states of stress have been measured at such depths which would retard the ingress of groundwater and obviate the incidence of faulting. A typical repository comprising three horizons each with a total area of 5 km 2 would have the capacity to store wastes with thermal output of 240 MW

  7. On-line repository of audiovisual material feminist research methodology

    Directory of Open Access Journals (Sweden)

    Lena Prado

    2014-12-01

    Full Text Available This paper includes a collection of audiovisual material available in the repository of the Interdisciplinary Seminar of Feminist Research Methodology SIMReF (http://www.simref.net.

  8. Aespoe Hard Rock Laboratory. Prototype Repository. Sensors data report (Period 010917-091201) Report No: 22

    International Nuclear Information System (INIS)

    Goudarzi, Reza; Johannesson, Lars-Erik

    2009-12-01

    The Prototype Repository Test consists of two sections. The installation of the first Section of Prototype Repository was made during summer and autumn 2001 and Section 2 was installed in spring and summer 2003. This report presents data from measurements in the Prototype Repository during the period 010917-091201. The report is organized so that the actual measured results are shown in Appendix 1-10, where Appendix 8 deals with measurements of canister displacements (by AITEMIN), Appendix 9 deals with geo-electric measurements in the backfill (by GRS), Appendix 10 deals with stress and strain measurement in the rock (by AaF) and Appendix 11 deals with measurement of water pressure in the rock (by VBB/VIAK). The main report and Appendix 1-7 deal with the rest of the measurements

  9. Aespoe Hard Rock Laboratory. Prototype Repository. Sensors data report (Period 010917-090601) Report No: 21

    International Nuclear Information System (INIS)

    Goudarzi, Reza; Johannesson, Lars-Erik

    2009-07-01

    The Prototype Repository Test consists of two sections. The installation of the first Section of Prototype Repository was made during summer and autumn 2001 and Section 2 was installed in spring and summer 2003. This report presents data from measurements in the Prototype Repository during the period 010917-090601. The report is organized so that the actual measured results are shown in Appendix 1-10, where Appendix 8 deals with measurements of canister displacements (by AITEMIN), Appendix 9 deals with geo-electric measurements in the backfill (by GRS), Appendix 10 deals with stress and strain measurement in the rock (by BBK) and Appendix 11 deals with measurement of water pressure in the rock (by VBB/VIAK). The main report and Appendix 1-7 deal with the rest of the measurements

  10. Aespoe Hard Rock Laboratory. Prototype Repository. Sensors data report (Period 010917-081201) Report No: 20

    International Nuclear Information System (INIS)

    Goudarzi, Reza; Johannesson, Lars-Erik

    2009-03-01

    The Prototype Repository Test consists of two sections. The installation of the first Section of Prototype Repository was made during summer and autumn 2001 and Section 2 was installed in spring and summer 2003. This report presents data from measurements in the Prototype Repository during the period 010917-081201. The report is organized so that the actual measured results are shown in Appendix 1-10, where Appendix 8 deals with measurements of canister displacements (by AITEMIN), Appendix 9 deals with geo-electric measurements in the backfill (by GRS), Appendix 10 deals with stress and strain measurement in the rock (by BBK) and Appendix 11 deals with measurement of water pressure in the rock (by VBB/VIAK). The main report and Appendix 1-7 deal with the rest of the measurements

  11. Aespoe Hard Rock Laboratory. Prototype Repository. Sensors data report (Period 010917-090601) Report No: 21

    Energy Technology Data Exchange (ETDEWEB)

    Goudarzi, Reza; Johannesson, Lars-Erik (Clay Technology AB, Lund (Sweden))

    2009-07-15

    The Prototype Repository Test consists of two sections. The installation of the first Section of Prototype Repository was made during summer and autumn 2001 and Section 2 was installed in spring and summer 2003. This report presents data from measurements in the Prototype Repository during the period 010917-090601. The report is organized so that the actual measured results are shown in Appendix 1-10, where Appendix 8 deals with measurements of canister displacements (by AITEMIN), Appendix 9 deals with geo-electric measurements in the backfill (by GRS), Appendix 10 deals with stress and strain measurement in the rock (by BBK) and Appendix 11 deals with measurement of water pressure in the rock (by VBB/VIAK). The main report and Appendix 1-7 deal with the rest of the measurements.

  12. Aespoe Hard Rock Laboratory. Prototype Repository. Sensors data report (Period 010917-081201) Report No: 20

    Energy Technology Data Exchange (ETDEWEB)

    Goudarzi, Reza; Johannesson, Lars-Erik (Clay Technology AB, Lund (Sweden))

    2009-03-15

    The Prototype Repository Test consists of two sections. The installation of the first Section of Prototype Repository was made during summer and autumn 2001 and Section 2 was installed in spring and summer 2003. This report presents data from measurements in the Prototype Repository during the period 010917-081201. The report is organized so that the actual measured results are shown in Appendix 1-10, where Appendix 8 deals with measurements of canister displacements (by AITEMIN), Appendix 9 deals with geo-electric measurements in the backfill (by GRS), Appendix 10 deals with stress and strain measurement in the rock (by BBK) and Appendix 11 deals with measurement of water pressure in the rock (by VBB/VIAK). The main report and Appendix 1-7 deal with the rest of the measurements.

  13. Aespoe Hard Rock Laboratory. Prototype Repository. Sensors data report (Period 010917-091201) Report No: 22

    Energy Technology Data Exchange (ETDEWEB)

    Goudarzi, Reza; Johannesson, Lars-Erik (Clay Technology AB, Lund (Sweden))

    2009-12-15

    The Prototype Repository Test consists of two sections. The installation of the first Section of Prototype Repository was made during summer and autumn 2001 and Section 2 was installed in spring and summer 2003. This report presents data from measurements in the Prototype Repository during the period 010917-091201. The report is organized so that the actual measured results are shown in Appendix 1-10, where Appendix 8 deals with measurements of canister displacements (by AITEMIN), Appendix 9 deals with geo-electric measurements in the backfill (by GRS), Appendix 10 deals with stress and strain measurement in the rock (by AaF) and Appendix 11 deals with measurement of water pressure in the rock (by VBB/VIAK). The main report and Appendix 1-7 deal with the rest of the measurements.

  14. Integrating rock mechanics issues with repository design through design process principles and methodology

    International Nuclear Information System (INIS)

    Bieniawski, Z.T.

    1996-01-01

    A good designer needs not only knowledge for designing (technical know-how that is used to generate alternative design solutions) but also must have knowledge about designing (appropriate principles and systematic methodology to follow). Concepts such as open-quotes design for manufactureclose quotes or open-quotes concurrent engineeringclose quotes are widely used in the industry. In the field of rock engineering, only limited attention has been paid to the design process because design of structures in rock masses presents unique challenges to the designers as a result of the uncertainties inherent in characterization of geologic media. However, a stage has now been reached where we are be able to sufficiently characterize rock masses for engineering purposes and identify the rock mechanics issues involved but are still lacking engineering design principles and methodology to maximize our design performance. This paper discusses the principles and methodology of the engineering design process directed to integrating site characterization activities with design, construction and performance of an underground repository. Using the latest information from the Yucca Mountain Project on geology, rock mechanics and starter tunnel design, the current lack of integration is pointed out and it is shown how rock mechanics issues can be effectively interwoven with repository design through a systematic design process methodology leading to improved repository performance. In essence, the design process is seen as the use of design principles within an integrating design methodology, leading to innovative problem solving. In particular, a new concept of open-quotes Design for Constructibility and Performanceclose quotes is introduced. This is discussed with respect to ten rock mechanics issues identified for repository design and performance

  15. Study on the locational criteria for submarine rock repositories of low and medium level radioactive wastes

    Energy Technology Data Exchange (ETDEWEB)

    Kim, G H; Kang, W J; Kim, T J. and others [Chungnam National Univ., Taejon (Korea, Republic of)

    1992-01-15

    Submarine repositories have significant advantages over their land counterparts locating close to the areas of daily human activities. Consequently, the construction of submarine repositories on the vast continental shelves around Korean seas is considered to be highly positive. In this context, the development of locational criteria primarily targeting the safety of submarine rock repositories is very important.The contents of the present study are: analyzing characteristics of marine environment: Search of potential hazards to, and environmental impact by, the submarine repositories; Investigation of the oceanographic, geochemical, ecological and sedimentological characteristics of estuaries and coastal seas. Locating potential hazards to submarine repositories by: Bibliographical search of accidents leading to the destruction of submarine structures by turbidity currents and other potentials; Review of turbidity currents. Consideration of environmental impact caused by submarine repositories: Logistics to minimize the environmental impacts in site selection; Removal and dispersion processes of radionuclides in sea water. Analyses of oceanographical characteristics of, and hazard potentials in, the Korean seas. Evaluation of the MOST 91-7 criteria for applicability to submarine repositories and the subsequent proposition of additional criteria.

  16. Study on the locational criteria for submarine rock repositories of low and medium level radioactive wastes

    International Nuclear Information System (INIS)

    Kim, G. H.; Kang, W. J.; Kim, T. J. and others

    1992-01-01

    Submarine repositories have significant advantages over their land counterparts locating close to the areas of daily human activities. Consequently, the construction of submarine repositories on the vast continental shelves around Korean seas is considered to be highly positive. In this context, the development of locational criteria primarily targeting the safety of submarine rock repositories is very important.The contents of the present study are: analyzing characteristics of marine environment: Search of potential hazards to, and environmental impact by, the submarine repositories; Investigation of the oceanographic, geochemical, ecological and sedimentological characteristics of estuaries and coastal seas. Locating potential hazards to submarine repositories by: Bibliographical search of accidents leading to the destruction of submarine structures by turbidity currents and other potentials; Review of turbidity currents. Consideration of environmental impact caused by submarine repositories: Logistics to minimize the environmental impacts in site selection; Removal and dispersion processes of radionuclides in sea water. Analyses of oceanographical characteristics of, and hazard potentials in, the Korean seas. Evaluation of the MOST 91-7 criteria for applicability to submarine repositories and the subsequent proposition of additional criteria

  17. Effects of gaseous radioactive nuclides on the design and operation of repositories for spent LWR fuel in rock salt

    International Nuclear Information System (INIS)

    Jenks, G.H.

    1979-12-01

    Information relating to the identities and amounts of gaseous radionuclides present in spent LWR fuel and to their release from canistered spent fuel under plausible storage and disposal conditions was assembled, reviewed, and analyzed. Information was also reviewed and analyzed on several other subjects that relate to the integrity of the carbon steel canister in which the spent fuel is to be encapsulated and to the expected rates of transfer of gaseous radionuclides through crushed salt backfill within a disposal room in a reference repository in rock salt. The advantages and disadvantages were considered for several different canister-backfill materials, and recommendations were made regarding preferred materials. Other recommendations relate to encapsulation procedures and specifications and to needs for additional experimental studies. The objective of this work was to provide reference information, conclusions, and recommendations that could be used to establish design and operating conditions and procedures for a bedded salt repository for spent LWR fuel and that could also be used to help evaluate the safety of the repository. The results of this work will also generally apply to spent fuel repositories in domal salt. However, because the domal salt may have little or no brine inclusions within it, there may be little or no possibility that brine will migrate into open spaces around an emplaced canister. Addordingly, some of the concerns that result from the possible occurrence of brine migration in bedded salt may be of no importance in domal salt

  18. Panel discussion on rock mechanics issues in repository design

    International Nuclear Information System (INIS)

    Bieniawski, Z.T.; Kim, K.S.; Nataraja, M.

    1996-01-01

    The panel discussion was introduced by Mr. Z.T.(Richard) Bieniawski and then continued with five additional speakers. The topics covered in the discussion included rock mechanics pertaining to the design of underground facilities for the disposal of radioactive wastes and the safety of such facilities. The speakers included: Mr. Kun-Soo Kim who is a specialist in the area of rock mechanics testing during the Basalt Waste Isolation Project; Dr. Mysore Nataraja who is the senior project manager with the NRC; Dr. Michael Voegele who is the project manager for Science Applications International Corporation (SAIC) on the Yucca Mountain Project; Dr. Edward Cording who is a member of the Nuclear Waste Technical Review Board; and Dr. Hemendra Kalia who is employed by Los Alamos National Laboratory and coordinates various activities of testing programs at the Yucca Mountain Site

  19. An innovative 3-D numerical modelling procedure for simulating repository-scale excavations in rock - SAFETI

    Energy Technology Data Exchange (ETDEWEB)

    Young, R. P.; Collins, D.; Hazzard, J.; Heath, A. [Department of Earth Sciences, Liverpool University, 4 Brownlow street, UK-0 L69 3GP Liverpool (United Kingdom); Pettitt, W.; Baker, C. [Applied Seismology Consultants LTD, 10 Belmont, Shropshire, UK-S41 ITE Shrewsbury (United Kingdom); Billaux, D.; Cundall, P.; Potyondy, D.; Dedecker, F. [Itasca Consultants S.A., Centre Scientifique A. Moiroux, 64, chemin des Mouilles, F69130 Ecully (France); Svemar, C. [Svensk Karnbranslemantering AB, SKB, Aspo Hard Rock Laboratory, PL 300, S-57295 Figeholm (Sweden); Lebon, P. [ANDRA, Parc de la Croix Blanche, 7, rue Jean Monnet, F-92298 Chatenay-Malabry (France)

    2004-07-01

    This paper presents current results from work performed within the European Commission project SAFETI. The main objective of SAFETI is to develop and test an innovative 3D numerical modelling procedure that will enable the 3-D simulation of nuclear waste repositories in rock. The modelling code is called AC/DC (Adaptive Continuum/ Dis-Continuum) and is partially based on Itasca Consulting Group's Particle Flow Code (PFC). Results are presented from the laboratory validation study where algorithms and procedures have been developed and tested to allow accurate 'Models for Rock' to be produced. Preliminary results are also presented on the use of AC/DC with parallel processors and adaptive logic. During the final year of the project a detailed model of the Prototype Repository Experiment at SKB's Hard Rock Laboratory will be produced using up to 128 processors on the parallel super computing facility at Liverpool University. (authors)

  20. Backfilling techniques and materials in underground excavations: Potential alternative backfill materials in use in Posiva's spent fuel repository concept

    International Nuclear Information System (INIS)

    Dixon, D.A.; Keto, P.

    2009-05-01

    A variety of geologic media options have been proposed as being suitable for safely and permanently disposing of spent nuclear fuel or fuel reprocessing wastes. In Finland the concept selected is construction of a deep repository in crystalline rock (Posiva 1999, 2006; SKB 1999), likely at the Olkiluoto site (Posiva 2006). Should that site prove suitable, excavation of tunnels and several vertical shafts will be necessary. These excavations will need to be backfilled and sealed as emplacement operations are completed and eventually all of the openings will need to be backfilled and sealed. Clay-based materials were selected after extensive review of materials options and the potential for practical implementation in a repository and work over a 30+ year period has led to the development of a number of workable clay-based backfilling options, although discussion persists as to the most suitable clay materials and placement technologies to use. As part of the continuous process of re-evaluating backfilling options in order to provide the best options possible, placement methods and materials that have been given less attention have been revisited. Primary among options that were and continue to be evaluated as a potential backfill are cementitious materials. These materials were included in the list of candidate materials initially screened in the late 1970's for use in repository backfilling. Conventional cement-based materials were quickly identified as having some serious technical limitations with respect their ability to fulfil the identified requirements of backfill. Concerns related to their ability to achieve the performance criteria defined for backfill resulted in their exclusion from large-scale use as backfill in a repository. Development of new, less chemically aggressive cementitious materials and installation technologies has resulted in their re-evaluation. Concrete and cementitious materials have and are being developed that have chemical, durability

  1. Creep in crystalline rock with application to high level nuclear waste repository

    International Nuclear Information System (INIS)

    Eloranta, P.; Simonen, A.

    1992-06-01

    The time-dependent strength and deformation properties of hard crystalline rock are studied. Theoretical models defining the phenomena which can effect these properties are reviewed. The time- dependent deformation of the openings in the proposed nuclear waste repository is analysed. The most important factors affecting the subcritical crack growth in crystalline rock are the stress state, the chemical environment, temperature and microstructure of the rock. There are several theoretical models for the analysis of creep and cyclic fatigue: deformation diagrams, rheological models thermodynamic models, reaction rate models, stochastic models, damage models and time-dependent safety factor model. They are defective in describing the three-axial stress condition and strength criteria. In addition, the required parameters are often too difficult to determine with adequate accuracy. Therefore these models are seldom applied in practice. The effect of microcrack- driven creep on the stability of the work shaft, the emplacement tunnel and the capsulation hole of a proposed nuclear waste repository was studied using a numerical model developed by Atomic Energy of Canada Ltd. According to the model, the microcrack driven creep progresses very slowly in good quality rock. Poor rock quality may accelerate the creep rate. The model is very sensitive to the properties of the rock and secondary stress state. The results show that creep causes no stability problems on excavations in good rock. The results overestimate the effect of the creep, because the analysis omitted the effect of support structures and backfilling

  2. Retrievability of high-level nuclear waste from geologic repositories - Regulatory and rock mechanics/design considerations

    International Nuclear Information System (INIS)

    Tanious, N.S.; Nataraja, M.S.; Daemen, J.J.K.

    1987-01-01

    Retrievability of nuclear waste from high-level geologic repositories is one of the performance objectives identified in 10CFR60 (Code of Federal Regulations, 1985). 10CFR60.111 states that the geologic repository operations area shall be designed to preserve the option of waste retrieval. In designing the repository operations area, rock mechanics considerations play a major role especially in evaluating the feasibility of retrieval operations. This paper discusses generic considerations affecting retrievability as they relate to repository design, construction, and operation, with emphasis on regulatory and rock mechanics aspects

  3. Numerical modeling of the geomechanical response of a rock mass to a radioactive waste repository

    International Nuclear Information System (INIS)

    Hardy, M.P.; St John, C.M.; Hocking, G.

    1979-06-01

    Geotechnical numerical models capable of predicting the thermomechanical response and groundwater movements around an underground radioactive waste repository are vital to the success of the nuclear waste disposal program. In the absence of directly related engineering experience, the design, risk assessment, and licensing procedures of a repository will be reliant on predictions made using such models. This paper reviews models being used to assist in repository design and summarizes the results of a recent parametric study of underground disposal in basaltic rocks. On the basis of preliminary site data, it is concluded that the allowable areal density of heat-generating waste will be controlled by the stability of placement rooms and the boreholes in which waste canisters are placed. Regional effects including thermally induced upward groundwater flow, appear to present less severe problems

  4. Postclosure safety assessment of a used fuel repository in sedimentary rock

    International Nuclear Information System (INIS)

    Gobien, M.; Garisto, F.; Hunt, N.; Kremer, E.

    2014-01-01

    The Nuclear Waste Management Organization (NWMO) is responsible for the implementation of Adaptive Phased Management (APM), the federally-approved plan for safe long-term management of Canada's used nuclear fuel. Under the APM plan, used nuclear fuel will ultimately be placed within a deep geological repository in a suitable rock formation. This paper summarizes an illustrative case study of the current multi-barrier design and postclosure safety of a deep geological repository in a hypothetical sedimentary Michigan Basin setting. The purpose of this postclosure safety assessment is to determine potential effects of the repository on the health and safety of persons and the environment. Results are compared against acceptance criteria established for the protection of persons and the environment from potential radiological and non-radiological hazards. (author)

  5. Postclosure safety assessment of a used fuel repository in sedimentary rock

    Energy Technology Data Exchange (ETDEWEB)

    Gobien, M.; Garisto, F.; Hunt, N.; Kremer, E. [Nuclear Waste Management Organization, Toronto, ON (Canada)

    2014-07-01

    The Nuclear Waste Management Organization (NWMO) is responsible for the implementation of Adaptive Phased Management (APM), the federally-approved plan for safe long-term management of Canada's used nuclear fuel. Under the APM plan, used nuclear fuel will ultimately be placed within a deep geological repository in a suitable rock formation. This paper summarizes an illustrative case study of the current multi-barrier design and postclosure safety of a deep geological repository in a hypothetical sedimentary Michigan Basin setting. The purpose of this postclosure safety assessment is to determine potential effects of the repository on the health and safety of persons and the environment. Results are compared against acceptance criteria established for the protection of persons and the environment from potential radiological and non-radiological hazards. (author)

  6. Assessment of the Durability of Cementitious Materials in Repository Environment

    International Nuclear Information System (INIS)

    Vicente, R.; Marumo, J.T.; Miyamoto, H.; Isiki, V.L.K.; Ferreira, E.G.

    2013-01-01

    The Radioactive Waste Management Laboratory of the Energy and Nuclear Research Institute is developing the concept of a borehole repository for disused sealed radioactive sources drilled in a deep granite batholite. In this concept, the annular space between the well steel casing and the geological formation is backfilled with cement paste. The hardened cement paste functions as an additional barrier against the escape of radionuclides from the repository and their migration to the environment. It also functions as an obstacle to the flow of groundwater between different layers of the geological setting crossed by the borehole. The long term behavior of hydrated cement compounds is yet incompletely known and therefore more research is needed to increase the confidence on the performance of the material under the repository conditions as required. For the repository to achieve the required performance, the cement paste must be durable. However, in a deep repository, the cementitious materials is exposed to the deleterious action of high temperatures and pressures, the radiation field created by the radioactive sources and aggressive ion species that may be present in groundwater. Furthermore, it is necessary to consider that the cement paste is unstable in the long term because its microstructure and mineralogy change with time as the cement gel components recrystallize and react chemically with materials of the repository environment. In principle, the lifetime of this material could be determined based on the study of its long-term behavior, which, in turn, could be estimated from the extrapolation of short-term results, by accelerating, under controlled laboratory conditions, the composition changes and the loss of mechanical strength and cohesion induced by any detrimental component of the repository environment. Loss of mechanical strength, dimensional variations, changes in chemical-mineralogical composition, and leaching of hydrate compounds are all possible

  7. Stress, strain, and temperature induced permeability changes in potential repository rocks

    International Nuclear Information System (INIS)

    Heard, H.C.; Duba, A.

    1977-01-01

    Work is in progress to assess the permeability characteristics of coarse-grained igneous rocks as affected by pressure, deviatoric stress, and temperature. In order to predict the long-term behavior of these rocks, both virgin and fractured, permeability and all principal strains resulting from an imposed deviatoric stress under various simulated lithostatic pressures are being measured. In addition, compressional as well as shear velocities and electrical conductivity are being evaluated along these principal directions. These simultaneous measurements are being made initially at 25 0 C on a 15 cm diameter by 30 cm long sample in a pressure apparatus controlled by a mini-computer. Correlation of these data with similar field observations should then allow simplified exploration for a suitable repository site as well as the prediction of the response of a mined cavity with both distance and time at this site. After emplacement of the waste canisters, the mechanical stability and hydrologic integrity of this mined repository will be directly influenced by the fracturing of the surrounding rock which results from local temperature differences and the thermal expansion of that rock. Temperatures (and, hence, these differences) in the vicinity of the repository are expected to be affected by the presence of pore fluids (single- or two-phase) in the rock, the heat capacity and the thermal conductivity of this system. In turn, these are all dependent upon lithostatic pressure, pore pressure, and stress. Thermal expansion (and fracturing) will also be affected by the lithostatic (and effective) pressure, the deviatoric stress field, and the initial anisotropy of the rock

  8. Transport of gaseous C-14 from a repository in unsaturated rock

    International Nuclear Information System (INIS)

    Light, W.B.; Chambre, P.L.; Lee, W. L.; Pigford, T.H.; California Univ., Berkeley, CA

    1990-09-01

    The authors predict the transport of gaseous 14 CO 2 from a nuclear waste repository in unsaturated rock using a porous-medium model. This model is justified if the appropriate modified Peclet number, which indicates equilibrium between gas in fractures and liquid in rock pores, is much less than unity. Numerical illustrations are given which are applicable to the proposed repository at Yucca Mountain which is 350 m underground. Maximum predicted concentrations of 14 CO 2 near the ground surface are comparable to the USNRC limit for unrestricted areas. Maximum predicted dose rates above ground are less than 1% of background. Travel times are predicted to be hundreds to thousands of years. For some cases, it is shown that the release rate from the source has negligible effect on concentrations at the ground surface. 15 refs., 10 figs., 1 tab

  9. Technology needs for selecting and evaluating high-level waste repository sites in crystalline rock

    International Nuclear Information System (INIS)

    1988-12-01

    This report describes properties and processes that govern the performance of the geological barrier in a nuclear waste isolation system in crystalline rock and the state-of-the-art in the understanding of these properties and processes. Areas and topics that require further research and development as well as technology needs for investigating and selecting repository sites are presented. Experiences from the Swedish site selection program are discussed, and a general investigation strategy is presented for an area characterization phase of an exploratory program in crystalline rocks. 255 refs., 65 figs., 10 tabs

  10. Workshop on rock mechanics issues in repository design and performance assessment

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-04-01

    The Center for Nuclear Waste Regulatory Analyses organized and hosted a workshop on ``Rock Mechanics Issues in Repository Design and Performance Assessment`` on behalf its sponsor the U.S. Nuclear Regulatory Commission (NRC). This workshop was held on September 19- 20, 1994 at the Holiday Inn Crowne Plaza, Rockville, Maryland. The objectives of the workshop were to stimulate exchange of technical information among parties actively investigating rock mechanics issues relevant to the proposed high-level waste repository at Yucca Mountain and identify/confirm rock mechanics issues important to repository design and performance assessment The workshop contained three technical sessions and two panel discussions. The participants included technical and research staffs representing the NRC and the Department of Energy and their contractors, as well as researchers from the academic, commercial, and international technical communities. These proceedings include most of the technical papers presented in the technical sessions and the transcripts for the two panel discussions. Selected papers have been indexed separately for inclusion the Energy Science and Technology Database.

  11. Far-field thermomechanical response of argillaceous rock to emplacement of a nuclear-waste repository

    International Nuclear Information System (INIS)

    McVey, D.F.; Thomas, R.K.; Lappin, A.R.

    1980-08-01

    Before heat-producing wastes can be emplaced safely in any argillaceous rock, it will be necessary to understand the far-field thermal and thermomechanical response of this rock to waste emplacement. This report presents the results of a first series of calculations aimed at estimating the far-field response of argillite to waste emplacement. Because the thermal and mechanical properties of argillite are affected by its content of expandable clay, its behavior is briefly compared and contrasted with that of a shale having the same matrix thermal properties, but containing no expandable clay. Under this assumption, modeled temperatures are the same for the two rock types at equivalent power densities and reflect the large dependence of in-situ temperatures on both initial power density and waste type. Thermomechanical calculations indicate that inclusion of contraction behavior of expandable clays in the assumed argillite thermal expansion behavior results, in some cases, in generation of a large zone in and near the repository that has undergone volumetric contraction but is surrounded by uniformly compressive stresses. Information available to date indicates that this contraction would likely result in locally increased fluid permeability and decreased in-situ thermal conductivity, but might well be advantageous as regards radionuclide retention, because of the increased surface area within the contracted zone. Assumption of continuous and positive expansion behavior for the shale eliminates the near-repository contraction and tensional zones, but results in near-surface tensional zones directly above the repository

  12. Damage-plasticity model of the host rock in a nuclear waste repository

    Energy Technology Data Exchange (ETDEWEB)

    Koudelka, Tomáš; Kruis, Jaroslav, E-mail: kruis@fsv.cvut.cz [Department of Mechanics, Faculty of Civil Engineering, Czech Technical University in Prague, Thákurova 7, 166 29 Prague (Czech Republic)

    2016-06-08

    The paper describes damage-plasticity model for the modelling of the host rock environment of a nuclear waste repository. Radioactive Waste Repository Authority in Czech Republic assumes the repository to be in a granite rock mass which exhibit anisotropic behaviour where the strength in tension is lower than in compression. In order to describe this phenomenon, the damage-plasticity model is formulated with the help of the Drucker-Prager yield criterion which can be set to capture the compression behaviour while the tensile stress states is described with the help of scalar isotropic damage model. The concept of damage-plasticity model was implemented in the SIFEL finite element code and consequently, the code was used for the simulation of the Äspö Pillar Stability Experiment (APSE) which was performed in order to determine yielding strength under various conditions in similar granite rocks as in Czech Republic. The results from the performed analysis are presented and discussed in the paper.

  13. LIFE Materials: Fuel Cycle and Repository Volume 11

    Energy Technology Data Exchange (ETDEWEB)

    Shaw, H; Blink, J A

    2008-12-12

    The fusion-fission LIFE engine concept provides a path to a sustainable energy future based on safe, carbon-free nuclear power with minimal nuclear waste. The LIFE design ultimately offers many advantages over current and proposed nuclear energy technologies, and could well lead to a true worldwide nuclear energy renaissance. When compared with existing and other proposed future nuclear reactor designs, the LIFE engine exceeds alternatives in the most important measures of proliferation resistance and waste minimization. The engine needs no refueling during its lifetime. It requires no removal of fuel or fissile material generated in the LIFE engine. It leaves no weapons-attractive material at the end of life. Although there is certainly a need for additional work, all indications are that the 'back end' of the fuel cycle does not to raise any 'showstopper' issues for LIFE. Indeed, the LIFE concept has numerous benefits: (1) Per unit of electricity generated, LIFE engines would generate 20-30 times less waste (in terms of mass of heavy metal) requiring disposal in a HLW repository than does the current once-through fuel cycle. (2) Although there may be advanced fuel cycles that can compete with LIFE's low mass flow of heavy metal, all such systems require reprocessing, with attendant proliferation concerns; LIFE engines can do this without enrichment or reprocessing. Moreover, none of the advanced fuel cycles can match the low transuranic content of LIFE waste. (3) The specific thermal power of LIFE waste is initially higher than that of spent LWR fuel. Nevertheless, this higher thermal load can be managed using appropriate engineering features during an interim storage period, and could be accommodated in a Yucca-Mountain-like repository by appropriate 'staging' of the emplacement of waste packages during the operational period of the repository. The planned ventilation rates for Yucca Mountain would be sufficient for LIFE waste

  14. LIFE Materials: Fuel Cycle and Repository Volume 11

    International Nuclear Information System (INIS)

    Shaw, H.; Blink, J.A.

    2008-01-01

    The fusion-fission LIFE engine concept provides a path to a sustainable energy future based on safe, carbon-free nuclear power with minimal nuclear waste. The LIFE design ultimately offers many advantages over current and proposed nuclear energy technologies, and could well lead to a true worldwide nuclear energy renaissance. When compared with existing and other proposed future nuclear reactor designs, the LIFE engine exceeds alternatives in the most important measures of proliferation resistance and waste minimization. The engine needs no refueling during its lifetime. It requires no removal of fuel or fissile material generated in the LIFE engine. It leaves no weapons-attractive material at the end of life. Although there is certainly a need for additional work, all indications are that the 'back end' of the fuel cycle does not to raise any 'showstopper' issues for LIFE. Indeed, the LIFE concept has numerous benefits: (1) Per unit of electricity generated, LIFE engines would generate 20-30 times less waste (in terms of mass of heavy metal) requiring disposal in a HLW repository than does the current once-through fuel cycle. (2) Although there may be advanced fuel cycles that can compete with LIFE's low mass flow of heavy metal, all such systems require reprocessing, with attendant proliferation concerns; LIFE engines can do this without enrichment or reprocessing. Moreover, none of the advanced fuel cycles can match the low transuranic content of LIFE waste. (3) The specific thermal power of LIFE waste is initially higher than that of spent LWR fuel. Nevertheless, this higher thermal load can be managed using appropriate engineering features during an interim storage period, and could be accommodated in a Yucca-Mountain-like repository by appropriate 'staging' of the emplacement of waste packages during the operational period of the repository. The planned ventilation rates for Yucca Mountain would be sufficient for LIFE waste to meet the thermal constraints of

  15. Reactive transport simulations of the evolution of a cementitious repository in clay-rich host rocks

    Science.gov (United States)

    Kosakowski, Georg; Berner, Urs; Kulik, Dmitrii A.

    2010-05-01

    In Switzerland, the deep geological disposal in clay-rich rocks is foreseen not only for high-level radioactive waste, but also for intermediate-level (ILW) and low-level (LLW) radioactive waste. Typically, ILW and LLW repositories contain huge amounts of cementitious materials used for waste conditioning, confinement, and as backfill for the emplacement caverns. We are investigating the interactions of such a repository with the surrounding clay rocks and with other clay-rich materials such as sand/bentonite mixtures that are foreseen for backfilling the access tunnels. With the help of a numerical reactive transport model, we are comparing the evolution of cement/clay interfaces for different geochemical and transport conditions. In this work, the reactive transport of chemical components is simulated with the multi-component reactive transport code OpenGeoSys-GEM. It employs the sequential non-iterative approach to couple the mass transport code OpenGeoSys (http://www.ufz.de/index.php?en=18345) with the GEMIPM2K (http://gems.web.psi.ch/) code for thermodynamic modeling of aquatic geochemical systems which is using the Gibbs Energy Minimization (GEM) method. Details regarding code development and verification can be found in Shao et al. (2009). The mineral composition and the pore solution of a CEM I 52.5 N HTS hydrated cement as described by Lothenbach & Wieland (2006) are used as an initial state of the cement compartment. The setup is based on the most recent CEMDATA07 thermodynamic database which includes several ideal solid solutions for hydrated cement minerals and is consistent with the Nagra/PSI thermodynamic database 01/01. The smectite/montmorillonite model includes cation exchange processes and amphotheric≡SOH sites and was calibrated on the basis of data by Bradbury & Baeyens (2002). In other reactive transport codes based on the Law of Mass Action (LMA) for solving geochemical equilibria, cation exchange processes are usually calculated assuming

  16. STAFAN, Fluid Flow, Mechanical Stress in Fractured Rock of Nuclear Waste Repository

    International Nuclear Information System (INIS)

    Huyakorn, P.; Golis, M.J.

    1989-01-01

    1 - Description of program or function: STAFAN (Stress And Flow Analysis) is a two-dimensional, finite-element code designed to model fluid flow and the interaction of fluid pressure and mechanical stresses in a fractured rock surrounding a nuclear waste repository. STAFAN considers flow behavior of a deformable fractured system with fracture-porous matrix interactions, the coupling effects of fluid pressure and mechanical stresses in a medium containing discrete joints, and the inelastic response of the individual joints of the rock mass subject to the combined fluid pressure and mechanical loading. 2 - Restrictions on the complexity of the problem: STAFAN does not presently contain thermal coupling, and it is unable to simulate inelastic deformation of the rock mass and variably saturated or two-phase flow in the fractured porous medium system

  17. A probabilistic approach to rock mechanical property characterization for nuclear waste repository design

    International Nuclear Information System (INIS)

    Kim, Kunsoo; Gao, Hang

    1996-01-01

    A probabilistic approach is proposed for the characterization of host rock mechanical properties at the Yucca Mountain site. This approach helps define the probability distribution of rock properties by utilizing extreme value statistics and Monte Carlo simulation. We analyze mechanical property data of tuff obtained by the NNWSI Project to assess the utility of the methodology. The analysis indicates that laboratory measured strength and deformation data of Calico Hills and Bullfrog tuffs follow an extremal. probability distribution (the third type asymptotic distribution of the smallest values). Monte Carlo simulation is carried out to estimate rock mass deformation moduli using a one-dimensional tuff model proposed by Zimmermann and Finley. We suggest that the results of these analyses be incorporated into the repository design

  18. Importance of creep failure of hard rock in the near field of a nuclear-waste repository

    International Nuclear Information System (INIS)

    Blacic, J.D.

    1981-01-01

    Potential damage resulting from slow creep deformation intuitively seems unlikely for a high-level nuclear waste repository excavated in hard rock. However, recent experimental and modeling results indicate that the processes of time-dependent microcracking and water-induced stress corrosion can lead to significant reductions in strength and alteration of other key rock properties in the near-field region of a repository. We review the small data base supporting these conclusions and stress the need for an extensive laboratory program to obtain the new data that will be required for design of a repository

  19. Hydrological and thermal issues concerning a nuclear waste repository in fractured rocks

    International Nuclear Information System (INIS)

    Wang, J.S.Y.

    1991-12-01

    The characterization of the ambient conditions of a potential site and the assessment of the perturbations induced by a nuclear waste repository require hydrological and thermal investigations of the geological formations at different spatial and temporal scales. For high-level wastes, the near-field impacts depend on the heat power of waste packages and the far-field long-term perturbations depend on the cumulative heat released by the emplaced wastes. Surface interim storage of wastes for several decades could lower the near-field impacts but would have relatively small long-term effects if spent fuels were the waste forms for the repository. One major uncertainty in the assessment of repository impacts is from the variation of hydrological properties in heterogeneous media, including the effects of fractures as high-permeability flow paths for containment migration. Under stress, a natural fracture cannot be represented by the parallel plate model. The rock surface roughness, the contact area, and the saturation state in the rock matrix could significantly change the fracture flow. In recent years, the concern of fast flow through fractures in saturated media has extended to the unsaturated zones. The interactions at different scales between fractures and matrix, between fractured matrix unites and porous units, and between formations and faults are discussed

  20. Plugs for deposition tunnels in a deep geologic repository in granitic rock. Concepts and experience

    International Nuclear Information System (INIS)

    Dixon, D. A.; Boergesson, L.; Gunnarsson, D.; Hansen, J.

    2009-11-01

    Regardless of the emplacement geometry selected in a geological repository for spent nuclear fuel, there will be a requirement for the access tunnels to remain open while repository operations are ongoing. The period of repository operation will stretch for many years (decades to more than a century depending on disposal concept and number of canisters to be installed). Requirements for extended monitoring of the repository before final closure may further extend the period over which the tunnels must remain open. The intersection of the emplacement rooms/drifts and the access tunnels needs to be physically closed in order to ensure that the canisters remain undisturbed and that no undesirable hydraulic conditions are allowed to develop within the backfilled volume. As a result of these requirements, generic guidelines and design concepts have been developed for 'Plugs' that are intended to provide mechanical restraint, physical security and hydraulic control functions over the short-term (repository operational and pre-closure monitoring periods). This report focuses on the role and requirements of plugs to be installed at emplacement room/ tunnel/drift entrances or in other locations within the repository that may require installation of temporary mechanical or hydraulic control structures. These plugs are not necessarily a permanent feature of the repository and may, if required, be removed for later installation of a permanent seal. Room/Drift plugs are also by their defined function, physically accessible during repository operation so their performance can be monitored and remedial actions taken if necessary (e.g. increased seepage past the plug). A considerable number of sealing demonstrations have been undertaken at several research laboratories that are focussed on development of technologies and materials for use in isolation of spent nuclear fuel and these are briefly reviewed in this report. Additionally, technologies developed for non

  1. Plugs for deposition tunnels in a deep geologic repository in granitic rock. Concepts and experience

    Energy Technology Data Exchange (ETDEWEB)

    Dixon, D.A. (AECL, Chalk River (Canada)); Boergesson, L. (Clay Technology, Lund (Sweden)); Gunnarsson, D. (Swedish Nuclear Fuel and Waste Management Co, Stockholm (Sweden)); Hansen, J. (Posiva Oy, Eurajoki (Finland))

    2009-11-15

    Regardless of the emplacement geometry selected in a geological repository for spent nuclear fuel, there will be a requirement for the access tunnels to remain open while repository operations are ongoing. The period of repository operation will stretch for many years (decades to more than a century depending on disposal concept and number of canisters to be installed). Requirements for extended monitoring of the repository before final closure may further extend the period over which the tunnels must remain open. The intersection of the emplacement rooms/drifts and the access tunnels needs to be physically closed in order to ensure that the canisters remain undisturbed and that no undesirable hydraulic conditions are allowed to develop within the backfilled volume. As a result of these requirements, generic guidelines and design concepts have been developed for 'Plugs' that are intended to provide mechanical restraint, physical security and hydraulic control functions over the short-term (repository operational and pre-closure monitoring periods). This report focuses on the role and requirements of plugs to be installed at emplacement room/ tunnel/drift entrances or in other locations within the repository that may require installation of temporary mechanical or hydraulic control structures. These plugs are not necessarily a permanent feature of the repository and may, if required, be removed for later installation of a permanent seal. Room/Drift plugs are also by their defined function, physically accessible during repository operation so their performance can be monitored and remedial actions taken if necessary (e.g. increased seepage past the plug). A considerable number of sealing demonstrations have been undertaken at several research laboratories that are focussed on development of technologies and materials for use in isolation of spent nuclear fuel and these are briefly reviewed in this report. Additionally, technologies developed for non

  2. Modeling gas migration experiments in repository host rocks for the MEGAS project

    International Nuclear Information System (INIS)

    Worgan, K.; Impey, M.; Volckaert, G.; DePreter, P.

    1993-01-01

    In response to concerns over the possibility of hydrogen gas generation within an underground repository for high-level radioactive waste, and its implications for repository safety, a joint European research study (MEGAS) is underway. Its aims are to understand and characterize the behavior of gas migration within an argillacious, host-rock. Laboratory experiments are being carried out by SCK/CEN, BGS and ISMES. SCK/CEN are also conducting in situ experiments at the underground laboratory at Mol, Belgium. Modeling of gas migration is being done in parallel with the experiments, by Intera Information Technologies. A two-phase flow code, TOPAZ, has been developed specifically for this work. In this paper the authors report on the results of some preliminary calculations performed with TOPAZ, in advance of the in situ experiments

  3. Sensitivity and uncertainty analysis applied to a repository in rock salt

    International Nuclear Information System (INIS)

    Polle, A.N.

    1996-12-01

    This document describes the sensitivity and uncertainty analysis with UNCSAM, as applied to a repository in rock salt for the EVEREST project. UNCSAM is a dedicated software package for sensitivity and uncertainty analysis, which was already used within the preceding PROSA project. The use of UNCSAM provides a flexible interface to EMOS ECN by substituting the sampled values in the various input files to be used by EMOS ECN ; the model calculations for this repository were performed with the EMOS ECN code. Preceding the sensitivity and uncertainty analysis, a number of preparations has been carried out to facilitate EMOS ECN with the probabilistic input data. For post-processing the EMOS ECN results, the characteristic output signals were processed. For the sensitivity and uncertainty analysis with UNCSAM the stochastic input, i.e. sampled values, and the output for the various EMOS ECN runs have been analyzed. (orig.)

  4. Assessment of disruptive scenarios of a Canadian used fuel repository in crystalline rock

    Energy Technology Data Exchange (ETDEWEB)

    Gobien, M.; Garisto, F.; Hunt, N.; Kremer, E.P. [Nuclear Waste Management Organization (NWMO), Toronto, Ontario (Canada)

    2015-06-15

    The NWMO has recently extended its modelling capabilities by performing simulations for four disruptive scenarios that, to date, have not yet been examined in detail. These scenarios complement those considered in an existing postclosure safety assessment for a conceptual geological repository located in a hypothetical crystalline rock formation. The four new disruptive scenarios are: Shaft Seal Failure, Undetected Fault, Open or Poorly Sealed Borehole and Open Borehole Due to Inadvertent Human Intrusion. All simulations are based on the FRAC3DVS-OPG Site-Scale Model. The Site-Scale Model includes a simplified representation of the full repository and a portion of the surrounding sub-regional flow system. All transport simulations are performed with only the radionuclide I-129. Transport rates to the surface and a domestic water supply well are compared to the Reference Case results from an earlier case study documented in Reference. (author)

  5. Assessment of disruptive scenarios of a Canadian used fuel repository in crystalline rock

    Energy Technology Data Exchange (ETDEWEB)

    Gobien, M.; Garisto, F.; Hunt, N.; Kremer, E.P., E-mail: mgobien@nwmo.ca [Nuclear Waste Management Organization, Toronto, ON (Canada)

    2015-07-01

    The NWMO has recently extended its modelling capabilities by performing simulations for four disruptive scenarios that, to date, have not yet been examined in detail. These scenarios complement those considered in an existing postclosure safety assessment for a conceptual geological repository located in a hypothetical crystalline rock formation. The four new disruptive scenarios are: Shaft Seal Failure, Undetected Fault, Open or Poorly Sealed Borehole and Open Borehole Due to Inadvertent Human Intrusion. All simulations are based on the FRAC3DVS-OPG [1] Site-Scale Model [2]. The Site-Scale Model includes a simplified representation of the full repository and a portion of the surrounding sub-regional flow system. All transport simulations are performed with only the radionuclide I-129. Transport rates to the surface and a domestic water supply well are compared to the Reference Case results from an earlier case study documented in Reference [2]. (author)

  6. Assessment of disruptive scenarios of a Canadian used fuel repository in crystalline rock

    International Nuclear Information System (INIS)

    Gobien, M.; Garisto, F.; Hunt, N.; Kremer, E.P.

    2015-01-01

    The NWMO has recently extended its modelling capabilities by performing simulations for four disruptive scenarios that, to date, have not yet been examined in detail. These scenarios complement those considered in an existing postclosure safety assessment for a conceptual geological repository located in a hypothetical crystalline rock formation. The four new disruptive scenarios are: Shaft Seal Failure, Undetected Fault, Open or Poorly Sealed Borehole and Open Borehole Due to Inadvertent Human Intrusion. All simulations are based on the FRAC3DVS-OPG [1] Site-Scale Model [2]. The Site-Scale Model includes a simplified representation of the full repository and a portion of the surrounding sub-regional flow system. All transport simulations are performed with only the radionuclide I-129. Transport rates to the surface and a domestic water supply well are compared to the Reference Case results from an earlier case study documented in Reference [2]. (author)

  7. Radionuclide Transport in Fractured Rock: Numerical Assessment for High Level Waste Repository

    Directory of Open Access Journals (Sweden)

    Claudia Siqueira da Silveira

    2013-01-01

    Full Text Available Deep and stable geological formations with low permeability have been considered for high level waste definitive repository. A common problem is the modeling of radionuclide migration in a fractured medium. Initially, we considered a system consisting of a rock matrix with a single planar fracture in water saturated porous rock. Transport in the fracture is assumed to obey an advection-diffusion equation, while molecular diffusion is considered the dominant mechanism of transport in porous matrix. The partial differential equations describing the movement of radionuclides were discretized by finite difference methods, namely, fully explicit, fully implicit, and Crank-Nicolson schemes. The convective term was discretized by the following numerical schemes: backward differences, centered differences, and forward differences. The model was validated using an analytical solution found in the literature. Finally, we carried out a simulation with relevant spent fuel nuclide data with a system consisting of a horizontal fracture and a vertical fracture for assessing the performance of a hypothetical repository inserted into the host rock. We have analysed the bentonite expanded performance at the beginning of fracture, the quantified radionuclide released from a borehole, and an estimated effective dose to an adult, obtained from ingestion of well water during one year.

  8. Acoustic remote monitoring of rock and concrete structures for nuclear waste repositories

    International Nuclear Information System (INIS)

    Young, R.P.

    2000-01-01

    Excavation and thermally induced damage is of significance for many types of engineering structures but no more so than in the case of nuclear waste repository design. My research and that of my group, formally at Queen's University Canada and Keele University UK and now at the University of Liverpool UK, has focused on the development of acoustic techniques for the in situ detection and quantification of induced damage and fracturing. The application of earthquake seismology to this problem has provided the opportunity to study the micro mechanics of damage mechanisms in situ and provide validation data for predictive geomechanical models used for engineering design. Since 1987 I have been a principal investigator at Atomic Energy of Canada's Underground Research Laboratory (URL), responsible for the development of acoustic emission techniques (AE). In the last twelve years, the application of acoustic techniques to rock damage assessment has been pioneered by my group at the URL and successfully applied in several other major international projects including the ZEDEX, Retrieval and Prototype repository experiments at the Aspo Hard Rock Laboratory (HRL) of SKB Sweden. In this paper I describe what information is available by remote acoustic monitoring of rock and concrete structures and demonstrate this with reference to two international scientific experiments carried out at the URL Canada and the HRL Sweden. (author)

  9. Application of Ga-Al discrimination plots in identification of high strength granitic host rocks for deep geological repository of high level radioactive waste

    International Nuclear Information System (INIS)

    Bajpai, R.K.; Narayan, P.K.; Trivedi, R.K.; Purohit, M.K.

    2010-01-01

    The permanent disposal of vitrified high level wastes and in some cases even spent fuel, is being planned in specifically designed and built deep geological repository located in the depth range of 500-600m in appropriate host rock at carefully selected sites. Such facilities are expected to provide very long term isolation and confinement to the disposed waste by means of long term mechanical stability of such structures that results from very high strength and homogeneity of the chosen rock, geochemical compatible environment around the disposed waste and general lack of groundwater. In Indian geological repository development programme, granites have been selected as target host rock and large scale characterization studies have been undertaken to develop database of mineralogy, petrology, geochemistry and rock mechanical characteristics. The paper proposes a new approach for demarcation of high strength homogeneous granite rocks from within an area of about 100 square kilometres wherein a cocktail of granites of different origins with varying rock mass characteristics co exists. The study area is characterised by the presence of A, S and I type granites toughly intermixed. The S type granites are derived from sedimentary parent material and therefore carry relics of parent fabric and at times undigested material with resultant reduction in their strength and increased inhomogeneity. On the other hand I type varieties are derived from igneous parents and are more homogeneous with sufficient strength. The A type granites are emplaced as molten mass in a complete non-tectonic setting with resultant homogeneous compositions, absence of tectonic fabric and very high strength. Besides they are silica rich with less vulnerability to alterations with time. Thus A type granites are most suited for construction of Deep Geological Repository. For developing a geochemical approach for establishing relation between chemical compositions and rock strength parameters, a

  10. Thermomechanical repository and shaft response analyses using the CAVS [Cracking And Void Strain] jointed rock model: Draft final report

    International Nuclear Information System (INIS)

    Dial, B.W.; Maxwell, D.E.

    1986-12-01

    Numerical studies of the far-field repository and near-field shaft response for a nuclear waste repository in bedded salt have been performed with the STEALTH computer code using the CAVS model for jointed rock. CAVS is a constitutive model that can simulate the slip and dilatancy of fracture planes in a jointed rock mass. The initiation and/or propagation of fractures can also be modeled when stress intensity criteria are met. The CAVS models are based on the joint models proposed with appropriate modifications for numerical simulations. The STEALTH/CAVS model has been previously used to model (1) explosive fracturing of a wellbore, (2) earthquake effects on tunnels in a generic nuclear waste repository, (3) horizontal emplacement for a nuclear waste repository in jointed granite, and (4) tunnel response in jointed rock. The use of CAVS to model far-field repository and near-field shaft response was different from previous approaches because it represented a spatially oriented approach to rock response and failure, rather than the traditional stress invariant formulation for yielding. In addition, CAVS tracked the response of the joint apertures to the time-dependent stress changes in the far-field repository and near-field shaft regions. 28 refs., 21 figs., 11 tabs

  11. Evaluation of backfill materials for a shallow-depth repository

    International Nuclear Information System (INIS)

    Buckley, L.P.; Arbique, G.M.; Tosello, N.B.; Woods, B.L.

    1986-11-01

    The focus of laboratory research effort on the disposal of low- and intermediate-level radioactive waste is to determine what conditions will dominate and which engineered barriers will be most effective for the retention of radionuclides. Initial studies have concentrated on the evaluation of a flooded repository and the assessment of backfill materials suitable for the adsorption of radioactivity, yet permeable enough to allow excess water to pass through the repository and into the underlying water table. Both physical and adsorption studies have been performed. Based on these preliminary experiments, it is felt that a mixture of 10 wt% clay and the remainder sand would satisfy the above criteria. Since both are available within the Ottawa Valley, they also have the added advantage of being more cost effective to use than imported materials

  12. Bentonite as a backfill material for shallow land repositories

    International Nuclear Information System (INIS)

    Yalmali, V.S.; Deshingkar, D.S.

    2001-01-01

    Two commercially available indigenous bentonite samples were evaluated for their cesium and strontium sorption properties in distilled water and surface water. By converting them into sodium form, the distribution coefficients for both cesium (I) and strontium (II) increased. Sodium bentonite was recommended because of high sorption capacity for Cs(I), Mg(II) and Sr(II) for use as backfill material in shallow land repositories where cement waste form containing Cs, Sr and Be wastes are disposed. (author)

  13. Material control and accountability procedures for a waste isolation repository

    International Nuclear Information System (INIS)

    Jenkins, J.D.; Allen, E.J.; Blakeman, E.D.

    1978-05-01

    The material control and accountability needs of a waste isolation repository are examined. Three levels of control are discussed: (1) item identification and control, (2) tamper indication, and (3) quantitative material assay. A summary of waste characteristics is presented and, based on these, plus a consideration of the accessibility of the various types of waste, material control by item identification and accountability (where the individual waste container is the basic unit) is recommended. Tamper indicating procedures are also recommended for the intermediate and low level waste categories

  14. Material constitutive model for jointed rock mass behavior

    International Nuclear Information System (INIS)

    Thomas, R.K.

    1980-11-01

    A material constitutive model is presented for jointed rock masses which exhibit preferred planes of weakness. This model is intended for use in finite element computations. The immediate application is the thermomechanical modelling of a nuclear waste repository in hard rock, but the model seems appropriate for a variety of other static and dynamic geotechnical problems as well. Starting with the finite element representations of a two-dimensional elastic body, joint planes are introduced in an explicit manner by direct modification of the material stiffness matrix. A novel feature of this approach is that joint set orientations, lengths and spacings are readily assigned through the sampling of a population distribution statistically determined from field measurement data. The result is that the fracture characteristics of the formations have the same statistical distribution in the model as is observed in the field. As a demonstration of the jointed rock mass model, numerical results are presented for the example problem of stress concentration at an underground opening

  15. Feasibility assessment of copper-base waste package container materials in a tuff repository

    International Nuclear Information System (INIS)

    Acton, C.F.; McCright, R.D.

    1986-01-01

    This report discussed progress made during the second year of a two-year study on the feasibility of using copper or a copper-base alloy as a container material for a waste package in a potential repository in tuff rock at the Yucca Mountain site in Nevada. Corrosion testing in potentially corrosive irradiated environments received emphasis during the feasibility study. Results of experiments to evaluate the effect of a radiation field on the uniform corrosion rate of the copper-base materials in repository-relevant aqueous environments are given as well as results of an electrochemical study of the copper-base materials in normal and concentrated J-13 water. Results of tests on the irradiation of J-13 water and on the subsequent formation of hydrogen peroxide are given. A theoretical study was initiated to predict the long-term corrosion behavior of copper in the repository. Tests were conducted to determine whether copper would adversely affect release rates of radionuclides to the environment because of degradation of the Zircaloy cladding. A manufacturing survey to determine the feasibility of producing copper containers utilizing existing equipment and processes was completed. The cost and availability of copper was also evaluated and predicted to the year 2000. Results of this feasibility assessment are summarized

  16. Exploration of cystalline rocks for nuclear waste repositories: Some strategies for area characterization

    International Nuclear Information System (INIS)

    Trask, N.J.; Roseboom, E.H.; Watts, R.D.; Bedinger, M.S.

    1991-01-01

    A general strategy for the exploration of crystalline rock massed in the eastern United States for the identification of potential sites for high-level radioactive waste repositories has been generated by consideration of the Department of Energy (DOE) Siting Guidelines, available information on these crystalline rocks, and the capabilities and limitations of various exploration methods. The DOE has recently screened over 200 crystalline rock massed in 17 states by means of literature surveys and has recommended 12 rock masses for more intensive investigation including field investigations. The suggested strategy applies to the next stage of screen where the objective is to identify those potential sites that merit detailed site characterization including an exploratory shaft and underground study. This document discusses strategies for reconnaissance and field investigations, including the early phases of drilling, to provide geoscience information on the areas under construction. A complete Area Characterization Plan, to be developed by DOE with involvement of the states within which the areas to be studied are located, will outline all of the investigations to be carried out in the area phase including their cost and scheduling. Here, we provide input for the Area Characterization Plan by discussing what we believe to be the most important issues that need to be addressed in this phase and suggesting methods for their resolution. This report is not intended as a complete outline of area phase geoscience investigations, however. 79 refs., 4 figs

  17. Salt repository sealing materials development program: 5-year work plan

    International Nuclear Information System (INIS)

    Myers, L.B.

    1986-06-01

    This plan covers 5 years (fiscal years 1986 through 1990) of work in the repository sealing materials program to support design decisions and licensing activities for a salt repository. The plan covers a development activity, not a research activity. There are firm deliverables as the end points of each part of the work. The major deliverables are: development plans for code development and materials testing; seal system components models; seal system performance specifications; seal materials specifications; and seal materials properties ''handbook.'' The work described in this plan is divided into three general tasks as follows: mathematical modeling; materials studies (salt, cementitious materials, and earthen materials); and large-scale testing. Each of the sections presents an overview, status, planned activities, and summary of program milestones. This plan will be the starting point for preparing the development plans described above, but is subject to change if preparation of the work plan indicates that a different approach or sequence is preferable to achieve the ultimate goal, i.e., support of design and licensing

  18. Compact rock material gas permeability properties

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Huanling, E-mail: whl_hm@163.com [Key Laboratory of Coastal Disaster and Defence, Ministry of Education, Hohai University, Nanjing 210098 (China); LML, University of Lille, Cite Scientifique, 59655 Villeneuve d’Ascq (France); Xu, Weiya; Zuo, Jing [Institutes of Geotechnical Engineering, Hohai University, Nanjing 210098 (China)

    2014-09-15

    Natural compact rocks, such as sandstone, granite, and rock salt, are the main materials and geological environment for storing underground oil, gas, CO{sub 2,} shale gas, and radioactive waste because they have extremely low permeabilities and high mechanical strengths. Using the inert gas argon as the fluid medium, the stress-dependent permeability and porosity of monzonitic granite and granite gneiss from an underground oil storage depot were measured using a permeability and porosity measurement system. Based on the test results, models for describing the relationships among the permeability, porosity, and confining pressure of rock specimens were analyzed and are discussed. A power law is suggested to describe the relationship between the stress-dependent porosity and permeability; for the monzonitic granite and granite gneiss (for monzonitic granite (A-2), the initial porosity is approximately 4.05%, and the permeability is approximately 10{sup −19} m{sup 2}; for the granite gneiss (B-2), the initial porosity is approximately 7.09%, the permeability is approximately 10{sup −17} m{sup 2}; and the porosity-sensitivity exponents that link porosity and permeability are 0.98 and 3.11, respectively). Compared with moderate-porosity and high-porosity rocks, for which φ > 15%, low-porosity rock permeability has a relatively lower sensitivity to stress, but the porosity is more sensitive to stress, and different types of rocks show similar trends. From the test results, it can be inferred that the test rock specimens’ permeability evolution is related to the relative particle movements and microcrack closure.

  19. Natural analogue studies in crystalline rock: the influence of water-bearing fractures on radionuclide immobilisation in a granitic rock repository

    International Nuclear Information System (INIS)

    Alexander, W.R.; MacKenzie, A.B.; Scott, R.D.; McKinley, I.G.

    1990-06-01

    Current Swiss concepts for the disposal of radioactive waste involve disposal in deep mined repositories to ensure that only insignificant quantities of radionuclides will ever reach the surface and so enter the biosphere. The rock formations presently considered as potential candidates for hosting radwaste repositories have thus been selected on the basis of their capacity to isolate radionuclides from the biosphere. An important factor in ensuring such containment is a very low solute transport rate through the host formation. However, it is considered likely that, in the formations of interest in the Swiss programme (eg. granites, argillaceous sediments, anhydrite), the rocks will be fractured to some extent even at repository depth. In the instance of the cumulative failure of near-field barriers in the repository, these hydraulically connected fractures in the host formation could be very important far-field routes of migration (and possible sites of retardation) of radionuclides dissolved in the groundwaters. In this context, the so-called 'matrix diffusion' mechanism is potentially very important for radionuclide retardation. This report is the culmination of a programme which has attempted to assess the potential influence of these water-bearing fractures on radionuclide transport in a crystalline rock radwaste repository. 162 refs., 36 figs., 16 tabs

  20. Studies of ancient concrete as analogs of cementitious sealing materials for a repository in tuff

    Energy Technology Data Exchange (ETDEWEB)

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

    1989-03-01

    The durability of ancient cementitious materials has been investigated to provide data applicable to determining the resistance to weathering of concrete materials for sealing a repository for storage of high-level radioactive waste. Because tuff and volcanic ash are used in the concretes in the vicinity of Rome, the results are especially applicable to a waste repository in tuff. Ancient mortars, plasters, and concretes collected from Rome, Ostia, and Cosa dating to the third century BC show remarkable durability. The aggregates used in the mortars, plasters, and concretes included basic volcanic and pyroclastic rocks (including tuff), terra-cotta, carbonates, sands, and volcanic ash. The matrices of ancient cementitious materials have been characterized and classified into four categories: (1) hydraulic hydrated lime and hydrated lime cements, (2) hydraulic aluminous and ferruginous hydrated lime cements ({plus_minus} siliceous components), (3) pozzolana/hydrated lime cements, and (4) gypsum cements. Most of the materials investigated are in category (3). The materials were characterized to elucidate aspects of the technology that produced them and their response to the environmental exposure throughout their centuries of existence. Their remarkable properties are the result of a combination of chemical, mineralogical, and microstructural factors. Their durability was found to be affected by the matrix mineralogy, particle size, and porosity; aggregate type, grading and proportioning; and the methodology of placement. 30 refs.

  1. Superhard nanophase materials for rock drilling applications

    Energy Technology Data Exchange (ETDEWEB)

    Sadangi, R.K.; Voronov, O.A.; Tompa, G.S. [Diamond Materials Inc., Pisctaway, NJ (United States); Kear, B.H. [Rutgers Univ., Piscataway, NJ (United States)

    1997-12-31

    Diamond Materials Incorporated is developing new class of superhard materials for rock drilling applications. In this paper, we will describe two types of superhard materials, (a) binderless polycrystalline diamond compacts (BPCD), and (b) functionally graded triphasic nanocomposite materials (FGTNC). BPCDs are true polycrystalline diamond ceramic with < 0.5 wt% binders and have demonstrated to maintain their wear properties in a granite-log test even after 700{degrees}C thermal treatment. FGTNCs are functionally-graded triphasic superhard material, comprising a nanophase WC/Co core and a diamond-enriched surface, that combine high strength and toughness with superior wear resistance, making FGTNC an attractive material for use as roller cone stud inserts.

  2. Unsaturated flow and transport through fractured rock related to high-level waste repositories

    International Nuclear Information System (INIS)

    Evans, D.D.; Rasmussen, T.C.

    1991-01-01

    Research results are summarized for a US Nuclear Regulatory Commission contract with the University of Arizona focusing on field and laboratory methods for characterizing unsaturated fluid flow and solute transport related to high-level radioactive waste repositories. Characterization activities are presented for the Apache Leap Tuff field site. The field site is located in unsaturated, fractured tuff in central Arizona. Hydraulic, pneumatic, and thermal characteristics of the tuff are summarized, along with methodologies employed to monitor and sample hydrologic and geochemical processes at the field site. Thermohydrologic experiments are reported which provide laboratory and field data related to the effects conditions and flow and transport in unsaturated, fractured rock. 29 refs., 17 figs., 21 tabs

  3. GEOCHEMISTRY OF ROCK UNITS AT THE POTENTIAL REPOSITORY LEVEL, YUCCA MOUNTAIN, NEVADA

    International Nuclear Information System (INIS)

    Peterman, Z.E.; Cloke, P.L.

    2000-01-01

    The compositional variability of the phenocryst-poor member of the 12.8-million-year Topopah Spring Tuff at the potential repository level was assessed by duplicate analysis of 20 core samples from the cross drift at Yucca Mountain, Nevada. Previous analyses of outcrop and core samples of the Topopah Spring Tuff showed that the phenocryst-poor rhyolite, which includes both lithophysal and nonlithophysal zones, is relatively uniform in composition. Analyses of rock samples from the cross drift, the first from the actual potential repository block, also indicate the chemical homogeneity of this unit excluding localized deposits of vapor-phase minerals and low-temperature calcite and opal in fractures, cavities, and faults, The possible influence of vapor-phase minerals and calcite and opal coatings on rock composition at a scale sufficiently large to incorporate these heterogeneously distributed deposits was evaluated and is considered to be relatively minor. Therefore, the composition of the phenocryst-poor member of the Topopah Spring Tuff is considered to be adequately represented by the analyses of samples from the cross drift. The mean composition as represented by the 10 most abundant oxides in weight percent or grams per hundred grams is: SiO 2 , 76.29; Al 2 O 3 , 12.55; FeO, 0.14; Fe 2 O 3 , 0.97; MgO, 0.13; CaO, 0.50; Na 2 O, 3.52; K 2 O, 4.83; TiO 2 , 0.11; and MnO, 0.07

  4. Influence of convective-energy transfer on calculated temperature distributions in proposed hard-rock nuclear waste repositories

    Energy Technology Data Exchange (ETDEWEB)

    Eaton, R R; Reda, D C [Sandia National Labs., Albuquerque, NM (USA)

    1982-06-01

    This study assesses the relative influence of convective-energy transfer on predicted temperature distributions for a nuclear-waste repository located in water-saturated rock. Using results for energy transfer by conduction only (no water motion) as a basis of comparison, it is shown that a considerable amount of energy can be removed from the repository by pumping out water that migrates into the drift from regions adjacent to the buried waste canisters. Furthermore, the results show that the influence of convective-energy transfer on mine drift cooling requirements can be significant for cases where the in-situ permeability of the rock is greater than one millidarcy (a regime potentially encountered in repository scenarios).

  5. Integrated Analytic Radionuclide Transport Model for a Spent Nuclear Fuel Repository in Saturated Fractured Rock

    International Nuclear Information System (INIS)

    Hedin, Allan

    2002-01-01

    Simple analytic expressions are presented for radionuclide transport from a KBS 3-type repository, where spent nuclear fuel is placed in copper canisters surrounded by bentonite clay and deposited at a depth of 500 m in fractured granitic rock.Dissolution of readily accessible and fuel matrix embedded nuclides, chain decay, and nuclide precipitation is treated within the canister. Transport in the canister void and buffer is modeled with a dual stirred tank analogy, where transport resistances represent an assumed small initial damage in the canister and transport features of the buffer-geosphere interface. Initial, transient diffusion in the buffer is treated with a simple correction term. Chain decay is not included in the buffer.Geosphere transport expressions handle advection, longitudinal dispersion, matrix diffusion, sorption, and radioactive decay, but not chain decay. The treatment is based on earlier results for an instantaneous inlet and for a constant inlet to the geosphere in the nondispersive case. A correction is added so that longitudinal dispersion is taken approximately into account. The correction utilizes analytical expressions for the temporal moments of the geosphere release curve in the dispersive case.The near-field/geosphere integration is treated in a simplified manner avoiding numerical convolutions. The instantaneous inlet expression for the geosphere release is used when the near-field release decreases rapidly in comparison to a typical response time in the geosphere; the constant inlet expression is used in the opposite case.Twenty-seven calculation cases from a safety assessment of a KBS 3 repository using borehole data from three different field investigation sites were repeated with the analytic expressions. The agreement in both near-field and geosphere releases is in general well within an order of magnitude for the variety of long- and short-lived, sorbing, nonsorbing, solubility limited, immediately accessible, and fuel matrix

  6. Choice of rock excavation methods for the Swedish deep repository for spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Baeckblom, Goeran [Conrox, Stockholm (Sweden); Christiansson, Rolf [Swedish Nuclear Fuel and Waste Management Co., Stockholm (Sweden); Lagerstedt, Leif [SwedPower AB, Stockholm (Sweden)

    2004-09-01

    Choice of rock excavation methods will or may have implications for a number of issues like repository layout, long term and operational safety, environmental impact, design of and operation of transport vehicles and methodology for backfilling the repository before closure as well as effects on costs and schedules. To fully analyse the issues at hand related to selection of excavation methods, SKB organized a project with the objectives: To investigate and compare principal technical solutions for rock excavation, both methods that are used at present but also methods that may be feasible 10 years from now; To assess how the selection of excavation method influences the design and operation of the deep repository; To present a definition of the Excavation Damaged/Disturbed Zone and practical methods for measurements of EDZ; To present advantages and disadvantages with different excavation methods for the various tunnels and underground openings as a basis for selection of preferred excavation methods; To present the Design Justification Statement for the selection of particular excavation methods for the different tunnels and openings in the deep repository to underpin a decision on excavation method; and To present background data that may be required for the evaluation of the long term safety of the deep repository. Main alternatives studied are very smooth blasting, excavation with a tunnel-boring machine (TBM) and excavation with horizontal pull-reaming using more or less conventional raise-boring equipment. The detailed studies were carried through in co-operation with major suppliers and end-users of the technology. An observation in this study is that all excavation technologies are mature; no major breakthroughs are foreseen within a 10 year period but it is likely that for any technology selected, SKB would specifically fine-tune the design of the equipment and work procedures in view of requirements and site specific conditions. Excavation methods have

  7. Choice of rock excavation methods for the Swedish deep repository for spent nuclear fuel

    International Nuclear Information System (INIS)

    Baeckblom, Goeran; Christiansson, Rolf; Lagerstedt, Leif

    2004-09-01

    Choice of rock excavation methods will or may have implications for a number of issues like repository layout, long term and operational safety, environmental impact, design of and operation of transport vehicles and methodology for backfilling the repository before closure as well as effects on costs and schedules. To fully analyse the issues at hand related to selection of excavation methods, SKB organized a project with the objectives: To investigate and compare principal technical solutions for rock excavation, both methods that are used at present but also methods that may be feasible 10 years from now; To assess how the selection of excavation method influences the design and operation of the deep repository; To present a definition of the Excavation Damaged/Disturbed Zone and practical methods for measurements of EDZ; To present advantages and disadvantages with different excavation methods for the various tunnels and underground openings as a basis for selection of preferred excavation methods; To present the Design Justification Statement for the selection of particular excavation methods for the different tunnels and openings in the deep repository to underpin a decision on excavation method; and To present background data that may be required for the evaluation of the long term safety of the deep repository. Main alternatives studied are very smooth blasting, excavation with a tunnel-boring machine (TBM) and excavation with horizontal pull-reaming using more or less conventional raise-boring equipment. The detailed studies were carried through in co-operation with major suppliers and end-users of the technology. An observation in this study is that all excavation technologies are mature; no major breakthroughs are foreseen within a 10 year period but it is likely that for any technology selected, SKB would specifically fine-tune the design of the equipment and work procedures in view of requirements and site specific conditions. Excavation methods have

  8. Modeling of irradiated graphite {sup 14}C transfer through engineered barriers of a generic geological repository in crystalline rocks

    Energy Technology Data Exchange (ETDEWEB)

    Poskas, Povilas; Grigaliuniene, Dalia, E-mail: Dalia.Grigaliuniene@lei.lt; Narkuniene, Asta; Kilda, Raimondas; Justinavicius, Darius

    2016-11-01

    There are two RBMK-1500 type graphite moderated reactors at the Ignalina nuclear power plant in Lithuania, and they are under decommissioning now. The graphite cannot be disposed of in a near surface repository, because of large amounts of {sup 14}C. Therefore, disposal of the graphite in a geological repository is a reasonable solution. This study presents evaluation of the {sup 14}C transfer by the groundwater pathway into the geosphere from the irradiated graphite in a generic geological repository in crystalline rocks and demonstration of the role of the different components of the engineered barrier system by performing local sensitivity analysis. The speciation of the released {sup 14}C into organic and inorganic compounds as well as the most recent information on {sup 14}C source term was taken into account. Two alternatives were considered in the analysis: disposal of graphite in containers with encapsulant and without it. It was evaluated that the maximal fractional flux of inorganic {sup 14}C into the geosphere can vary from 10{sup −} {sup 11} y{sup −} {sup 1} (for non-encapsulated graphite) to 10{sup −} {sup 12} y{sup −} {sup 1} (for encapsulated graphite) while of organic {sup 14}C it was about 10{sup −} {sup 3} y{sup −} {sup 1} of its inventory. Such difference demonstrates that investigations on the {sup 14}C inventory and chemical form in which it is released are especially important. The parameter with the highest influence on the maximal flux into the geosphere for inorganic {sup 14}C transfer was the sorption coefficient in the backfill and for organic {sup 14}C transfer – the backfill hydraulic conductivity. - Highlights: • Graphite moderated nuclear reactors are being decommissioned. • We studied interaction of disposed material with surrounding environment. • Specifically {sup 14}C transfer through engineered barriers of a geological repository. • Organic {sup 14}C flux to geosphere is considerably higher than inorganic

  9. Dessicant materials screening for backfill in a salt repository

    International Nuclear Information System (INIS)

    Simpson, D.R.

    1980-10-01

    Maintaining an anhydrous environment around nuclear waste stored in a salt repository is a concern which can be alleviated by using a desiccant material for backfilling. Such a desiccant should desiccate a brine yet be non deliquescent, the hydrated product should have moderate thermal stability, and the desiccant should have a high capacity and be readily available. From a literature search MgO and CaO were identified for detailed study. These oxides, and an intimate mixture of the two obtained by calcining dolomite, were used in experiments to further determine their suitability. They proved to be excellent desiccants with a high water capacity. The hydrates of both have moderate thermal stability and a high water content. Both MgO and CaO react in an alkaline chloride brine forming oxychloride compounds with different waters of crystallization. Some of these compounds are the Sorel Cements. CaO hydrates to Ca(OH) 2 which carbonates with CO 2 in air to form CaCO 3 and release the hydrated water. Thus the intimate mixture of CaO and MgO from calcined dolomite may serve as a desiccant and remove CO 2 from the repository atmosphere

  10. Dessicant materials screening for backfill in a salt repository

    Energy Technology Data Exchange (ETDEWEB)

    Simpson, D.R.

    1980-10-01

    Maintaining an anhydrous environment around nuclear waste stored in a salt repository is a concern which can be alleviated by using a desiccant material for backfilling. Such a desiccant should desiccate a brine yet be non deliquescent, the hydrated product should have moderate thermal stability, and the desiccant should have a high capacity and be readily available. From a literature search MgO and CaO were identified for detailed study. These oxides, and an intimate mixture of the two obtained by calcining dolomite, were used in experiments to further determine their suitability. They proved to be excellent desiccants with a high water capacity. The hydrates of both have moderate thermal stability and a high water content. Both MgO and CaO react in an alkaline chloride brine forming oxychloride compounds with different waters of crystallization. Some of these compounds are the Sorel Cements. CaO hydrates to Ca(OH)/sub 2/ which carbonates with CO/sub 2/ in air to form CaCO/sub 3/ and release the hydrated water. Thus the intimate mixture of CaO and MgO from calcined dolomite may serve as a desiccant and remove CO/sub 2/ from the repository atmosphere.

  11. Formation and fate of gases in the caverns of a repository in salt rock

    International Nuclear Information System (INIS)

    Mueller, W.; Morlock, G.; Gronemeyer, C.

    1992-01-01

    The report summarizes the knowledge avaible today of the mechanisms governing the formation and transport of gases in a salt mine repository for radioactive wastes. The work under review deals with the formation of gases-by way of radiolysis, corrosion, microbial degradation, thermally induced or primary gas generation - and analyses the efficiency of predicting and modelling the gas generation mechanisms in terms of the role of parameters involved, and accuracy. Existing gaps in available knowledge are shown and defined in terms of significance, leading to an analysis of interdependencies between the various mechanisms and to a statement concerning the necessity of establishing materials balances. (orig./EF) [de

  12. Repository seal materials performance for a SALT Repository Project 5-year code/model development plan: Draft

    International Nuclear Information System (INIS)

    1987-06-01

    This document describes an integrated laboratory testing and model development effort for the seal system for a high-level nuclear waste repository in salt. The testing and modeling efforts are designed to determine seal material response in the repository environment, to provide models of seal system components for performance assessment, and to assist in the development of seal system designs. A code/model development and performance analysis program will be performed to predict the short- and long-term response of seal materials and seal components. The results from these analyses will be used to support the material testing activities on this contract and to support performance assessment activities that are conducted in other parts of the Salt Repository Project (SRP). 48 refs., 15 figs., 4 tabs

  13. Weathering products of basic rocks as sorptive materials of natural radionuclides

    International Nuclear Information System (INIS)

    Omelianenko, B.I.; Niconov, B.S.; Ryzhov, B.I.; Shikina, N.D.

    1994-06-01

    The principal requirements for employing natural minerals as buffer and backfill material in high-level waste (HLW) repositories are high sorptive properties, low water permeability, relatively high thermal conductivity, and thermostability. The major task of the buffer is to prevent the penetration of radionuclides into groundwater. The authors of this report examined weathered basic rocks from three regions of Russia in consideration as a suitable radioactive waste barrier

  14. Rock quality designation of the hydraulic properties in the near field of a final repository for spent nuclear fuel

    International Nuclear Information System (INIS)

    Carlsson, Hans; Carlsson, Leif; Pusch, Roland

    1989-06-01

    Quality assurance of a final repository for spent nuclear fuel requires detailed information on the characteristics of the rock, backfill, canisters and the waste itself. Furthermore, and of fundamental importance, is the knowledge of the behaviour of the integrated system of the waste and the different barriers. The in-situ characteristics of the rock must therefore be assessed and their influence on and interactions with the remaining barriers must be predicted and verified. A rock quality designation process of the hydraulic properties in the near-field is out-lined both for the KBS-3 system as well as for the WP-cave system. The process, once updated and approved, will be included in a Quality Assurance Program for the final repository for spent nuclear fuel. Some of the available methods for the near-field designation process are presented as well as techniques that need further development or are not developed at all. Finally, a presentation is given of a generic designation process of the KBS-3 and WP-cave repository systems in the previously investigated area in Central Sweden, where the final repository for reactor waste, SFR, is located. Geological and hydrogeological data are here at hand and it is therefore possible to carry out a simulation of how the designation process would be accomplished. (authors) (72 figs., 12 tabs., 43 refs.)

  15. 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)

  16. Corrosion of container and infrastructure materials under clay repository conditions

    International Nuclear Information System (INIS)

    Debruyn, W.; Dresselaers, J.; Vermeiren, P.; Kelchtermans, J.; Tas, H.

    1991-01-01

    With regard to the disposal of high-level radioactive waste, it was recommended in a IAEA Technical Committee meeting to perform tests in realistic environments corresponding with normal and accidental conditions, to qualify and apply corrosion monitoring techniques for corrosion evaluation under real repository conditions and to develop corrosion and near-field evolution models. The actual Belgian experimental programme for the qualification of a container for long-term HLW storage in clay formations complies with these recommendations. The emphasis in the programme is indeed on in situ corrosion testing and monitoring and on in situ control of the near-field chemistry. Initial field experiments were performed in a near-surface clay quarry at Terhaegen. Based on a broad laboratory material screening programme and in agreement with the Commission of the European Communities, three reference materials were chosen for extensive in situ overpack testing. Ti/0.2 Pd and Hastelloy C-4 were chosen as reference corrosion resistant materials and a low-carbon steel as corrosion allowance reference material. This report summarizes progress made in the material qualification programme since the CEC contract of 1983-84. 57 Figs.; 15 Tabs.; 18 Refs

  17. Rock mechanics methods and in situ heater tests for design of a nuclear waste repository in basalt

    International Nuclear Information System (INIS)

    Board, M.P.

    1978-01-01

    Methods of integrating data from the Near-Surface Test Facility into the overall Waste Isolation Program are examined. Discussions are presented dealing primarily with the application of numerical models to the design of a waste repository. The various types of models currently available are discussed with reference to design in basalt and the breakdown of the problem of repository design is summarized. It is shown that the most efficient method for analyzing repository design is to break the problem down into several problems which are based on physical scale. These include the area directly surrounding a single waste canister (the very near field), the area including many canisters and canister emplacement rooms (the near field), and the area including the entire repository and the rock mass to the free surface (the far field). The methods by which numerical models are used for design are discussed. Flow charts are used to show the basic input data required, the calculational processes used, and the preliminary criteria for judgment of suitable repository performance. It is shown that the ultimate design of the allowable gross thermal loading density, and, thus, the layout of the underground workings is highly dependent upon the rock mass properties supplied as base line input data to the numerical models. Of the many input properties required, the thermal conductivity, the thermal expansion coefficient, and elastic moduli of the rock mass have, perhaps, the greatest effect on the calculation of induced temperatures, stresses, and displacements and, thus, repository design. To ensure that the design continues with confidence, field (in situ) values of input data must be obtained. The role of the Near-Surface Test Facility in situ testing in obtaining these basic required data is discussed

  18. Building the safety case for a hypothetical underground repository in crystalline rock. Final report. Vol. 2. Safety file

    International Nuclear Information System (INIS)

    Biurrun, E.; Engelmann, H.J.; Jobmann, M.; Lommerzheim, A.; Popp, W.; Frentz, R.R. v.; Wahl, A.

    1996-10-01

    The study was intended as a desk simulation of the process of preparing a licensing application for a deep repository for spent fuel and high level waste in crystalline rock. After clarifying of organizational aspects of table of contents specifying all aspects in a safety life for license application were considered. The volume II is subdivided in two parts. Part A describes the general information, waste description, site characteristics, disposal facility design, reporitory construction and operation, quality assurance, operational safety, repository closure, organization and financial aspects, and long-term safety assessment. Part B deals with the impact of retrievability. (DG)

  19. Selection of candidate canister materials for high-level nuclear waste containment in a tuff repository

    International Nuclear Information System (INIS)

    McCright, R.D.; Weiss, H.; Juhas, M.C.; Logan, R.W.

    1983-11-01

    A repository located at Yucca Mountain at the Nevada Test Site is a potential site for permanent geological disposal of high-level nuclear waste. The repository can be located in a horizon in welded tuff, a volcanic rock, which is above the static water level at this site. The environmental conditions in this unsaturated zone are expected to be air and water vapor dominated for much of the containment period. Type 304L stainless steel is the reference material for fabricating canisters to contain the solid high-level wastes. Alternative stainless alloys are considered because of possible susceptibility of 304L to localized and stress forms of corrosion. For the reprocessed glass wastes, the canisters serve as the recipient for pouring the glass with the result that a sensitized microstructure may develop because of the times at elevated temperatures. Corrosion testing of the reference and alternative materials has begun in tuff-conditioned water and steam environments. 21 references, 8 figures, 8 tables

  20. Geochemical Characteristics of the Gyeongju LILW Repository II. Rock and Minera

    International Nuclear Information System (INIS)

    Kim, Geon Young; Koh, Yong Kwon; Choi, Byoung Young; Shin, Seon Ho; Kim, Doo Haeng

    2008-01-01

    Geochemical study on the rocks and minerals of the Gyeongju low and intermediate level waste repository was carried out in order to provide geochemical data for the safety assessment and geochemical modeling. Polarized microscopy, X-ray diffraction method, chemical analysis for the major and trace elements, scanning electron microscopy (SEM), and stable isotope analysis were applied. Fracture zones are locally developed with various degrees of alteration in the study area. The study area is mainly composed of granodiorite and diorite and their relation is gradational in the field. However, they could be easily distinguished by their chemical property. The granodiorite showed higher Sig 2 content and lower MgO and Fe 2 O 3 contents than the diorite. Variation trends of the major elements of the granodiorite and diorite were plotted on the same line according to the increase of Sig 2 content suggesting that they were differentiated from the same magma. Spatial distribution of the various elements showed that the diorite region had lower Sig 2 , Al 2 O 3 , Na 2 O and K 2 O contents, and higher CaO, Fe 2 O 3 contents than the granodiorite region. Especially, because the differences in the CaO and Na 2 O distribution were most distinct and their trends were reciprocal, the chemical variation of the plagioclase of the granitic rocks was the main parameter of the chemical variation of the host rocks in the study area. Identified fracture-filling minerals from the drill core were montmorillonite, zeolite minerals, chlorite, illite, calcite and pyrite. Especially pyrite and laumontite, which are known as indicating minerals of hydrothermal alteration, were widely distributed in the study area indicating that the study area was affected by mineralization and/or hydrothermal alteration. Sulfur isotope analysis for the pyrite and oxygen-hydrogen stable isotope analysis for the clay minerals indicated that they were originated from the magma. Therefore, it is considered that

  1. Safety case approach for a KBS-3 type repository in crystalline rock

    International Nuclear Information System (INIS)

    Pastina, Barbara; Lehikoinen, Jarmo; Puigdomenech, Ignasi

    2012-01-01

    Barbara Pastina of Saanio and Riekkola described the approach to considering cementitious materials in a safety case for a KBS-3 repository in Finland. In this concept, cements will be used predominantly as tunnel plugs and seals. Part of the Finnish approach has involved identifying the cement-related FEPs. For example, FEPs representing the effects of cement on spent fuel, on the canister and on radionuclide transport include: - Fuel matrix dissolution at high pH. - Copper corrosion at high pH. - Radionuclide speciation and solubility at high pH. - Radionuclide sorption and diffusion at high pH. - Radionuclide transport due to organic materials (e.g. super-plasticisers). - Colloid formation at a high pH plume front. FEPs representing the effects of cement on bentonite in the buffer and backfill include: - Potential changes in swelling pressure due to mass loss, decrease in clay density, and precipitation of secondary minerals. - Potential cracking and increase of hydraulic conductivity due to cementation. - Increase of the cation exchange capacity due to the loss of silica from the montmorillonite structure. Amongst the cement-related FEPs, the main concerns are related to effects on the performance of the bentonite buffer. Cement-bentonite interactions are complex, there are few experimental data, and there are significant modelling uncertainties (e.g. limited knowledge about the reactions that may occur and their rates, and the effects of temperature). Accepting the existence of various uncertainties, preliminary modelling studies performed using the TOUGHREACT code illustrate the potential for porosity reduction and clogging of porosity in bentonite affected by cementitious pore waters. The modelling also suggests that that the high pH of the pore waters moving from the cementitious materials into the bentonite may be rapidly lowered as a result of reactions with the bentonite close to the cement-bentonite interface. Taking account of the various research and

  2. Alteration of national glass in radioactive waste repository host rocks: A conceptional review

    International Nuclear Information System (INIS)

    Apps, J.A.

    1987-01-01

    The storage of high-level radioactive wastes in host rocks containing natural glass has potential chemical advantages, especially if the initial waste temperatures are as high as 250 0 C. However, it is not certain how natural glasses will decompose when exposed to an aqueous phase in a repository environment. The hydration and devitrification of both rhyolitic and natural basaltic natural glasses are reviewed in the context of hypothetical thermodynamic phase relations, infrared spectroscopic data and laboratory studies of synthetic glasses exposed to steam. The findings are compared with field observations and laboratory studies of hydrating and devitrifying natural glasses. The peculiarities of the dependence of hydration and devitrification behavior on compositional variation is noted. There is substantial circumstantial evidence to support the belief that rhyolitic glasses differ from basaltic glasses in their thermodynamic stability and their lattice structure, and that this is manifested by a tendency of the former to hydrate rather than devitrify when exposed to water. Further research remains to be done to confirm the differences in glass structure, and to determine both physically and chemically dependent properties of natural glasses as a function of composition

  3. GRS/ISTec strategy for the treatment of gas-related issues for repositories located in rock salt

    International Nuclear Information System (INIS)

    Muller-Lyda, I.; Javeri, V.; Muller, W.

    2001-01-01

    The treatment of gas-related issues for repositories located in rock salt by GRS and ISTec has followed a strategy which has been developed with increasing complexity and degree of detail in the past. The strategy today clearly indicates the direction to establish a comprehensive safety case and the work that remains to be done. For gas generation mainly long-term aspects are an issue to increase accuracy of predictions. Physical modelling especially for HLW is still incomplete with regard to the coupling of fluid flow with geomechanics, solution/precipitation effects and geochemistry. The appropriate tools to transform the physical models into numerical solutions are at hand in principle but have to be further developed collaterally to the physical modelling. The first full-scale demonstration of safety regarding gas issues in rock salt will have to be provided for the licensing of the Morsleben repository shut-down in the near future. (authors)

  4. The long-term strength and deformation properties of crystalline rock in a high level nuclear waste repository

    International Nuclear Information System (INIS)

    Tuokko, T.

    1990-12-01

    The time-dependent phenomena which can affect the strength and deformation properties of hard crystal line rock are clarified. Suitable measuring methods for field conditions are also summarized. The significance of time is evaluated around a shaft in a high level nuclear waste repository. According to the investigation it is generally held that creep and cyclic fatigue are the most important phenomena. They arise from subcritical crack growth which is most affected by stress intensity, chemical environment, temperature, and microstructure. There are many theoretical models, which can be used to analyse creep and cyclic fatigue, but they are defective in describing the triaxial stress condition and strength criteria. Additionally, the required parameters are often too difficult to determine with adequate accuracy. The joint creep rate depends on the affecting stress regime, on the water conditions, and on the properties of filling material. The acoustic emission method is suited to observe long-term microcrack development in field conditions. The computer program developed by Atomic Energy of Canada Limited (AECL) is used to evaluate the time-dependent de-formation around a main shaft. According to the model the enlargement of the shaft radius by 30 cm takes millions of years. The possible reduction of shaft radius by 3 mm will happen during 200 years. The model is very sensitive to changes in stress state, in the uniaxial compressive strength, and in the stress corrosion index

  5. Aespoe Hard Rock Laboratory. Prototype Repository. Acoustic emission and ultrasonic monitoring results from deposition hole DA3545G01 in the Prototype Repository between October 2007 and March 2008

    International Nuclear Information System (INIS)

    Duckworth, D.; Haycox, J.; Pettitt, W.S.

    2008-12-01

    This report describes results from acoustic emission (AE) and ultrasonic monitoring around a canister deposition hole (DA3545G01) in the Prototype Repository Experiment at SKB's Hard Rock Laboratory (HRL), Sweden. The experiment has been designed to simulate a disposal tunnel in a real deep repository environment for storage of high-level radioactive waste. The test consists of a 90 m long, 5 m diameter subhorizontal tunnel excavated in dioritic granite. The monitoring aims to examine changes in the rock mass caused by an experimental repository environment, in particular due to thermal stresses induced from canister heating and pore pressures induced from tunnel sealing

  6. Aespoe Hard Rock Laboratory. Prototype Repository. Acoustic emission and ultrasonic monitoring results from deposition hole DA3545G01 in the Prototype Repository between October 2007 and March 2008

    Energy Technology Data Exchange (ETDEWEB)

    Duckworth, D.; Haycox, J.; Pettitt, W.S. (Applied Seismology Consultants, Shrewsbury (United Kingdom))

    2008-12-15

    This report describes results from acoustic emission (AE) and ultrasonic monitoring around a canister deposition hole (DA3545G01) in the Prototype Repository Experiment at SKB's Hard Rock Laboratory (HRL), Sweden. The experiment has been designed to simulate a disposal tunnel in a real deep repository environment for storage of high-level radioactive waste. The test consists of a 90 m long, 5 m diameter subhorizontal tunnel excavated in dioritic granite. The monitoring aims to examine changes in the rock mass caused by an experimental repository environment, in particular due to thermal stresses induced from canister heating and pore pressures induced from tunnel sealing.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-11-15

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

  8. Degradation of rocks, through cracking caused by differential thermal expansion, in relation to nuclear waste repositories

    International Nuclear Information System (INIS)

    McLaren, J.R.; Davidge, R.W.; Titchell, I.; Sincock, K.; Bromley, A.

    1982-01-01

    Heating to temperatures up to 500 0 C gives a reduction in Young's modulus and increases in permeability of granitic rocks and it is likely that a major reason is grain boundary cracking. The cracking of grain boundary facets in polycrystalline multiphase materials showing anistropic thermal expansion behaviour is controlled by several microstructural factors in addition to the intrinsic thermal and elastic properties. Of specific interest are the relative orientations of the two grains meeting at the facet, and the size of the facet; these factors thus introduce two statistical aspects to the problem and these are introduced to give quantitative data on crack density versus temperature. The theory is compared with experimental measurements of Young's modulus and permeability for various rocks as a function of temperature. There is good qualitative agreement, and the additional (mainly microstructural) data required for a quantitative comparison are defined. 6 figures, 2 tables

  9. Archive of information about geological samples available for research from the Ohio State University Byrd Polar and Climate Research Center (BPCRC) Polar Rock Repository

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — The Polar Rock Repository (PRR) operated by the Byrd Polar and Climate Research Center (BPCRC) at the Ohio State University is a partner in the Index to Marine and...

  10. Development of low alkaline cementitious grouting materials for a deep geological repository

    International Nuclear Information System (INIS)

    Suzuki, Kenichiro; Miura, Norihiko; Iriya, Keishiro; Kobayashi, Yasushi

    2012-01-01

    In order to reduce uncertainties of long-term safety assessment for a High Level radioactive Waste (HLW) repository system, low alkaline cementitious grouting materials have been studied. The pH of the leachate from the grouting material is targeted to be below 11.0, since the degradation of the bentonite buffer and host rock is limited. The current work focused on the effects of pozzolanic reactions to reduce pH and the development of low alkaline cementitious injection materials in which super-micro ordinary Portland cement (SOPC) was partially replaced by silica fume (SF), micro silica (MS) and fly ash (FA). As it is important to realize how the grouting material will respond to a high injection pressure into the fracture, and in order to understand the penetrability of different low alkaline cement mixes and to observe their flow behavior through the fracture, injection tests were conducted by using a simulated model fracture of 2 m diameter made from parallel plates of acrylic acid resin and stainless steel. Experimental results of the basic properties for selecting suitable materials and that of injecting into a simulated fracture to assess the grouting performance are described

  11. Engineering feasibility for the fabrication and emplacement of cementitious repository materials: results from the EC-ESDRED project

    International Nuclear Information System (INIS)

    Alonso, Maria Cruz; Garcia-Sineriz, Jose Luis

    2012-01-01

    Maria Cruz Alonso of the Spanish National Research Council gave a presentation that summarised relevant findings on cementitious materials from the EC ESDRED (Engineering Studies and Demonstration of Repository Designs) Project. Concrete will be used for different purposes during the construction of geologic repositories for radioactive waste. These purposes include grouting, tunnel and drift lining, and tunnel plugging and sealing. Although some of the concrete may be removed before repository closure, a significant amount of concrete will remain in the repository. An important concern regarding the use of cementitious materials in geologic repositories for HLW and spent fuel is their interaction with the bentonite buffer, backfill material, and the host rock close to the repository near-field. For this reason, the ESDRED project has developed a low-pH concrete formulation as an alternative to standard ordinary Portland cement (OPC) concrete formulations with the aim of reducing the interaction of the cementitious materials with the near-field components. The main functional requirement required in the development of the low-pH material was a pore fluid pH < 11, which is considered acceptable for preventing or reducing the alteration of the bentonite EBS. Other functional requirements considered in the development of the low-pH concrete were: - Hydraulic conductivity. - Mechanical properties. - Durability. - Workability and pumpability. - Slumping. - Peak hydration temperature. - Thermal conductivity. - Use of organic components. - Use of other products. The development of the low-pH concrete involved laboratory work, as well as field testing at the Aespoe underground research laboratory (URL) in Sweden, and in the Grimsel URL and at the Hagerbach site in Switzerland. The ESDRED project demonstrated that low-pH cements can be formulated and used for production of concrete plugs and rock support. OPC can be used as the cement included in low-pH blends, but at least

  12. National survey of crystalline rocks and recommendations of regions to be explored for high-level radioactive waste repository sites

    International Nuclear Information System (INIS)

    Smedes, H.W.

    1983-04-01

    A reconnaissance of the geological literature on large regions of exposed crystalline rocks in the United States provides the basis for evaluating if any of those regions warrant further exploration toward identifying potential sites for development of a high-level radioactive waste repository. The reconnaissance does not serve as a detailed evaluation of regions or of any smaller subunits within the regions. Site performance criteria were selected and applied insofar as a national data base exists, and guidelines were adopted that relate the data to those criteria. The criteria include consideration of size, vertical movements, faulting, earthquakes, seismically induced ground motion, Quaternary volcanic rocks, mineral deposits, high-temperature convective ground-water systems, hydraulic gradients, and erosion. Brief summaries of each major region of exposed crystalline rock, and national maps of relevant data provided the means for applying the guidelines and for recommending regions for further study. It is concluded that there is a reasonable likelihood that geologically suitable repository sites exist in each of the major regions of crystalline rocks. The recommendation is made that further studies first be conducted of the Lake Superior, Northern Appalachian and Adirondack, and the Southern Appalachian Regions. It is believed that those regions could be explored more effectively and suitable sites probably could be found, characterized, verified, and licensed more readily there than in the other regions

  13. National survey of crystalline rocks and recommendations of regions to be explored for high-level radioactive waste repository sites

    Energy Technology Data Exchange (ETDEWEB)

    Smedes, H.W.

    1983-04-01

    A reconnaissance of the geological literature on large regions of exposed crystalline rocks in the United States provides the basis for evaluating if any of those regions warrant further exploration toward identifying potential sites for development of a high-level radioactive waste repository. The reconnaissance does not serve as a detailed evaluation of regions or of any smaller subunits within the regions. Site performance criteria were selected and applied insofar as a national data base exists, and guidelines were adopted that relate the data to those criteria. The criteria include consideration of size, vertical movements, faulting, earthquakes, seismically induced ground motion, Quaternary volcanic rocks, mineral deposits, high-temperature convective ground-water systems, hydraulic gradients, and erosion. Brief summaries of each major region of exposed crystalline rock, and national maps of relevant data provided the means for applying the guidelines and for recommending regions for further study. It is concluded that there is a reasonable likelihood that geologically suitable repository sites exist in each of the major regions of crystalline rocks. The recommendation is made that further studies first be conducted of the Lake Superior, Northern Appalachian and Adirondack, and the Southern Appalachian Regions. It is believed that those regions could be explored more effectively and suitable sites probably could be found, characterized, verified, and licensed more readily there than in the other regions.

  14. The Usability of Rock-Like Materials for Numerical Studies on Rocks

    Science.gov (United States)

    Zengin, Enes; Abiddin Erguler, Zeynal

    2017-04-01

    The approaches of synthetic rock material and mass are widely used by many researchers for understanding the failure behavior of different rocks. In order to model the failure behavior of rock material, researchers take advantageous of different techniques and software. But, the majority of all these instruments are based on distinct element method (DEM). For modeling the failure behavior of rocks, and so to create a fundamental synthetic rock material model, it is required to perform related laboratory experiments for providing strength parameters. In modelling studies, model calibration processes are performed by using parameters of intact rocks such as porosity, grain size, modulus of elasticity and Poisson ratio. In some cases, it can be difficult or even impossible to acquire representative rock samples for laboratory experiments from heavily jointed rock masses and vuggy rocks. Considering this limitation, in this study, it was aimed to investigate the applicability of rock-like material (e.g. concrete) to understand and model the failure behavior of rock materials having complex inherent structures. For this purpose, concrete samples having a mixture of %65 cement dust and %35 water were utilized. Accordingly, intact concrete samples representing rocks were prepared in laboratory conditions and their physical properties such as porosity, pore size and density etc. were determined. In addition, to acquire the mechanical parameters of concrete samples, uniaxial compressive strength (UCS) tests were also performed by simultaneously measuring strain during testing. The measured physical and mechanical properties of these extracted concrete samples were used to create synthetic material and then uniaxial compressive tests were modeled and performed by using two dimensional discontinuum program known as Particle Flow Code (PFC2D). After modeling studies in PFC2D, approximately similar failure mechanism and testing results were achieved from both experimental and

  15. Repository for high level radioactive wastes in Brazil: the importance of geochemical (Micro thermometric) studies and fluid migration in potential host rocks

    International Nuclear Information System (INIS)

    Rios, Francisco Javier; Fuzikawa, Kazuo; Alves, James Vieira; Neves, Jose Marques Correia

    2003-01-01

    A detailed fluid inclusion study of host rocks, is of fundamental importance in the selection of geologically suitable areas for high level nuclear waste repository constructions (HLRW). The LIFM-CDTN is enabled to develop studies that confirm: the presence or not, of corrosive fluid in minerals from host rocks of the repository and the possible presence of micro fractures (and fluid leakage) when these rocks are submitted to high temperatures. These fluid geochemistry studies, with permeability determinations by means of pressurized air injection must be carried out in rocks hosting nuclear waste. Micro fracture determination is of vital importance since many naturally corrosive solutions, present in the mineral rocks, could flow out through these plans affecting the walls of the repository. (author)

  16. Shaft sealing concepts for high-level radioactive waste repositories based on the host-rock options rock salt and clay stone; Schachtverschlusskonzepte fuer zukuenftige Endlager fuer hochradioaktive Abfaelle fuer die Wirtsgesteinsoptionen Steinsalz und Ton

    Energy Technology Data Exchange (ETDEWEB)

    Kudla, Wolfram; Gruner, Matthias [TU Bergakademie Freiberg (Germany). Inst. fuer Erdbau und Spezialtiefbau; Herold, Philipp; Jobmann, Michael [DBE Technology GmbH, Peine (Germany)

    2015-07-01

    Unlike the shaft barriers used for the dry preservation of former mine workings and underground storage sites, shaft seals designed for radioactive-waste repositories must also fulfil additional requirements associated with the design diversity of the sealing system. This diversity makes use of the simple redundancy principle in order to prevent the proliferation of defects. In practice this means combining several sealing elements made from different materials or from materials with different properties. The R and D project, Shaft sealing systems for final repositories for high-level radioactive waste (ELSA) - phase 2: concept design for shaft seals and testing of the functional elements of shaft seals', which was funded by the Federal Ministry for Economic Affairs and Energy (BMWi), set out to investigate potential sealing elements for the two host-rock options rock salt and mudstone. This paper combines the text that the authors presented at the First International Freiberg Shaft Colloquium held at the Freiberg University of Mining and Technology on 01.10.2014 with a presentation on the sealing elements that were investigated as part of the R and D project.

  17. Aespoe Hard Rock Laboratory. Prototype Repository. Analyses of microorganisms, gases and water chemistry in buffer and backfill, 2009

    Energy Technology Data Exchange (ETDEWEB)

    Lydmark, Sara (Microbial Analytics Sweden AB (Sweden))

    2010-09-15

    The Prototype repository is an international project to build and study a full-scale model of the planned Swedish final repository for spent nuclear fuel. The Prototype repository differs from a real storage in that it is drained. For example, this makes the swelling pressure lower in the Prototype repository compared with a real storage. The project is being conducted at the Aespoe Hard Rock Laboratory (HRL) in crystalline rock at a depth of approximately 450 m. A monitoring programme is investigating the evolution of the water chemistry, gas, and microbial activity at the site, and one of the specific aims is to monitor the microbial consumption of oxygen in situ in the Prototype repository. This document describes the results of the analyses of microbes, gases, and chemistry inside and outside the Prototype in 2009. Hydrogen, helium, nitrogen, oxygen, carbon monoxide, carbon dioxide, methane, ethane, and ethene were analysed in the following sampling points in the Prototype repository: KBU10001, KBU10002, KBU10004, KBU10006, KBU10008, KFA01 and KFA04. Where the sampling points in the Prototype delivered pore water, the water was analysed for amount of ATP (i.e., the biovolume), cultivable heterotrophic aerobic bacteria (CHAB), sulphate-reducing bacteria (SRB), methane-oxidizing bacteria (MOB), autotrophic acetogens (AA) and in some cases iron-reducing bacteria (IRB). Cultivation methods were also compared with qPCR molecular techniques to evaluate these before next year's decommission of the Prototype repository. The collected pore water from the Prototype repository was subject to chemistry analysis (as many analyses were conducted as the amount of water allowed). In addition, groundwater from two borehole sections in the rock surrounding the Prototype was analysed regarding its gas composition, microbiology and redox. Chemistry data from a previous investigation of the groundwater outside the Prototype repository were compared with the pore water

  18. Aespoe Hard Rock Laboratory. Prototype Repository. Analyses of microorganisms, gases and water chemistry in buffer and backfill, 2009

    International Nuclear Information System (INIS)

    Lydmark, Sara

    2010-09-01

    The Prototype repository is an international project to build and study a full-scale model of the planned Swedish final repository for spent nuclear fuel. The Prototype repository differs from a real storage in that it is drained. For example, this makes the swelling pressure lower in the Prototype repository compared with a real storage. The project is being conducted at the Aespoe Hard Rock Laboratory (HRL) in crystalline rock at a depth of approximately 450 m. A monitoring programme is investigating the evolution of the water chemistry, gas, and microbial activity at the site, and one of the specific aims is to monitor the microbial consumption of oxygen in situ in the Prototype repository. This document describes the results of the analyses of microbes, gases, and chemistry inside and outside the Prototype in 2009. Hydrogen, helium, nitrogen, oxygen, carbon monoxide, carbon dioxide, methane, ethane, and ethene were analysed in the following sampling points in the Prototype repository: KBU10001, KBU10002, KBU10004, KBU10006, KBU10008, KFA01 and KFA04. Where the sampling points in the Prototype delivered pore water, the water was analysed for amount of ATP (i.e., the biovolume), cultivable heterotrophic aerobic bacteria (CHAB), sulphate-reducing bacteria (SRB), methane-oxidizing bacteria (MOB), autotrophic acetogens (AA) and in some cases iron-reducing bacteria (IRB). Cultivation methods were also compared with qPCR molecular techniques to evaluate these before next year's decommission of the Prototype repository. The collected pore water from the Prototype repository was subject to chemistry analysis (as many analyses were conducted as the amount of water allowed). In addition, groundwater from two borehole sections in the rock surrounding the Prototype was analysed regarding its gas composition, microbiology and redox. Chemistry data from a previous investigation of the groundwater outside the Prototype repository were compared with the pore water chemistry

  19. Distribution coefficient of radionuclides on rocks for performance assessment of high-level radioactive waste repository

    International Nuclear Information System (INIS)

    Shibutani, Tomoki; Shibata, Masahiro; Suyama, Tadahiro

    1999-11-01

    Distribution coefficients of radionuclides on rocks are selected for safety assessment in the 'Second Progress Report on Research and Development for the geological disposal of HLW in Japan (H12 Report)'. The categorized types of rock are granitic rocks (crystalline and acidic rocks), basaltic rocks (crystalline and basic rocks), psammitic rocks (neogene sedimentary (soft)), and tuffaceous-pelitic rocks (pre-neogene sedimentary rocks (hard)). The types of groundwater are FRHP (fresh reducing high-pH), FRLP (fresh reducing low-pH), SRHP (saline reducing high-pH), SRLP (saline reducing low-pH), MRNP (mixing reducing neutral-pH) and FOHP (fresh oxidizing high-pH) groundwater. The elements to be surveyed are Ni, Se, Zr, Nb, Tc, Pd, Sn, Cs, Sm, Pb, Ra, Ac, Th, Pa, U, Np, Pu, Am and Cm. Distribution coefficients are collected from literatures describing batch sorption experimental results, and are selected under consideration of conservativity. (author)

  20. Preliminary environmental assessments of disposal of rock mined during excavation of a federal repository for radioactive waste

    International Nuclear Information System (INIS)

    1977-09-01

    Since the environmental impact of mined rock handling will be dependent not only upon the nature of the material and the way in which it might be disposed but also upon the features of the disposal site area and surroundings, it was necessary to select ''reference environmental locii'' within the regions of geological interest to typify the environmental setting into which the rock would be placed. Reference locii (locations) were developed for consideration of the environmental implications of mined rock from: bedded rock salt from the Salina region, bedded rock salt from the Permian region, dome rock salt from the Gulf Interior region, Pierre shale from the Argillaceous region, granite from the crystalline rock region, volcanic basalt rock from the crystalline ash region, and carbonate rock from the limestone region. Each of these reference locii was examined with respect to those demographic, geographic, physical and ecological attributes which might be impacted by various mined rock disposal alternatives. Alternatives considered included: onsite surface storage, industrial or commercial use, offsite disposal, and environmental blending. Potential impact assessment consists of a qualitative look at the environmental implications of various alternatives for handling the mined rock, given baseline characteristics of an area typified by those represented by the ''reference locus''

  1. Generic repository concept for RBMK-1500 spent nuclear fuel disposal in crystalline rocks in Lithuania

    International Nuclear Information System (INIS)

    Poskas, P.; Brazauskaite, A.; Narkunas, E.; Smaizys, A.; Sirvydas, A.

    2006-01-01

    During 2002-2005 investigations on possibilities to dispose of spent nuclear fuel (SNF) in Lithuania were performed with support of Swedish experts. Disposal concept for RBMK-1500 SNF in crystalline rocks in Lithuania is based on Swedish KBS-3 concept with SNF emplacement into the copper canister with cast iron insert. The bentonite and its mixture with crushed rock are also foreseen as buffer and backfill material. In this paper modelling results on thermal, criticality and other important disposal characteristics for RBMK-1500 SNF fuel emplaced in copper canisters are presented. Based on thermal calculations, the distances between the canisters and between the tunnels were justified. Criticality calculations for the canister with fresh fuel with 2.8 % 235 U enrichment demonstrated that effective neutron multiplication factor k eff values are less than allowable value of 0.95. Dose calculations have shown that total equivalent dose rate from the canister with 50 years stored RBMK-1500 SNF is rather high and is defined mainly by the γ radiation. (author)

  2. The potential for methane hydrate formation in deep repositories of spent nuclear fuel in granitic rocks

    International Nuclear Information System (INIS)

    Tohidi, Bahman; Chapoy, Antonin; Smellie, John; Puigdomenech, Ignasi

    2010-12-01

    The main aim of this work was to establish whether the pertaining pressure and temperature conditions and dissolved gas concentration in groundwater is conducive to gas hydrate formation using a modelling approach. The hydrate stability pressure-temperature zone of dissolved methane in the presence of salt has been obtained through calculations which show that a decrease in the system pressure and/or an increase in salt concentration favours hydrate formation, as both factors reduce equilibrium gas solubility in the aqueous phase. This behaviour is unlike that of the system including a gas phase, where the water phase is always saturated with methane, and hence the methane solubility in water is not a limiting factor. The main conclusion is that hydrate formation is not possible at the reported methane concentrations and water salinities for the Forsmark and Laxemar sites in Sweden and Olkiluoto in Finland. At the highest salinities and methane concentrations encountered, namely ∼0.00073 mole fraction methane and ∼10 mass % NaCl at a depth of 1,000 m in Olkiluoto, Finland, hydrates could form if the system temperatures and pressures are below 2.5 deg C and 60 bar, respectively, i.e. values that are much lower than those prevailing at that depth (∼20 deg C and ∼100 bar, respectively). Furthermore, the calculated results provide the necessary data to estimate the effect of increase in dissolved methane concentration on potential hydrate formation, as well as two phase flow. The available depth dependency of methane concentration at the sites studied in Sweden and Finland was used in another study to estimate the diffusive flow of methane in the rock volumes. These diffusion rates, which are highest at Olkiluoto, indicate that even if the conditions were to become favourable to methane hydrate formation, then it would take several millions of years before a thin layer of hydrates could be formed, a condition which is outside the required period of satisfactory

  3. The potential for methane hydrate formation in deep repositories of spent nuclear fuel in granitic rocks

    Energy Technology Data Exchange (ETDEWEB)

    Tohidi, Bahman; Chapoy, Antonin (Hydrafact Ltd, Inst. of Petroleum Engineering, Heriot-Watt Univ., Edinburgh (United Kingdom)); Smellie, John (Conterra AB, Uppsala (Sweden)); Puigdomenech, Ignasi (Swedish Nuclear Fuel and Waste Management Co., Stockholm (Sweden))

    2010-12-15

    The main aim of this work was to establish whether the pertaining pressure and temperature conditions and dissolved gas concentration in groundwater is conducive to gas hydrate formation using a modelling approach. The hydrate stability pressure-temperature zone of dissolved methane in the presence of salt has been obtained through calculations which show that a decrease in the system pressure and/or an increase in salt concentration favours hydrate formation, as both factors reduce equilibrium gas solubility in the aqueous phase. This behaviour is unlike that of the system including a gas phase, where the water phase is always saturated with methane, and hence the methane solubility in water is not a limiting factor. The main conclusion is that hydrate formation is not possible at the reported methane concentrations and water salinities for the Forsmark and Laxemar sites in Sweden and Olkiluoto in Finland. At the highest salinities and methane concentrations encountered, namely approx0.00073 mole fraction methane and approx10 mass % NaCl at a depth of 1,000 m in Olkiluoto, Finland, hydrates could form if the system temperatures and pressures are below 2.5 deg C and 60 bar, respectively, i.e. values that are much lower than those prevailing at that depth (approx20 deg C and approx100 bar, respectively). Furthermore, the calculated results provide the necessary data to estimate the effect of increase in dissolved methane concentration on potential hydrate formation, as well as two phase flow. The available depth dependency of methane concentration at the sites studied in Sweden and Finland was used in another study to estimate the diffusive flow of methane in the rock volumes. These diffusion rates, which are highest at Olkiluoto, indicate that even if the conditions were to become favourable to methane hydrate formation, then it would take several millions of years before a thin layer of hydrates could be formed, a condition which is outside the required period of

  4. Basic rock properties for the thermo-hydro-mechanical analysis of a high-level radioactive waste repository

    International Nuclear Information System (INIS)

    Kim, Jhin Wung; Kang, Chul Hyung

    1999-04-01

    Deep geological radioactive waste disposal is generally based on the isolation of the waste from the biosphere by multiple barriers. The host rock is one of these barriers which should provide a stable mechanical and chemical environment for the engineered barriers. In the evaluation of the safety of the high-level radioactive waste disposal systems, an important part of the safety analysis is an assessment of the coupling or interaction between thermal, hydrological, and mechanical effects. In order to do this assessment, adequate data on the characteristics of different host rocks are necessary. The properties of the rock and rock discontinuity are very complex and their values vary in a wide range. The accuracy of the result of the assessment depends on the values of these properties used. The present study is an attempt to bring together and condense data for the basic properties of various rock masses, which are needed in the thermo-hydro-mechanical analysis for the deep geological radioactive waste repository. The testing and measurement methods for these basic properties are also presented. Domestic data for deep geological media should be supplemented in the future, due to the insufficiency and the lack of accuracy of the data available at present. (author). 28 refs., 21 figs

  5. Basic rock properties for the thermo-hydro-mechanical analysis of a high-level radioactive waste repository

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jhin Wung; Kang, Chul Hyung

    1999-04-01

    Deep geological radioactive waste disposal is generally based on the isolation of the waste from the biosphere by multiple barriers. The host rock is one of these barriers which should provide a stable mechanical and chemical environment for the engineered barriers. In the evaluation of the safety of the high-level radioactive waste disposal systems, an important part of the safety analysis is an assessment of the coupling or interaction between thermal, hydrological, and mechanical effects. In order to do this assessment, adequate data on the characteristics of different host rocks are necessary. The properties of the rock and rock discontinuity are very complex and their values vary in a wide range. The accuracy of the result of the assessment depends on the values of these properties used. The present study is an attempt to bring together and condense data for the basic properties of various rock masses, which are needed in the thermo-hydro-mechanical analysis for the deep geological radioactive waste repository. The testing and measurement methods for these basic properties are also presented. Domestic data for deep geological media should be supplemented in the future, due to the insufficiency and the lack of accuracy of the data available at present. (author). 28 refs., 21 figs.

  6. Establishment of characterizing parameters of clay as a filling material and coverage for repository

    Energy Technology Data Exchange (ETDEWEB)

    Santos, Daisy M.M. dos; Tello, Cledola Cassia Oliveira de, E-mail: dmms@cdtn.br, E-mail: tellocc@cdtn.br [Centro de Desenvolvimento da Tecnologia Nucelar (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2015-07-01

    The multiple barriers of a repository should be able to provide adequate containment of radionuclides during all the previewed time for the operation and institutional control. One of these barriers is the backfill layer, located between the waste packages and other barriers. Furthermore, after shutting the disposal units with concrete, various materials are used to compose the final coverage of the deposition area. The backfill and the cover layer can be composed of clay or clay mixed with cement, with soil or with rocks. The last layer is a vegetation cover. The selection of the best clay should take into consideration some physical-chemical and mechanical properties. Bentonite is a clay with high absorption capacity, and large volume change in moistening and drying processes, being also effective in the contaminant retention. Additionally, it presents unique properties, such as high swelling potential. Some bentonite characterization works have been developed in the Laboratory of Cementation at CDTN/CNEN (LABCIM/CDTN). The sequence of experiments was: granulometric analysis, moisture, compaction test, hydraulic conductivity and cation exchange capacity. Some initial characterization results are presented and discussed. The paper summarizes these previous studies in order to have the basis for creating a protocol for characterization of a bentonite as a reference material. (author)

  7. Establishment of characterizing parameters of clay as a filling material and coverage for repository

    International Nuclear Information System (INIS)

    Santos, Daisy M.M. dos; Tello, Cledola Cassia Oliveira de

    2015-01-01

    The multiple barriers of a repository should be able to provide adequate containment of radionuclides during all the previewed time for the operation and institutional control. One of these barriers is the backfill layer, located between the waste packages and other barriers. Furthermore, after shutting the disposal units with concrete, various materials are used to compose the final coverage of the deposition area. The backfill and the cover layer can be composed of clay or clay mixed with cement, with soil or with rocks. The last layer is a vegetation cover. The selection of the best clay should take into consideration some physical-chemical and mechanical properties. Bentonite is a clay with high absorption capacity, and large volume change in moistening and drying processes, being also effective in the contaminant retention. Additionally, it presents unique properties, such as high swelling potential. Some bentonite characterization works have been developed in the Laboratory of Cementation at CDTN/CNEN (LABCIM/CDTN). The sequence of experiments was: granulometric analysis, moisture, compaction test, hydraulic conductivity and cation exchange capacity. Some initial characterization results are presented and discussed. The paper summarizes these previous studies in order to have the basis for creating a protocol for characterization of a bentonite as a reference material. (author)

  8. Performance of concrete backfilling materials for shafts and tunnels in rock formations

    International Nuclear Information System (INIS)

    Storer, G.; Mistry, N.; Galliara, J.

    1985-10-01

    This report (Part 2) describes the mathematical modelling studies carried out within a research project into the performance of concrete backfilling materials for shafts and tunnels comprising a hard rock geological disposal repository for High Level, Heat Generating Wastes (HLW/HGW) or Intermediate Level Wastes (ILW) with long lived isotopes. A complementary volume (Part 1) describes laboratory research studies into the development, manufacture and testing of a pre-placed aggregate concrete (PAC). The ongoing objective is to demonstrate that concrete will serve as a beneficial engineered barrier, part of a multi-barrier system, in isolating potentially harmful radionuclides from the biosphere. The report recognises that the backfill cannot be considered in isolation and that there are many interactions between the primary repository elements of host rock, waste and backfill. The interactions considered include mechanical, thermal, creep and moisture movement. Analyses were carried out using the ADINA finite element system, by programmed analytical formulae and using the TEMPOR program (for thermally driven moisture migration in concrete). The emphasis has been directed at establishing basic mathematical approaches to the understanding and quantification of the phenomena involved and applying them to simplified and idealised repository scenarios. The methods devised lay foundations for future work on more defined disposal scenarios. (author)

  9. Self-sealing barriers of sand/bentonite-mixtures in a clay repository. SB-experiment in the Mont Terri Rock Laboratory. Final report

    International Nuclear Information System (INIS)

    Rothfuchs, Tilmann; Czaikowski, Oliver; Hartwig, Lothar; Hellwald, Karsten; Komischke, Michael; Miehe, Ruediger; Zhang, Chun-Liang

    2012-10-01

    Several years ago, GRS performed laboratory investigations on the suitability of clay/mineral mixtures as optimized sealing materials in underground repositories for radioactive wastes /JOC 00/ /MIE 03/. The investigations yielded promising results so that plans were developed for testing the sealing properties of those materials under representative in-situ conditions in the Mont Terri Rock Laboratory (MTRL). The project was proposed to the ''Projekttraeger Wassertechnologie und Entsorgung (PtWT+E)'', and finally launched in January 2003 under the name SB-project (''Self-sealing Barriers of Clay/Mineral Mixtures in a Clay Repository''). The project was divided in two parts, a pre-project running from January 2003 until June 2004 under contract No. 02E9713 /ROT 04/ and the main project running from January 2004 until June 2012 under contract No. 02E9894 with originally PtWT+E, later renamed as PTKA-WTE. In the course of the pre-project it was decided to incorporate the SB main project as a cost shared action of PtWT+E and the European Commission (contract No. FI6W-CT-2004-508851) into the EC Integrated Project ESDRED (Engineering Studies and Demonstrations of Repository Designs) performed by 11 European project partners within the 6th European framework programme. The ESDRED project was terminated prior to the termination of the SB project. Interim results were reported by mid 2009 in two ESDRED reports /DEB09/ /SEI 09/. This report presents the results achieved in the whole SB-project comprising preceding laboratory investigations for the final selection of suited material mixtures, the conduction of mock-up tests in the geotechnical laboratory of GRS in Braunschweig and the execution of in-situ experiments at the MTRL.

  10. An integrated approach to isotopic study of crystalline rock for a high-level waste repository: Area phase

    International Nuclear Information System (INIS)

    Gilbert, L.A.

    1986-01-01

    An integrated approach to assessing isotopic systems in crystalline rock is planned for area phase studies. This approach combines radiogenic isotope systems with petrography in order to characterize potential crystalline repository media. The coeval use of selected isotope systems will minimize the limitations of each method and provide intensive parameters yielding data on alteration timing, secondary mineral formation, temperature history, and radionuclide species migration. Isotope systems will be selected in order to measure differences in sensitivity to thermal disturbances and mobility due to fluid interaction. Comparative evaluation of isotope pair behavior may be used in combination with mineral versus whole-rock dates to provide data on heating and mobilization of alkali elements, lanthanides, and gases, caused by future introduction of waste

  11. Ecohydrological Responses to Diversion of Groundwater: Case Study of a Deep-Rock Repository for Spent Nuclear Fuel in Sweden

    International Nuclear Information System (INIS)

    Werner, Kent; Collinder, Per; Berglund, Sten; Maartensson, Erik

    2013-01-01

    Planning and license applications concerning groundwater diversion in areas containing water-dependent or water-favored habitats must take into account both hydrological effects and associated ecological consequences. There is at present no established methodology to assess such ecohydrological responses. Thus, this paper describes a new stepwise methodology to assess ecohydrological responses to groundwater diversion from, e.g., water-drained pits, shafts, tunnels, and caverns in rock below the groundwater table. The methodology is illustrated using the planned deep-rock repository for spent nuclear fuel at Forsmark in central Sweden as a case study, offering access to a unique hydrological and ecological dataset. The case study demonstrates that results of ecohydrological assessments can provide useful inputs to planning of monitoring programs and mitigation measures in infrastructure projects. As a result of the assessment, artificial water supply to wetlands is planned in order to preserve biological diversity, nature values, and vulnerable species

  12. Excavated rock materials from tunnels for sprayed concrete

    OpenAIRE

    Luong, Judy Yuen Wah; Aarstad, Kari; De Weerdt, Klaartje; Bjøntegaard, Øyvind

    2017-01-01

    Sand extracted from natural resources is widely used in concrete production nowadays. The increase in demand for concrete production has resulted in shortage of natural sand resources, especially in terms of suitable materials for concrete production. At the same time, large amounts of excavated rock materials are and have been generated from tunnelling projects and discarded. Hence, there is an opportunity to use these excavated rock materials as aggregates for concrete production. The chall...

  13. Foreign materials in a deep repository for spent nuclear fuels

    International Nuclear Information System (INIS)

    Jones, C.; Christiansson, Aa.; Wiborgh, M.

    1999-12-01

    The effects of foreign substances introduced into a spent-fuel repository are reviewed. Possible impacts on processes and barrier-functions are examined, and the following areas are identified: Corrosion of the spent-fuel canister through the presence of sulfur and substances that favor microbial growth; impacts on the bentonite properties through the presence of cations as calcium, potassium and iron; radionuclide transport through the presence of complex-formers and surface-active substances

  14. Analysis of hydraulic gradients across the host rock at the proposed Texas Panhandle nuclear-waste repository site

    International Nuclear Information System (INIS)

    Bair, E.S.

    1987-01-01

    Analysis of the direction of ground-water flow across the host rock at the proposed high-level nuclear-waste repository site in Deaf Smith County, Texas, is complicated by vertical and lateral changes in the density of formation fluids in the various hydrogeologic units that overlie and underlie the proposed host rock. Because the concept of hydraulic head is not valid when evaluating vertical hydraulic gradients in a variably-density flow system, other methods were used to determine the direction and magnitude of vertical hydraulic gradients at the proposed site where the specific gravity of formation fluids varies between 1.00 and 1.28. The direction of ground-water flow across the proposed host rock, an 80-foot-thick salt bed in the Lower San Andres Formation, was determined by calculating vertical hydraulic gradients based on formation pressure and fluid density data, and by analysis of pressure-depth diagrams. Based on data from the vicinity of the proposed site, both methods indicate the potential for downflow across the host rock. Downflow or predominantly horizontal flow is considered a favorable prewaste emplacement condition because it prolongs the travel time to the biosphere of any naturally or accidentally released radionuclides

  15. Study on water migration of tunnel surrounding rock in nuclear waste repository based on coupling theory

    International Nuclear Information System (INIS)

    Jiang Zhongming; Zhang Xinmin

    2008-01-01

    Excavation of tunnel changes not only the stresses and deformation of tunnel surrounding rock, but also disturbs the underground water environment in tunnel surrounding rock Water migration happens due to variation of pore water pressure and redistribution. Based on the mechanics of porous media, saturated and unsaturated hydro-mechanical coupling analysis method is employed to study the variation of the stresses, deformation and pore pressure of the surrounding rock. Case study indicates that the excavation of tunnel will induce redistribution of stress and pore water pressure. Redistribution of pore water pressure will seriously affect on evaluation of surrounding rock stability and diffusion of nucleon in the pore water. (authors)

  16. The role of the disturbed rock zone in radioactive waste repository safety and performance assessment. A topical discussion and international overview

    International Nuclear Information System (INIS)

    Winberg, A.

    1991-06-01

    A discussion was presented of the role and relative importance of the disturbed rock zone (DRZ) around the underground openings of a repository for nuclear waste in crystalline rock. The term disturbed rock zone was defined and possible criteria to be sued to distinguish if from undisturbed rock was suggested. The processes decisive for the hydraulic characteristics of the DRZ were discussed. With regard to the integral hydraulic characteristics of the DRZ, the effects of the excavation methodology, stress redistribution, thermal changes, chemical changes and backfill were discussed. A review of in-situ observations of the DRZ was provided. Model analysis where the DRZ has been explicitly or implicitly represented, either from a phenomenological and performance assessment aspect were reviewed. The implications of the disturbed rock zone for the safe performance of a nuclear waste repository were discussed. Conceptual models for the geometry of the DRZ and hydraulic conductivity distribution in the DRZ were suggested. (au) (82 refs.)

  17. Development of a method for the comparison of final repository sites in different host rock formations; Weiterentwicklung einer Methode zum Vergleich von Endlagerstandorten in unterschiedlichen Wirtsgesteinsformationen

    Energy Technology Data Exchange (ETDEWEB)

    Fischer-Appelt, Klaus; Frieling, Gerd; Kock, Ingo; and others

    2017-10-15

    The report on the development of a method for the comparison of final repository sites in different host rock formations covers the following issues: influence of the requirement of retrievability on the methodology, study on the possible extension of the methodology for repository sites with crystalline host rocks: boundary conditions in Germany, final disposal concept for crystalline host rocks, generic extension of the VerSi method, identification, classification and relevance weighting of safety functions, relevance of the safety functions for the crystalline host rock formation, review of the methodological need for changes for crystalline rock sites under low-permeability covering; study on the applicability of the methodology for the determination of site regions for surface exploitation (phase 1).

  18. Summary of United States Geological Survey investigations of fluid-rock-waste reactions in evaporite environments under repository conditions

    International Nuclear Information System (INIS)

    Stewart, D.B.; Jones, B.F.; Roedder, E.; Potter, R.W. II

    1980-01-01

    The interstitial and inclusion fluids contained in rock salt and anhydrite, though present in amounts less than 1 weight per cent, are chemically aggressive and may react with canisters or wastes. The three basic types of fluids are: (1) bitterns residual from saline mineral precipitation including later recrystallization reactions; (2) brines containing residual solutes from the formation of evaporite that have been extensively modified by reactions with contiguous carbonate of clastic rocks; and (3) re-solution brines resulting from secondary dehydration of evaporite minerals or solution of saline minerals by undersaturated infiltrating waters. Fluid composition can indicate that meteoric flow systems have contacted evaporites or that fluids from evaporites have migrated into other formations. The movement of fluids trapped in fluid inclusions in salt from southeast New Mexico is most sensitive to ambient temperature and to inclusion size, although several other factors such as thermal gradient and vapour/liquid ratio are also important. There is no evidence of a threshold temperature for movement of inclusions. Empirical data are given for determining the amount of brine reaching the heat source if the temperature, approximate amount of total dissolved solids, and Ca:Mg ratio in the brine are known. SrCl 2 and CsCl can reach high concentrations in saturated NaCl solutions and greatly depress the liquidus. The possibility that such fluids, if generated, could migrate from a high-level waste repository must be minimized because the fluid would contain its own radiogenic energy source in the first decades after repository closure, thus changing the thermal evolution of the repository from designed values. (author)

  19. Radiation effects on materials in the near-field of a nuclear waste repository. 1997 annual progress report

    International Nuclear Information System (INIS)

    Ewing, R.C.; Wang, L.M.

    1997-01-01

    'Sheet silicates (e.g. micas and clays) are important constituents of a wide variety of geological formations such as granite, basalt, and sandstone. Sheet silicates, particularly clays such as bentonite are common materials in near-field engineered barriers in high-level nuclear waste (HLW) repositories. This is because migration of radionuclides from an underground HLW repository to the geosphere may be significantly reduced by sorption of radionuclides (e.g., Pu, U and Np) onto sheet silicates (e.g., clays and micas) that line the fractures and pores of the rocks along groundwater flowpaths. In addition to surface sorption, it has been suggested that some sheet silicates may also be able to incorporate many radionuclides, such as Cs and Sr, in the inter-layer sites of the sheet structure. However, the ability of the sheet silicates to incorporate radionuclides and retard release and migration of radionuclides may be significantly affected by the near-field radiation due to the decay of fission products and actinides. for example, the unique properties of the sheet structures will be lost completely if the structure becomes amorphous due to irradiation effects. Thus, the study of irradiation effects on sheet-structures, such as structural damage and modification of chemical properties, are critical to the performance assessment of long-term repository behavior.'

  20. Mechanisms and consequences of creep in the nearfield rock of a KBS-3 repository

    International Nuclear Information System (INIS)

    Pusch, R.; Hoekmark, H.

    1992-12-01

    Creep in rock depends on the structure as well as on the stress and temperature. Log time creep is often observed and can be explained on the basis of statistical mechanics. Simple Kelvin behavior can be used as an approximation. The code FLAC is concluded to be useful for predicting creep strain, assuming that the rock obeys the Kelvin law. 22 refs

  1. Evaluation of radiological safety assessment of a repository in a clay rock formation. Evaluacion del comportamiento y de la seguridad de un almacenamiento profundo en arcilla

    Energy Technology Data Exchange (ETDEWEB)

    1999-12-15

    This report presents a comprehensive description of the post-closure radiological safety assessment of a repository for the spent fuel arisings resulting from the Spanish nuclear program excavated in a clay host rock formation. In this report three scenarios have been analysed in detail. The first scenario represents the normal in detail. The first scenario represents the normal evolution of the repository (Reference Scenario); and includes a set of variants to investigate the relative importance of the various repository components and examine the sensitivity of the performance to parameters variations. Two altered scenarios have also been considered: deep well construction and poor sealing of the repository. This document contains a detailed description of the repository system, the methodology adopted for the scenarios generation, the process modelling approach and the results of the consequences analysis. (Author)

  2. Assessment of rock mass quality based on rock quality designation and rock block index. Taking the Borehole BS01 in Beishan HLW disposal repository as example

    International Nuclear Information System (INIS)

    Xu Jian; Wang Ju

    2006-01-01

    Rock mass quality assessment plays an important role in the security for all kinds of large-scale buildings, especially for the underground buildings. In this paper, based on two parameters of RQD and RBI, taking the Borehole BS01 as example, lots of measured data prove that the rock block index can reflect the integrity and corresponding variation of mechanical properties of core from Borehole BS01 to some extent. Meanwhile, the rock mass classification around the Borehole BS01 is given in this paper. Finally, comparison of the results for rock mass assessment between RBI and RQD is made. The research result shows that the rock block index has remarkable significance in engineering and advantages in rock mass quality assessment. (authors)

  3. An assessment of gas impact on geological repository. Methodology and material property of gas migration analysis in engineered barrier system

    International Nuclear Information System (INIS)

    Yamamoto, Mikihiko; Mihara, Morihiro; Ooi, Takao

    2004-01-01

    Gas production in a geological repository has potential hazard, as overpressurisation and enhanced release of radionuclides. Amongst data needed for assessment of gas impact, gas migration properties of engineered barriers, focused on clayey and cementitious material, was evaluated in this report. Gas injection experiments of saturated bentonite sand mixture, mortar and cement paste were carried out. In the experiments, gas entry phenomenon and gas outflow rate were observed for these materials. Based on the experimental results, two-phase flow parameters were evaluated quantitatively. A conventional continuum two-phase flow model, which is only practically used multidimensional multi-phase flow model, was applied to fit the experimental results. The simulation results have been in good agreement with the gas entry time and the outflow flux of gas and water observed in the experiments. It was confirmed that application of the continuum two-phase flow model to gas migration in cementitious materials provides sufficient degree of accuracy for assessment of repository performance. But, for sand bentonite mixture, further extension of basic two-phase flow model is needed especially for effect of stress field. Furthermore, gas migration property of other barrier materials, including rocks, but long-term gas injection test, clarification of influence of chemicals environment and large-scale gas injection test is needed for multi-barrier assessment tool development and their verification. (author)

  4. Prediction of Fracture Behavior in Rock and Rock-like Materials Using Discrete Element Models

    Science.gov (United States)

    Katsaga, T.; Young, P.

    2009-05-01

    The study of fracture initiation and propagation in heterogeneous materials such as rock and rock-like materials are of principal interest in the field of rock mechanics and rock engineering. It is crucial to study and investigate failure prediction and safety measures in civil and mining structures. Our work offers a practical approach to predict fracture behaviour using discrete element models. In this approach, the microstructures of materials are presented through the combination of clusters of bonded particles with different inter-cluster particle and bond properties, and intra-cluster bond properties. The geometry of clusters is transferred from information available from thin sections, computed tomography (CT) images and other visual presentation of the modeled material using customized AutoCAD built-in dialog- based Visual Basic Application. Exact microstructures of the tested sample, including fractures, faults, inclusions and void spaces can be duplicated in the discrete element models. Although the microstructural fabrics of rocks and rock-like structures may have different scale, fracture formation and propagation through these materials are alike and will follow similar mechanics. Synthetic material provides an excellent condition for validating the modelling approaches, as fracture behaviours are known with the well-defined composite's properties. Calibration of the macro-properties of matrix material and inclusions (aggregates), were followed with the overall mechanical material responses calibration by adjusting the interfacial properties. The discrete element model predicted similar fracture propagation features and path as that of the real sample material. The path of the fractures and matrix-inclusion interaction was compared using computed tomography images. Initiation and fracture formation in the model and real material were compared using Acoustic Emission data. Analysing the temporal and spatial evolution of AE events, collected during the

  5. State-of-the-art report on potentially useful materials for sealing nuclear waste repositories

    International Nuclear Information System (INIS)

    Coons, W.; Bergstroem, A.; Gnirk, P.; Gray, M.; Knecht, B.; Pusch, R.; Steadman, J.; Stillborg, B.; Tokonami, Masayasu; Vaajasaari, M.

    1987-06-01

    Seals, including fracture seals, may be used to limit groundwater flow into and away and to limit the release of radionuclides that may be transported by groundwater movement. Seals, if required to achieve repository performance or desirable from a performance standpoint, should have as long service life as possible; the primary means to assure long-term sealing functions is to assure long-term stability of the materials selected for sealing. Seal materials selection and seal design will depend on quantitative sealing criteria; these criteria have not been established and probably cannot be established generically; each repository will have different sealing criteria and individually selected seal materials and designs. In light of the above, however, the priority fracture seal materials, i.e., bentonite grouts and cementitious grouts and their mixtures, will probably be widely applicable and will meet sealing requirements that may be imposed by any of the participants' repository programs. (orig./HP)

  6. Postcards from the past: Archaeological and industrial analogs for deep repository materials

    International Nuclear Information System (INIS)

    Miller, B.; Chapman, N.

    1995-01-01

    Many recent performance assessments of deep geological repositories for radioactive wastes suggest that the engineered barrier system plays the cominant role in reducing releases of radionuclides to the surface environment. There is a considerable impetus to demonstrate the longevity of engineered barrier system components. Although many of the materials are familiar, the requirement for predictable behavior and longevity in a repository is unlike any other requirements of the past. A full appreciation of the acceptability of repository materials can only be reached from a combination of complementary field, laboratory, and natural analog studies. This article discusses analogs from archaelogy and industry. Topics covered include what makes a good analog; long term material behavior (from archeological studies) of metals, glass,cements and concrete, bitumens, and betonite; investigations of radionuclide transport and material interactions. 4 figs., 3 tabs

  7. 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

  8. An overview of a possible approach to calculate rock movements due to earthquakes at Finnish nuclear waste repository sites

    International Nuclear Information System (INIS)

    LaPointe, P.R.; Cladouhos, T.T.

    1999-02-01

    The report outlines a possible approach to estimating rock movements due to earthquakes that may diminish canister safety. The method is based upon an approach developed for studying similar problems in Sweden at three generic Swedish sites. In the first part of the report, the problem of rock movements during earthquakes is described. The second section of the report outlines the approach used to estimate rock movements in Sweden, and discusses how the approach could be adapted to evaluating movements at Finnish repositories. This section also discusses data needs and potential problems in applying the approach in Finland. The next section presents some simple earthquake calculations for the four Finnish sites. These simulations use the discrete fracture network model geometric parameters developed by VTT (Technical Research Centre of Finland) for the use in hydrological calculations. The calculations are not meant for performance assessment purposes for reasons discussed in the report, but are designed to show (1) the importance of fracture size, intensity and orientation on induced displacement magnitudes; (2) the need for additional studies with regards to fracture size and intensity; and (3) the need to resolve issues regarding the role of post-glacial faulting, glacial rebound and tectonic processes in present-day and future earthquakes. (orig.)

  9. Microbial corrosion of metallic materials in a deep nuclear-waste repository

    Directory of Open Access Journals (Sweden)

    Stoulil J.

    2016-06-01

    Full Text Available The study summarises current knowledge on microbial corrosion in a deep nuclear-waste repository. The first part evaluates the general impact of microbial activity on corrosion mechanisms. Especially, the impact of microbial metabolism on the environment and the impact of biofilms on the surface of structure materials were evaluated. The next part focuses on microbial corrosion in a deep nuclear-waste repository. The study aims to suggest the development of the repository environment and in that respect the viability of bacteria, depending on the probable conditions of the environment, such as humidity of bentonite, pressure in compact bentonite, the impact of ionizing radiation, etc. The last part is aimed at possible techniques for microbial corrosion mechanism monitoring in the conditions of a deep repository. Namely, electrochemical and microscopic techniques were discussed.

  10. Digital Repositories of Learning Material as a Support Tool for Knowledge Management and Capacity Building

    International Nuclear Information System (INIS)

    Marmonti, E.

    2016-01-01

    Full text: For some years, digital repositories are emerging as a de facto standard service for storing, preserving and disseminate knowledge: academic, scientific information and, more recently, primary research data of institutions. Some of the digital repositories host also collections of material classified as learning objects; some others are created to manage only learning objects (LO), as the Learning Objects Digital Repositories, or were built to function as learning objects aggregators. The term “learning object” itself is involving different types of structures, organization and complexity. This paper will show how digital repositories, metadata standards and semantic web technologies can be valuable tools for managing educational content, which can contribute to build a learning and knowledge driven organization. (author

  11. Preliminary concepts: materials management in an internationally safeguarded nuclear-waste geologic repository

    International Nuclear Information System (INIS)

    Ostenak, C.A.; Whitty, W.J.; Dietz, R.J.

    1979-11-01

    Preliminary concepts of materials accountability are presented for an internationally safeguarded nuclear-waste geologic repository. A hypothetical reference repository that receives nuclear waste for emplacement in a geologic medium serves to illustrate specific safeguards concepts. Nuclear wastes received at the reference repository derive from prior fuel-cycle operations. Alternative safeguards techniques ranging from item accounting to nondestructive assay and waste characteristics that affect the necessary level of safeguards are examined. Downgrading of safeguards prior to shipment to the repository is recommended whenever possible. The point in the waste cycle where international safeguards may be terminate depends on the fissile content, feasibility of separation, and practicable recoverability of the waste: termination may not be possible if spent fuels are declared as waste

  12. Hydrogen transfer experiments and modelization in clay rocks for radioactive waste deep geological repository

    International Nuclear Information System (INIS)

    Boulin, P.

    2008-10-01

    Gases will be generated by corrosion of high radioactive waste containers in deep geological repositories. A gas phase will be generated. Gas pressure will build up and penetrated the geological formation. If gases do not penetrate the geological barrier efficiently, the pressure build up may create a risk of fracturing and of creation of preferential pathways for radionuclide migration. The present work focuses on Callovo-Oxfordian argillites characterisation. An experiment, designed to measure very low permeabilities, was used with hydrogen/helium and analysed using the Dusty Gas Model. Argillites close to saturation have an accessible porosity to gas transfer that is lower than 0,1% to 1% of the porosity. Analysis of the Knudsen effect suggests that this accessible network should be made of 50 nm to 200 nm diameter pores. The permeabilities values were integrated to an ANDRA operating model. The model showed that the maximum pressure expected near the repository would be 83 bar. (author)

  13. Geotechnical core and rock mass characterization for the UK radioactive waste repository design

    International Nuclear Information System (INIS)

    Rawlings, C.G.; Barton, N.; Loset, F.; Vik, G.; Bhasin, R.K.; Smallwood, A.; Davies, N.

    1996-01-01

    The NGI method of characterizing joints (using JRC, JCS and φ r ) and characterizing rock masses (using the Q-system) have been and are currently being used extensively in geotechnical consultancy projects. One such project recently completed for UK Nirex Ltd included the logging of 8 km of 100-mm-diameter drill core from boreholes up to 2km in depth. Preliminary rock reinforcement designs were derived from the Q-system statistics, which were logged in parallel with JRC, JCS and φ r . The data from the NGI method of characterizing joints and the Q-system for characterizing rock masses have also been used as the basis for UDEC-BB numerical modelling of the proposed cavern excavations for the disposal of solid, low- and intermediate-level radioactive wastes. The purpose of this numerical modelling was to investigate the stability of rock caverns and in particular the rock reinforcement requirements (giving predicted bolt loads and rock deformations), the extent of the disturbed zone (joint shearing and hydraulic aperture) with respect to cavern orientation, the effect of various pillar widths, and the effect of the cavern excavation sequence. (Author)

  14. Transferability of geodata from European to Canadian (Ontario) sedimentary rocks to study gas transport from nuclear wastes repositories

    International Nuclear Information System (INIS)

    Fall, M.; Ghafari, H.; Evgin, E.; Nguyen, T.S.

    2010-01-01

    Document available in extended abstract form only. A deep geological repository (DGR) for low and intermediate level waste in southern Ontario is currently proposed, at a depth of approximately 680 m in an argillaceous limestone formation (Cobourg Limestone) overlain by 200 m of low permeability shale (Ordovician Shale). Significant quantities of gas could be generated in the aforementioned DGR from several processes (e.g., degradation of waste forms, corrosion of waste containers). The accumulation and release of such gases from the repository system may affect a number of processes that influence its long-term safety. Consequently, safety assessments of the proposed DGR need to be supported by a solid understanding of the main mechanisms associated with gas generation and migration and the capability to mathematically model those mechanisms. The development of those mathematical models would usually require the consideration of complex coupled thermo-hydro-mechanical- chemical (THMC) processes. A research program is being conducted in the Department of Civil Engineering of the University of Ottawa in collaboration with the Canadian Nuclear Safety Commission (CNSC) to model the coupled THMC processes associated with gas migration and their impacts on the safety of DGR in southern Ontario. The development and validation of such model as well as the assessment of the impact of gas migration need the acquisition of sufficient amount of (good quality) data on the geomechanical, geochemical, hydraulic, thermal properties of the sedimentary rocks in Southern Ontario as well as relevant gas transport parameters, such as gas entry pressure, Klinkenberg effect, intrinsic permeability, capillary pressure-water saturation relationship. During the past fifteen years, several laboratory and field investigations have been conducted in several countries to acquire geo-data to study and model the THMC processes associated with gas migration in DGR in sedimentary rocks. However

  15. Laboratory investigations into fracture propagation characteristics of rock material

    Science.gov (United States)

    Prasad, B. N. V. Siva; Murthy, V. M. S. R.

    2018-04-01

    After Industrial Revolution, demand of materials for building up structures have increased enormously. Unfortunately, failures of such structures resulted in loss of life and property. Rock is anisotropic and discontinuous in nature with inherent flaws or so-called discontinuities in it. Rock is apparently used for construction in mining, civil, tunnelling, hydropower, geothermal and nuclear sectors [1]. Therefore, the strength of the structure built up considering rockmass as the construction material needs proper technical evaluation during designing stage itself to prevent and predict the scenarios of catastrophic failures due to these inherent fractures [2]. In this study, samples collected from nine different drilling sites have been investigated in laboratory for understanding the fracture propagation characteristics in rock. Rock material properties, ultrasonic velocities through pulse transmission technique and Mode I Fracture Toughness Testing of different variants of Dolomites and Graywackes are determined in laboratory and the resistance of the rock material to catastrophic crack extension or propagation has been determined. Based on the Fracture Toughness values and the rock properties, critical Energy Release Rates have been estimated. However further studies in this direction is to be carried out to understand the fracture propagation characteristics in three-dimensional space.

  16. THM-issues in repository rock. Thermal, mechanical, thermo-mechanical and hydro-mechanical evolution of the rock at the Forsmark and Laxemar sites

    Energy Technology Data Exchange (ETDEWEB)

    Hoekmark, Harald; Loennqvist, Margareta; Faelth, Billy (Clay Technology AB, Lund (Sweden))

    2010-05-15

    The present report addresses aspects of the Thermo-Hydro-Mechanical (THM) evolution of the repository host rock that are of potential importance to the SR-Site safety assessment of a KBS-3 type spent nuclear fuel repository. The report covers the evolution of rock temperatures, rock stresses, pore pressures and fracture transmissivities during the excavation and operational phase, the temperate phase and a glacial cycle on different scales. The glacial cycle is assumed to include a period of pre-glacial permafrost with lowered temperatures and with increased pore pressures in the rock beneath the impermeable permafrost layer. The report also addresses the question of the peak temperature reached during the early temperate phase in the bentonite buffer surrounding the spent fuel canisters. The main text is devoted exclusively to the projected THM evolution of the rock at the Forsmark site in central Sweden. The focus is on the potential for stress-induced failures, i.e. spalling, in the walls of the deposition holes and on changes in the transmissivity of fractures and deformation zones. All analyses are conducted by a combination of numerical tools (3DEC) and analytical solutions. All phases are treated separately and independently of each other, although in reality construction will overlap with heat generation because of the step-by-step excavation/deposition approach with some 50 years between deposition of the first and last canisters. It is demonstrated here that the thermal and thermo-mechanical evolution of the near-field will be independent of heat generated by canisters that were deposited in the past, provided that deposition is made in an orderly fashion, deposition area by deposition area. Peak temperatures and near-field stresses can, consequently, be calculated as if all canisters were deposited simultaneously. The canister and tunnel spacing is specified such that the peak buffer temperature will not exceed 100 deg C in any deposition hole, i.e. not

  17. Programme for repository host rock characterisation in the ONKALO (ReRoC)

    International Nuclear Information System (INIS)

    Aalto, P.; Aaltonen, I.; Kemppainen, K.

    2009-04-01

    The excavation of the ONKALO is now entering the deep bedrock regime, where the ambient rock conditions are representative of those to be found in the vicinity of future deposition tunnels and deposition holes. It is proposed to study the properties of the rock under these conditions using specially-excavated rock rooms, investigation niches or stations and characterisation holes. This report provides an overview of these plans, which are designed to obtain the relevant site-specific knowledge. This report summarises the outstanding issues of the site modelling that are driving the deep rock investigations. It also lists the main long-term safety needs from the site characterisation and presents a short description of the data needs raised by the RSC (Rock Suitability Criteria) programme. The report presents the general characterisation programme of the ONKALO access tunnel. It includes geological mapping in the tunnel and the shafts and investigations in pilot, probe and characterisation hole. It also presents a programme for the pre-grouting hole studies in the shafts. The main aim of the report is to present the experimental studies that are to be carried out in the niches and studies that will be carried out below hydrogeological zone HZ20 at different locations in the tunnel, in order to obtain information on rock properties that are comparable to the rock at the disposal depth. This document provides a detailed discussion of the following experiments: (1) the sulphate reduction experiment at a depth of between 300-350 m to investigate the production, presence and effects of sulphide in the groundwater; (2) the hydrogeological interference experiment at tunnel chainage 3620 or 3748 for a detailed characterisation of connected fracture networks in the rock mass, representative of those in the near field of the deposition holes; (3) the rock matrix diffusion experiment(s) below chainage 4000 to determine the bedrock's essential retention properties for

  18. Background studies in support of a feasibility assessment on the use of copper-base materials for nuclear waste packages in a repository in tuff

    International Nuclear Information System (INIS)

    Van Konynenburg, R.A.; Kundig, K.J.A.; Lyman, W.S.; Prager, M.; Meyers, J.R.; Servi, I.S.

    1990-06-01

    This report combines six work units performed in FY'85--86 by the Copper Development Association and the International Copper Research Association under contract with the University of California. The work includes literature surveys and state-of-the-art summaries on several considerations influencing the feasibility of the use of copper-base materials for fabricating high-level nuclear waste packages for the proposed repository in tuff rock at Yucca Mountain, Nevada. The general conclusion from this work was that copper-base materials are viable candidates for inclusion in the materials selection process for this application. 55 refs., 48 figs., 22 tabs

  19. Background studies in support of a feasibility assessment on the use of copper-base materials for nuclear waste packages in a repository in tuff

    Energy Technology Data Exchange (ETDEWEB)

    Van Konynenburg, R.A. [Lawrence Livermore National Lab., CA (USA); Kundig, K.J.A.; Lyman, W.S.; Prager, M.; Meyers, J.R.; Servi, I.S. [CDA/INCRA Joint Advisory Group, Greenwich, CT (USA)

    1990-06-01

    This report combines six work units performed in FY`85--86 by the Copper Development Association and the International Copper Research Association under contract with the University of California. The work includes literature surveys and state-of-the-art summaries on several considerations influencing the feasibility of the use of copper-base materials for fabricating high-level nuclear waste packages for the proposed repository in tuff rock at Yucca Mountain, Nevada. The general conclusion from this work was that copper-base materials are viable candidates for inclusion in the materials selection process for this application. 55 refs., 48 figs., 22 tabs.

  20. Fluid geochemistry associated associated to rocks: preliminary tests om minerals of granite rocks potentially hostess of radioactive waste repository; Geoquimica de fluidos associados a rochas: testes preliminares em minerais de rochas granitoides potencialmente hospedeiras de repositorios de rejeitos radioativos

    Energy Technology Data Exchange (ETDEWEB)

    Amorim, Lucas E.D.; Rios, Francisco J.; Oliveira, Lucilia A.R. de; Alves, James V.; Fuzikawa, Kazuo; Garcia, Luiz; Ribeiro, Yuri, E-mail: LDAmorim@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil); Matos, Evandro C. de [Industrias Nucleares do Brasil S.A. (INB), Caetite, BA (Brazil)

    2009-07-01

    Fluid inclusions (FI) are micro cavities present on crystals and imprison the mineralizer fluids, and are formed during or posterior to the mineral formation. Those kind of studies are very important for orientation of the engineer barrier projects for this purpose, in order to avoid that the solutions present in the mineral FI can affect the repository walls. This work proposes the development of FI micro compositional studies in the the hostess minerals viewing the contribution for a better understanding of the solution composition present in the metamorphosis granitoid rocks. So, micro thermometric, microchemical and characterization of the material confined in the FI, and the hostess minerals. Great part of the found FI are present in the quartz and plagioclase crystals. The obtained data on the mineral compositions and their inclusions will allow to formulate hypothesis on the process which could occurs at the repository walls, decurrens from of the corrosive character (or not) of the fluids present in the FI, and propose measurements to avoid them

  1. Analysis of the stability of underground high-level nuclear waste repository in discontinuous rock mass using 3DEC

    International Nuclear Information System (INIS)

    Kwon, Sang Ki; Park, Jeong Hwa; Choi, Jong Won; Kang, Chul Hyung

    2001-03-01

    For the safe design of a high-level nuclear waste repository in deep location, it is necessary to confirm the stability of the underground excavations under the high overburden pressure and also to investigate the influence of discontinuities such as fault, fracture zone, and joints. In this study, computer simulations using 3DEC, which is a Distince Element (DEM) code, were carried out for determining important parameters on the stability of the disposal tunnel and deposition holes excavated in 500 m deep granite body. The development of plastic zone and stress and strain distributions were analyzed with various modelling conditions with variation on the parameters including joint numbers, tunnel size, joint properties, rock properties, and stress ratio. Furthermore, the influence of fracture zone, which is located around the underground excavations, on the stability of the excavation was investigated. In this study, the variation of stress and strain distribution due to the variation of fracture zone location, dip, and width was analyzed

  2. Self-sealing of rock fractures. A possibility around the repositories of high-level radioactive wastes

    International Nuclear Information System (INIS)

    Chigira, Masahiro; Nakata, Eiji

    1995-01-01

    To the goal of the safe geological disposal of high-level radioactive wastes (HLW), we must provide long-term confidence for the isolation of HLW in various ways. In particular, groundwater flow, the most likely transport media of radioactive nuclides from HLW, must be restricted around a repository for long time. For that purpose, grouting techniques using cement, bentonite, or other materials have been studied in many countries. In this paper we report the results of a series of experiments on silica precipitation behavior in a flow path with negative temperature gradients in granite and also describe a natural example of hydrothermal alteration of diatomite intruded by andesite. Based on these, we will discuss the possibility of self-sealing around HLW repository. (J.P.N.)

  3. Cost-Effective Cementitious Material Compatible with Yucca Mountain Repository Geochemistry

    Energy Technology Data Exchange (ETDEWEB)

    Dole, LR

    2004-12-17

    The current plans for the Yucca Mountain (YM) repository project (YMP) use steel structures to stabilize the disposal drifts and connecting tunnels that are collectively over 100 kilometers in length. The potential exist to reduce the underground construction cost by 100s of millions of dollars and improve the repository's performance. These economic and engineering goals can be achieved by using the appropriate cementitious materials to build out these tunnels. This report describes the required properties of YM compatible cements and reviews the literature that proves the efficacy of this approach. This report also describes a comprehensive program to develop and test materials for a suite of underground construction technologies.

  4. Main organic materials in a repository for high level radioactive waste

    International Nuclear Information System (INIS)

    Hallbeck, Lotta; Grive, Mireia; Gaona, Xavier; Duro, Lara; Bruno, Jordi

    2007-11-01

    A compilation of the origin and composition of organic material possibly left in a repository is made. Recommendations of precautions and actions for the different material are listed as well. As a brief summary, the different categories of organic material of relevance for the repository are: 1. Microorganisms. Their effect would be mainly a reduction of the redox potential in the initial stages after the repository closure. They may contribute to the depletion of the oxygen entrapped due to the repository construction. This effect would not jeopardize the stability of the repository. If the dominating microorganisms in the anaerobic environment are sulphate-reducing bacteria, oxidation of organic material would lead to formation of HS - . The produced sulphide can corrode copper under anaerobic conditions, if it reaches the canisters. Another effect of microorganisms would be the increase of the complexing capacity of the groundwater due to excreted metabolites. The impact of these compounds is not yet clear, although it will surely not be very important, due to the low amounts of the excreted substances. 2. Materials in the ventilation air. Their effect will probably be a contribution to the maintenance of reducing conditions in the area, although it is likely that this effect will be minimal or negligible. 3. Construction materials. Among them we can highlight organic materials present in concrete, asphalt, bentonite and wood. The most important compounds from the repository safety perspective will be those hydrocarbons from asphalt that may contribute to decreasing the redox potential around the repository, and the products of degradation of cellulose. This last category of compounds may contribute to enhance the complexing capacity of the groundwater around the repository and it is recommended to minimize the amount of cellulose left in the repository. 4. Fuels and engine emissions. No important effects from these organics in the repository are expected

  5. Main organic materials in a repository for high level radioactive waste

    Energy Technology Data Exchange (ETDEWEB)

    Hallbeck, Lotta [Vita vegrandis, Hindaas (Sweden); Grive, Mireia; Gaona, Xavier; Duro, Lara; Bruno, Jordi [Enviros Consulting, Valldoreix, Barcelona (Spain)

    2007-11-15

    A compilation of the origin and composition of organic material possibly left in a repository is made. Recommendations of precautions and actions for the different material are listed as well. As a brief summary, the different categories of organic material of relevance for the repository are: 1. Microorganisms. Their effect would be mainly a reduction of the redox potential in the initial stages after the repository closure. They may contribute to the depletion of the oxygen entrapped due to the repository construction. This effect would not jeopardize the stability of the repository. If the dominating microorganisms in the anaerobic environment are sulphate-reducing bacteria, oxidation of organic material would lead to formation of HS{sup -}. The produced sulphide can corrode copper under anaerobic conditions, if it reaches the canisters. Another effect of microorganisms would be the increase of the complexing capacity of the groundwater due to excreted metabolites. The impact of these compounds is not yet clear, although it will surely not be very important, due to the low amounts of the excreted substances. 2. Materials in the ventilation air. Their effect will probably be a contribution to the maintenance of reducing conditions in the area, although it is likely that this effect will be minimal or negligible. 3. Construction materials. Among them we can highlight organic materials present in concrete, asphalt, bentonite and wood. The most important compounds from the repository safety perspective will be those hydrocarbons from asphalt that may contribute to decreasing the redox potential around the repository, and the products of degradation of cellulose. This last category of compounds may contribute to enhance the complexing capacity of the groundwater around the repository and it is recommended to minimize the amount of cellulose left in the repository. 4. Fuels and engine emissions. No important effects from these organics in the repository are expected

  6. Redox front formation in an uplifting sedimentary rock sequence: An analogue for redox-controlling processes in the geosphere around deep geological repositories for radioactive waste

    International Nuclear Information System (INIS)

    Yoshida, H.; Metcalfe, R.; Yamamoto, K.; Murakami, Y.; Hoshii, D.; Kanekiyo, A.; Naganuma, T.; Hayashi, T.

    2008-01-01

    Subsurface redox fronts control the mobilization and fixation of many trace elements, including potential pollutants such as certain radionuclides. Any safety assessment for a deep geological repository for radioactive wastes needs to take into account adequately the long-term redox processes in the geosphere surrounding the repository. To build confidence in understanding these processes, a redox front in a reduced siliceous sedimentary rock distributed in an uplifting area in Japan has been studied in detail. Geochemical analyses show increased concentrations of Fe and trace elements, including rare earth elements (REEs), at the redox front, even though concentrations of reduced rock matrix constituents show little change. Detailed SEM observations revealed that fossilized microorganisms composed of amorphous granules made exclusively of Fe and Si occur in the rock's pore space. Microbial 16S rDNA analysis suggests that there is presently a zonation of different bacterial groups within the redox band, and bacterial zonation played an important role in the concentration of Fe-oxyhydroxides at the redox front. These water-rock-microbe interactions can be considered analogous to the processes occurring in the redox fronts that would develop around geological repositories for radioactive waste. Once formed, the Fe-oxyhydroxides within such a front would be preserved even after reducing conditions resume following repository closure

  7. Redox front formation in an uplifting sedimentary rock sequence: An analogue for redox-controlling processes in the geosphere around deep geological repositories for radioactive waste

    Energy Technology Data Exchange (ETDEWEB)

    Yoshida, H. [Nagoya University Museum, Material Research Section, Furocho, Nagoya 464-8602 (Japan)], E-mail: dora@num.nagoya-u.ac.jp; Metcalfe, R. [Quintessa Japan, Queen' s Tower A7-707, Minatomirai, Yokohama 220-6007 (Japan); Yamamoto, K. [Nagoya University Museum, Material Research Section, Furocho, Nagoya 464-8602 (Japan); Murakami, Y. [Japan Atomic Energy Agency (JAEA), Tono Geoscience Centre (Japan); Hoshii, D.; Kanekiyo, A.; Naganuma, T. [Hiroshima University, Higashi Hiroshima, Kagamiyama 1-4-4 (Japan); Hayashi, T. [Asahi University, Department of Dental Pharmacology, Hozumi, Gifu (Japan)

    2008-08-15

    Subsurface redox fronts control the mobilization and fixation of many trace elements, including potential pollutants such as certain radionuclides. Any safety assessment for a deep geological repository for radioactive wastes needs to take into account adequately the long-term redox processes in the geosphere surrounding the repository. To build confidence in understanding these processes, a redox front in a reduced siliceous sedimentary rock distributed in an uplifting area in Japan has been studied in detail. Geochemical analyses show increased concentrations of Fe and trace elements, including rare earth elements (REEs), at the redox front, even though concentrations of reduced rock matrix constituents show little change. Detailed SEM observations revealed that fossilized microorganisms composed of amorphous granules made exclusively of Fe and Si occur in the rock's pore space. Microbial 16S rDNA analysis suggests that there is presently a zonation of different bacterial groups within the redox band, and bacterial zonation played an important role in the concentration of Fe-oxyhydroxides at the redox front. These water-rock-microbe interactions can be considered analogous to the processes occurring in the redox fronts that would develop around geological repositories for radioactive waste. Once formed, the Fe-oxyhydroxides within such a front would be preserved even after reducing conditions resume following repository closure.

  8. A theoretical and numerical consideration of rock mass behaviour under thermal loading of radioactive waste repository

    International Nuclear Information System (INIS)

    Reivinen, M.; Freund, J.; Eloranta, E.

    1996-08-01

    The aim of the study is to model the geodynamic response of a ground rock block under horizontal stresses and also consider the thermal fields and deformations, especially on the ground surface, caused by the heat produced by nuclear waste. (12 refs.)

  9. The diffusion of some radionuclides in local rocks collected from potential repository in Syria

    International Nuclear Information System (INIS)

    Othman, Ibrahim; Abou Jamous, Jamal

    1992-07-01

    Diffusion factor was estimated for 137 Cs in local rocks marl, limestone, and basalt. Slab activity measuring was constructed. Factors affecting the 137 Cs diffusion has been studied. These are dynamic state of water, length of contacting time and the concentration of radioisotope. (author). 9 refs., 12 figs., 6 tabs

  10. Preliminary constitutive properties for salt and nonsalt rocks from four potential repository sites

    International Nuclear Information System (INIS)

    Pfeifle, T.W.; Mellegard, K.D.; Senseny, P.E.

    1983-07-01

    Results are presented from laboratory strength and creep tests performed on salt and nonsalt specimens from the Richton Dome in Mississippi, the Vacherie Dome in Louisiana, the Permian Basin in Texas, and the Paradox Basin in Utah. The constititive properties obtained for salt are the elastic moduli and the failure envelope at 24 0 C and parameter values for the exponential-time creep law. Some additional data are presented to indicate how the elastic moduli and strength change with temperature. The nonsalt constitutive properties reported are the elastic moduli, the unconfined compressive strength and the tensile strength at 24 0 C. The properties given in this report will be used in subsequent numerical simulations that will provide information to assist in the screening and selection of site locations for a nuclear waste repository and to assist in the repository design at the selected site. The matrix of tests performed is the minimum effort required to obtain these constitutive properties. The preliminary values obtained will be supplemented by additional testing for sites that are selected for further investigation

  11. What requirements does the KBS-3 repository make on the host rock? Geoscientific suitability indicators and criteria for siting and site evaluation

    International Nuclear Information System (INIS)

    Andersson, Johan; Stroem, Anders; Svemar, Christer; Almen, Karl-Erik; Ericsson, Lars O.

    2000-04-01

    This report gives an account of what requirements are made on the rock, what conditions in the rock are advantageous and how the fulfilment of requirements and preferences is to be judged prior to the selection of sites for a site investigation and during a site investigation. The conclusions and results of the report are based on the knowledge and experience acquired by SKB over many years of research and development. The results, and particularly the stipulated criteria, apply to a repository for spent fuel of the KBS-3 type, i.e. a repository where the fuel is contained in copper canisters embedded in bentonite clay at a depth of 400 - 700 m in the Swedish crystalline basement. The report analyzes how the rock's different geological conditions, mechanical, thermal, hydrogeological, chemical and transport properties influence the functions of the deep repository, and whether it is possible to determine requirements and preferences regarding the influence of these properties. Where possible, these requirements or preferences have then been translated into requirements or preferences regarding the individual properties. Criteria are formulated that are based on the quantities that can be measured or estimated at the relevant stage of the investigation. The following requirements are made on the rock: The rock in the repository's deposition zone may not have any ore potential. Regional plastic shear zones shall be avoided if it cannot be demonstrated that the properties of the zone do not deviate from those of the rest of the rock. There may, however, be so-called 'tectonic lenses' near regional plastic shear zones where the bedrock is homogeneous and relatively unaffected. Deposition tunnels and deposition holes for canisters may not pass through or be positioned too close to major regional and major local fracture zones. Deposition holes may not intersect identified local minor fracture zones. The rock's strength, fracture geometry and initial stresses may not be

  12. Superhard nanophase cutter materials for rock drilling applications; FINAL

    International Nuclear Information System (INIS)

    Voronov, O.; Tompa, G.; Sadangi, R.; Kear, B.; Wilson, C.; Yan, P.

    2000-01-01

    The Low Pressure-High Temperature (LPHT) System has been developed for sintering of nanophase cutter and anvil materials. Microstructured and nanostructured cutters were sintered and studied for rock drilling applications. The WC/Co anvils were sintered and used for development of High Pressure-High Temperature (HPHT) Systems. Binderless diamond and superhard nanophase cutter materials were manufactured with help of HPHT Systems. The diamond materials were studied for rock machining and drilling applications. Binderless Polycrystalline Diamonds (BPCD) have high thermal stability and can be used in geothermal drilling of hard rock formations. Nanophase Polycrystalline Diamonds (NPCD) are under study in precision machining of optical lenses. Triphasic Diamond/Carbide/Metal Composites (TDCC) will be commercialized in drilling and machining applications

  13. Evaluation of possible host rocks for China's high level radioactive waste repository and the progress in site characterization at the Beishan potential site in NW China's Gansu province

    International Nuclear Information System (INIS)

    Wang Ju; Jin Yuanxin; Chen Zhangru; Chen Weiming; Wang Wenguang

    2000-01-01

    Evaluation of possible host rocks for China's high level radioactive waste repository is summarized in this paper. The distribution and characteristics of granite, tuff, clay stone, salt and loess in China are described, while maps showing the distribution of host rocks are presented. Because of the wide distribution, large scale, good heat conductivity and suitable mechanical properties, granite is considered as the most potential host rock. Some granite bodies distributed in NW China, SW China, South China and Inner Mongolia have been selected as potential areas. Detailed site characterization at Beishan area, Gansu Province NW China is in progress

  14. Effects of repository environment on diffusion behavior of radionuclides in buffer materials

    International Nuclear Information System (INIS)

    Kozaki, Tamotsu; Sato, Seichi

    2004-03-01

    Compacted bentonite is considered as a candidate buffer material in the geological disposal of high-level radioactive waste. An important function of the compacted bentonite is to retard the transport of radionuclides from waste forms to the surrounding host rock after degradation of an overpack. Therefore, diffusion behavior of radionuclides in the compacted bentonite has been extensively studied by many researchers for the performance assessments of the geological disposal. However, diffusion mechanism of radionuclides in the bentonite cannot be fully understood, and most experimental data have been obtained at room temperature for the bentonite saturated with low salinity water, which would disagree often with real repository conditions. In this study, therefore, apparent diffusion coefficients were determined at various diffusion temperatures for chloride ions in Na-montmorillonite samples saturated with NaCl solution of high salinity. Activation energies for the apparent diffusion were also obtained from the temperature dependence of the diffusion coefficients at different salinity. As the salinity increased, the apparent diffusion coefficients of chloride ions in montmorillonite were found to increase slightly. On the other hand, the activation energies for the chloride diffusion were found to be almost constant (approximately 12 kJ mol -1 ) and less than that in free water (17.4 kJ mol -1 ). Effects of salinity on diffusion behavior of radionuclides in montmorillonite were discussed from the viewpoints of microstructure of montmorillonite and distribution of ions in the montmorillonite. As a result, the diffusion behavior of sodium ions could be explained by the changes of the predominant diffusion process among pore water diffusion, surface diffusion, and interlayer diffusion that could be caused by the increase of salinity. (author)

  15. Geochemical simulation of the evolution of granitic rocks and clay minerals submitted to a temperature increase in the vicinity of a repository for spent nuclear fuel

    International Nuclear Information System (INIS)

    Fritz, B.; Kam, M.; Tardy, Y.

    1984-07-01

    The alteration of a granitic rock around a repository for spent nuclear fuel has been simulated considering the effect of an increase of temperature due to this kind of induced geothermal system. The results of the simulation have been interpreted in terms of mass transfer and volumic consequences. The alteration proceeds by dissolution of minerals (with an increase of the volumes of fissures and cracks) and precipitation of secondary miminerals as calcite and clay minerals particularly (with a decrease of the porosity). The increase of the temperature from 10 degrees C to about 100 degrees C will favour the alteration of the granitic rock around the repository by the solution filling the porosity. The rock is characterized by a very low fissure porosity and a consequent very low water velocity. This too, favours intense water rock interactions and production of secondary clays and the total possible mass transfer will decrease the porosity. A combination of these thermodynamic mass balance calculations with a kinetic approach of mineral dissolutions gives a first attempt to calibrate the modelling in the time scale: the decrease of porosity can be roughly estimated between 2 and 20% for 100,000 years. The particular problem of Na-bentonites behaviour in the proximate vicinity of the repository has been studied too. One must distinguish between two types of clay-water interactions: -within the rock around the repository, Na-bentonites should evolute with illitization in slighltly open system with low clay/water ratios, -within the repository itself, the clay reacts in a closed system for a long time with high clay/water ratios and a self-buffering effect should maintain the bentonite type. This chemical buffering effect is a positive point for the use of this clay as chemical barrier. (Author)

  16. Numerical modeling of rock stresses within a basaltic nuclear waste repository. Final report

    International Nuclear Information System (INIS)

    Hardy, M.P.; Hocking, G.

    1978-01-01

    The modeling undertaken during this project incorporated a wide range of problems that impact the design of the waste repository. Interaction of groundwater, heat and stress were considered on a regional scale, whereas on the room and canister scale thermo-mechanical analyses were undertaken. In the Phase II report, preliminary guidelines for waste densities were established based primarily on short-term stress criteria required to maintain stability during the retrievability period. Additional analyses are required to evaluate the effect of joints, borehole linings, room support and ventilation on these preliminary waste loading densities. The regional analyses did not indicate any adverse effect that could control the allowable waste loading densities. However, further refinements of geologic structure, hydrologic models, seismicity and possible induced seismicity are required before firm estimates of the loading densities can be made

  17. Understanding large scale groundwater flow in fractured crystalline rocks to aid in repository siting

    International Nuclear Information System (INIS)

    Davison, C.; Brown, A.; Gascoyne, M.; Stevenson, D.; Ophori, D.

    2000-01-01

    Atomic Energy of Canada Limited (AECL) conducted a ten-year long groundwater flow study of a 1050 km 2 region of fractured crystalline rock in southeastern Manitoba to illustrate how an understanding of large scale groundwater flow can be used to assist in selecting a hydraulically favourable location for the deep geological disposal of nuclear fuel waste. The study involved extensive field investigations that included the drilling testing, sampling and monitoring of twenty deep boreholes distributed at detailed study areas across the region. The surface and borehole geotechnical investigations were used to construct a conceptual model of the main litho-structural features that controlled groundwater flow through the crystalline rocks of the region. Eighty-three large fracture zones and other spatial domains of moderately fractured and sparsely fractured rocks were represented in a finite element model of the area to simulate regional groundwater flow. The groundwater flow model was calibrated to match the observed groundwater recharge rate and the hydraulic heads measured in the network of deep boreholes. Particle tracking was used to determine the pathways and travel times from different depths in the velocity field of the calibrated groundwater flow model. The results were used to identify locations in the regional flow field that maximize the time it takes for groundwater to travel to surface discharge areas through long, slow groundwater pathways. One of these locations was chosen as a good hypothetical location for situating a nuclear fuel waste disposal vault at 750 m depth. (authors)

  18. A THM stress-strain framework for modelling the performance of argillaceous materials in deep repositories for radioactive waste

    International Nuclear Information System (INIS)

    Laloui, L.; Francois, B.

    2007-01-01

    In the scenarios for deep, geological nuclear-waste repositories, clayey soils will be hydrated, heated, cooled and dried. The numerical modelling of these mechanical processes is a key issue. Performance assessment of deep repositories for heat-generating radioactive waste would benefit from improvements in mechanical stress-strain constitutive modelling of the coupled thermo-hydro-mechanical behaviour. The presented framework allows progress in understanding the most involved phenomena relevant to nuclear-waste repositories and their coupled nature. It could be used both in the design and in the performance assessment of repositories. It may be applied to disposal in clay formations and to hard-rock repositories where artificially compacted clay is to be used as buffer and backfill. Such a constitutive framework may help in understanding some unexplained or controversial behaviours and in defining experimental programmes to answer key questions. (author)

  19. The influence of organic materials on the near field of an intermediate level radioactive waste repository

    International Nuclear Information System (INIS)

    Wilkins, J.D.

    1988-01-01

    The influence of organic materials which are present in some intermediate level wastes on the chemistry of the near field of a radioactive waste repository is discussed. Particular attention is given to the possible formation of water soluble complexing agents as a result of the radiation field and chemical conditions. The present state of the research is reviewed. (author)

  20. Aespoe Hard Rock Laboratory. Prototype Repository. Acoustic emission and ultrasonic monitoring results from deposition hole DA3545G01 in the Prototype Repository between April 2008 and September 2008

    International Nuclear Information System (INIS)

    Duckworth, D.; Haycox, J.; Pettitt, W.S.

    2009-03-01

    This report describes results from acoustic emission (AE) and ultrasonic monitoring around a canister deposition hole (DA3545G01) in the Prototype Repository Experiment at SKB's Hard Rock Laboratory (HRL), Sweden. The monitoring aims to examine changes in the rock mass caused by an experimental repository environment, in particular due to thermal stresses induced from canister heating and pore pressures induced from tunnel sealing. Monitoring of this volume has previously been performed during excavation [Pettitt et al., 1999], and during stages of canister heating and tunnel pressurisation [Haycox et al., 2005a and 2005b; Haycox et al., 2006a and 2006b; Zolezzi et al., 2007 and Duckworth et al., 2008]. Further information on this monitoring can be found in Appendix I. This report covers the period between 1st April 2008 and 30th September 2008 and is the seventh instalment of the 6-monthly processing and interpretation of the results from the experiment

  1. Aespoe Hard Rock Laboratory. Prototype Repository. Acoustic emission and ultrasonic monitoring results from deposition hole DA3545G01 in the Prototype Repository between April 2008 and September 2008

    Energy Technology Data Exchange (ETDEWEB)

    Duckworth, D.; Haycox, J.; Pettitt, W.S. (Applied Seismology Consultants, Shrewsbury (United Kingdom))

    2009-03-15

    This report describes results from acoustic emission (AE) and ultrasonic monitoring around a canister deposition hole (DA3545G01) in the Prototype Repository Experiment at SKB's Hard Rock Laboratory (HRL), Sweden. The monitoring aims to examine changes in the rock mass caused by an experimental repository environment, in particular due to thermal stresses induced from canister heating and pore pressures induced from tunnel sealing. Monitoring of this volume has previously been performed during excavation [Pettitt et al., 1999], and during stages of canister heating and tunnel pressurisation [Haycox et al., 2005a and 2005b; Haycox et al., 2006a and 2006b; Zolezzi et al., 2007 and Duckworth et al., 2008]. Further information on this monitoring can be found in Appendix I. This report covers the period between 1st April 2008 and 30th September 2008 and is the seventh instalment of the 6-monthly processing and interpretation of the results from the experiment.

  2. Proposed format and content of environmental reports for deep geologic terminal repositories for radioactive material

    International Nuclear Information System (INIS)

    Carrell, D.J.; Jones, G.L.

    1978-01-01

    As the Nuclear Regulatory Commission has not yet issued a format guide for the preparation of an environmental impact statement for radioactive waste repositories, Rockwell Hanford operations has developed an annotated outline which will serve as the basis for the environmental evaluation activities until replaced by an appropriate NRC regulatory guide. According to the outline, the applicant should summarize the major environmental effects that are expected to occur during the construction, operation, and terminal isolation phases of the radioactive material repository. Compare these environmental effects with the possible effect of continued use of interim storage facilities. Unless unforeseen environmental effects become apparent, the summary should be a positive statement indicating that the short-term environmental effects are outweighed by the long-term benefits of the repository

  3. Deep repository - Engineered barrier system. Erosion and sealing processes in tunnel backfill materials investigated in laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Sanden, Torbjoern; Boergesson, Lennart; Dueck, Ann; Goudarzi, Reza; Loennqvist, Margareta (Clay Technology AB, Lund (Sweden))

    2008-12-15

    SKB in Sweden and Posiva in Finland are developing and plan to implement similar disposal concepts for the final disposal of spent nuclear fuel. Co-operation and joint development work between Posiva and SKB with the overall objective to develop backfill concepts and techniques for sealing and closure of the repository have been going on for several years. The investigation described in this report is intended to acquire more knowledge regarding the behavior of some of the candidate backfilling materials. Blocks made of three different materials (Friedland clay, Asha 230 or a bentonite/ballast 30/70 mixture) as well as different bentonite pellets have been examined. The backfill materials will be exposed to an environment simulating that in a tunnel, with high relative humidity and water inflow from the rock. The processes and properties investigated are: 1. Erosion properties of blocks and pellets (Friedland blocks, MX-80 pellets, Cebogel QSE pellets, Minelco and Friedland granules). 2. Displacements of blocks after emplacement in a deposition drift (Blocks of Friedland, Asha 230 and Mixture 30/70). 3. The ability of these materials to seal a leaking in-situ cast plug cement/rock but also other fractures in the rock (MX-80 pellets). 4. The self healing ability after a piping scenario (Blocks of Friedland, Asha 230 Mixture 30/70 and also MX-80 pellets). 5. Swelling and cracking of the compacted backfill blocks caused by relative humidity. The erosion properties of Friedland blocks were also investigated in Phase 2 of the joint SKBPosiva project 'Backfilling and Closure of the Deep Repository, BACLO, which included laboratory scale experiments. In this phase of the project (3) some completing tests were performed with new blocks produced for different field tests. These blocks had a lower density than intended and this has an influence on the erosion properties measured. The erosion properties of MX-80 pellets were also investigated earlier in the project but

  4. Study of materials for using at waste layer in repositories

    International Nuclear Information System (INIS)

    Amaral, Andre F.; Tello, Cledola C.O. de

    2009-01-01

    This research has an objective to characterize Brazilian clays and to implant a data base containing the information obtained form tests and suppliers. Such information will allow to buy and and to select optimum material for its utilization in the stuffing layer. Brazilian suppliers were contacted for obtaining information and samples, the various clays were tested and these tests comprehend the following: identification of the mineral constituents, determination of the compaction curve as function of the humidity, hydraulic conductivity, humidity and organic material contents, cationic exchange capacity, specific surface, and etc

  5. Long term thermo-hydro-mechanical interaction behavior study of the saturated, discontinuous granitic rock mass around the radwaste repository using a steady state flow algorithm

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jhin Wung; Bae, Dae Suk; Kang, Chul Hyung; Choi, Jong Won [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2002-02-01

    The objective of the present study is to understand the long term (500 years) thermo-hydro-mechanical interaction behavior of the 500 m depth underground radwaste repository in the saturated, discontinuous granitic rock mass using a steady state flow algorithm. The numerical model includes a saturated granitic rock mass with joints around the repository and a 45 .deg. C fault passing through the tunnel roof-wall intersection, and a canister with PWR spent fuels surrounded by the compacted bentonite and mixed-bentonite. Barton-Bandis joint constitutive model from the UDEC code is used for the joints. For the hydraulic analysis, a steady state flow algorithm is used for the groundwater flow through the rock joints. For the thermal analysis, heat transfer is modeled as isotropic conduction and heat decays exponentially with time. The results show that the variations of the hydraulic aperture, hydraulic conductivity, normal stress, normal displacements, and shear displacements of the joints are high in the vicinity of the repository and stay fairly constant on the region away from the repository. 14 refs., 15 figs., 11 tabs. (Author)

  6. Installation of depository for radioactive material in rocks

    International Nuclear Information System (INIS)

    Bergman, S.G.A.; Sagefors, K.I.; Aakesson, B.Aa.

    1985-01-01

    The rock outside the depository has a hollow space which is filled by elastoplastic material possible to deform. The solid body of the depository has a central vertical shaft and concentric vertical outer shafts. Between the shafts there are vertically oriented layers with tunnels for storage of waste. The tunnels slope down from the central shaft. (G.B.)

  7. What requirements does the KBS-3 repository make on the host rock? Geoscientific suitability indicators and criteria for siting and site evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Andersson, Johan [Golder Grundteknik AB (Sweden); Stroem, Anders; Svemar, Christer [Swedish Nuclear Fuel and Waste Management Co., Stockholm (Sweden); Almen, Karl-Erik [KEA Geo-Konsult AB, Naessjoe (Sweden); Ericsson, Lars O. [Chalmers University of Technology, Goeteborg (Sweden)

    2000-04-01

    This report gives an account of what requirements are made on the rock, what conditions in the rock are advantageous and how the fulfilment of requirements and preferences is to be judged prior to the selection of sites for a site investigation and during a site investigation. The conclusions and results of the report are based on the knowledge and experience acquired by SKB over many years of research and development. The results, and particularly the stipulated criteria, apply to a repository for spent fuel of the KBS-3 type, i.e. a repository where the fuel is contained in copper canisters embedded in bentonite clay at a depth of 400 - 700 m in the Swedish crystalline basement. The report analyzes how the rock's different geological conditions, mechanical, thermal, hydrogeological, chemical and transport properties influence the functions of the deep repository, and whether it is possible to determine requirements and preferences regarding the influence of these properties. Where possible, these requirements or preferences have then been translated into requirements or preferences regarding the individual properties. Criteria are formulated that are based on the quantities that can be measured or estimated at the relevant stage of the investigation. The following requirements are made on the rock: The rock in the repository's deposition zone may not have any ore potential. Regional plastic shear zones shall be avoided if it cannot be demonstrated that the properties of the zone do not deviate from those of the rest of the rock. There may, however, be so-called 'tectonic lenses' near regional plastic shear zones where the bedrock is homogeneous and relatively unaffected. Deposition tunnels and deposition holes for canisters may not pass through or be positioned too close to major regional and major local fracture zones. Deposition holes may not intersect identified local minor fracture zones. The rock's strength, fracture geometry and

  8. Use of safety analysis to site comfirmation procedure in case of hard rock repository

    International Nuclear Information System (INIS)

    Peltonen, E.K.

    1984-02-01

    The role of safety analysis in a confirmation procedure of a candidate disposal site of radioactive wastes is discussed. Items dealt with include principle reasons and practical goals of the use of safety analysis, methodology of safety analysis and assessment, as well as usefulness and adequacy of the present safety analysis. Safety analysis is a tool, which enables one to estimate quantitatively the possible radiological impacts from the disposal. The results can be compared with the criteria and the suitability conclusions drawn. Because of its systems analytical nature safety analysis is an effective method to reveal, what are the most important factors of the disposal system and the most critical site characteristics inside the lumped parameters often provided by the experimental site investigation methods. Furthermore it gives information on the accuracy needs of different site properties. This can be utilized to judge whether the quality and quantity of the measurements for the characterization are sufficient as well as to guide the further site investigations. A more practical discussion regarding the applicability of the use of safety analysis is presented by an example concerning the assessment of a Finnish candidate site for low- and intermediate-level radioactive waste repository. (author)

  9. Repositories; Repositorios

    Energy Technology Data Exchange (ETDEWEB)

    Freire, Carolina Braccini; Tello, Cledola Cassia Oliveira de [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)]. E-mails: cbf@cdtn.br; tellocc@cdtn.br

    2007-11-15

    The use of the nuclear energy is increasing in all areas. Then the radioactive waste management is in continuous development to comply the national and international established requirements. The final objective is to assure that it will not have any contamination of the public or the environmental, and that the exposition doses will be lower than the radiological protection limits. The multi barrier concept for the repository is internationally recognized. Among the repository types, the most used are: near surface, geological formations and of deposition in rock cavities. This article explains the concept and the types of repository and gives some examples of them. (author)

  10. Control of materials harmful to water in the German Konrad repository - 16125

    International Nuclear Information System (INIS)

    Kugel, Karin; Brennecke, Peter; Steyer, Stephan; Gruendler, Detlef; Boetsch, Wilma; Haider, Claudia

    2009-01-01

    In order to avoid a pollution of the near surface ground water during the post closure phase of the Konrad repository the acceptable amount of material harmful to water in the radioactive waste is restricted. For this purpose the KONRAD plan approval order includes waste requirements referring to the German water law ('water law permission'). In a first part of this contribution the water law permission for the KONRAD repository is introduced. This permission contains a list of materials harmful to water with the respective limitations in mass and many instructions and proposals regarding the registering and balancing of these materials as well as quality assurance aspects. The second part deals with the implementation of the water law permission in the waste acceptance criteria. The waste producer has to describe his waste in a standardized way with respect to the material composition. The operator of the repository has to check this description and to register and balance the materials and substances harmful to water. This procedure is based on a standardized list of materials and a list of containers. In the third part quality control measures used for the proof of the compliance with the acceptance criteria (with respect to the water law permission) are described. In particular objective of the quality control, possible quality control options and acceptable margins are dealt with. (authors)

  11. Rock stability considerations for siting and constructing a KBS-3 repository. Based on experiences from Aespoe HRL, AECL's URL, tunnelling and mining

    Energy Technology Data Exchange (ETDEWEB)

    Martin, C.D. [Univ. of Alberta, Edmonton (Canada); Christiansson, Rolf [Swedish Nuclear Fuel and Waste Management Co., Stockholm (Sweden); Soederhaell, J. [VBB VIAK AB, Stockholm (Sweden)

    2001-12-01

    Over the past 25 years the international nuclear community has carried out extensive research into the deep geological disposal of nuclear waste in hard rocks. In two cases this research has resulted in the construction of dedicated underground research facilities: SKB's Aespoe Hard Rock Laboratory, Sweden and AECL's Underground Research Laboratory, Canada. Both laboratories are located in hard rocks considered representative of the Fennoscandian and Canadian Shields, respectively. This report is intended to synthesize the important rock mechanics findings from these research programs. In particular the application of these finding to assessing the stability of underground openings. As such the report draws heavily on the published results from the SKB's ZEDEX Experiment in Sweden and AECL's Mine- by Experiment in Canada. The objectives of this report are to: 1. Describe, using the current state of knowledge, the role rock engineering can play in siting and constructing a KBS-3 repository. 2. Define the key rock mechanics parameters that should be determined in order to facilitate repository siting and construction. 3. Discuss possible construction issues, linked to rock stability, that may arise during the excavation of the underground openings of a KBS-3 repository. 4. Form a reference document for the rock stability analysis that has to be carried out as a part of the design works parallel to the site investigations. While there is no unique or single rock mechanics property or condition that would render the performance of a nuclear waste repository unacceptable, certain conditions can be treated as negative factors. Outlined below are major rock mechanics issues that should be addressed during the siting, construction and closure of a nuclear waste repository in Sweden in hard crystalline rock. During the site investigations phase, rock mechanics information will be predominately gathered from examination and testing of the rock core and

  12. Rock stability considerations for siting and constructing a KBS-3 repository. Based on experiences from Aespoe HRL, AECL's URL, tunnelling and mining

    International Nuclear Information System (INIS)

    Martin, C.D.; Christiansson, Rolf; Soederhaell, J.

    2001-12-01

    Over the past 25 years the international nuclear community has carried out extensive research into the deep geological disposal of nuclear waste in hard rocks. In two cases this research has resulted in the construction of dedicated underground research facilities: SKB's Aespoe Hard Rock Laboratory, Sweden and AECL's Underground Research Laboratory, Canada. Both laboratories are located in hard rocks considered representative of the Fennoscandian and Canadian Shields, respectively. This report is intended to synthesize the important rock mechanics findings from these research programs. In particular the application of these finding to assessing the stability of underground openings. As such the report draws heavily on the published results from the SKB's ZEDEX Experiment in Sweden and AECL's Mine- by Experiment in Canada. The objectives of this report are to: 1. Describe, using the current state of knowledge, the role rock engineering can play in siting and constructing a KBS-3 repository. 2. Define the key rock mechanics parameters that should be determined in order to facilitate repository siting and construction. 3. Discuss possible construction issues, linked to rock stability, that may arise during the excavation of the underground openings of a KBS-3 repository. 4. Form a reference document for the rock stability analysis that has to be carried out as a part of the design works parallel to the site investigations. While there is no unique or single rock mechanics property or condition that would render the performance of a nuclear waste repository unacceptable, certain conditions can be treated as negative factors. Outlined below are major rock mechanics issues that should be addressed during the siting, construction and closure of a nuclear waste repository in Sweden in hard crystalline rock. During the site investigations phase, rock mechanics information will be predominately gathered from examination and testing of the rock core and mapping of the

  13. Materials Degradation Issues in the U.S. High-Level Nuclear Waste Repository

    Energy Technology Data Exchange (ETDEWEB)

    K.G. Mon; F. Hua

    2005-04-12

    This paper reviews the state-of-the-art understanding of the degradation processes by the Yucca Mountain Project (YMP) with focus on interaction between the in-drift environmental conditions and long-term materials degradation of waste packages and drip shields within the repository system during the first 10,000-years after repository closure. This paper provides an overview of the degradation of the waste packages and drip shields in the repository after permanent closure of the facility. The degradation modes discussed in this paper include aging and phase instability, dry oxidation, general and localized corrosion, stress corrosion cracking, and hydrogen induced cracking of Alloy 22 and titanium alloys. The effects of microbial activity and radiation on the degradation of Alloy 22 and titanium alloys are also discussed. Further, for titanium alloys, the effects of fluorides, bromides, and galvanic coupling to less noble metals are considered. It is concluded that the materials and design adopted will provide sufficient safety margins for at least 10,000-years after repository closure.

  14. Materials Degradation Issues in the U.S. High-Level Nuclear Waste Repository

    International Nuclear Information System (INIS)

    Mon, K.G.; Hua, F.

    2005-01-01

    This paper reviews the state-of-the-art understanding of the degradation processes by the Yucca Mountain Project (YMP) with focus on interaction between the in-drift environmental conditions and long-term materials degradation of waste packages and drip shields within the repository system during the first 10,000-years after repository closure. This paper provides an overview of the degradation of the waste packages and drip shields in the repository after permanent closure of the facility. The degradation modes discussed in this paper include aging and phase instability, dry oxidation, general and localized corrosion, stress corrosion cracking, and hydrogen induced cracking of Alloy 22 and titanium alloys. The effects of microbial activity and radiation on the degradation of Alloy 22 and titanium alloys are also discussed. Further, for titanium alloys, the effects of fluorides, bromides, and galvanic coupling to less noble metals are considered. It is concluded that the materials and design adopted will provide sufficient safety margins for at least 10,000-years after repository closure

  15. Educational and Community Outreach Efforts by the United States Polar Rock Repository during the International Polar Year

    Science.gov (United States)

    Grunow, A.; Codispoti, J. E.

    2010-12-01

    The US Polar Rock Repository (USPRR) houses more than 19,000 rock samples from polar regions and these samples are made available to the scientific, educational and museum community. The USPRR has been active in promoting polar earth science to educational and community groups. During the past year, outreach efforts reached over 12,000 people. The USPRR outreach involve tours of the facility, school presentations, online laboratory exercises, working with the Columbus Metro Parks, teaching at summer camps, teaching special geology field assignments at the middle school level, as well as offering an ‘Antarctic Rock Box’ that contains representative samples of the three types of rocks, minerals, fossils, and books and activities about geology and Antarctica. The rock box activities have been designed and reviewed by educators and scientists to use as an educational supplement to the Earth Science course of study. The activities have been designed around the Academic Content Standards: k-12 Science manual published by the Ohio Department of Education to ensure that the activities and topics are focused on those mandated by the state of Ohio. The USPRR website has a Virtual Web Antarctic Expedition with many activities for Middle to High School age students. The students learn about how to plan a field season, safety techniques, how to make a remote field camp, identify what equipment is needed, learn about the different transportation choices, weather issues, understanding GPS, etc. Educational and community networks have been built in part, by directly contacting individuals at an institution and partnering with them on educational outreach. The institutions have been very interested in doing this because it brings scientists to the classroom and to the public. This type of outreach has also served as an opening for children to consider possible career choices in science that they may not have considered before. In many of the presentations, a female geologist

  16. Development of rock bolt grout and shotcrete for rock support and corrosion of steel in low-pH cementitious materials

    Energy Technology Data Exchange (ETDEWEB)

    Boden, Anders (Vattenfall Power Consultant AB, Vaellingby (Sweden)); Pettersson, Stig (Swedish Nuclear Fuel and Waste Management Co., Stockholm (Sweden))

    2011-04-15

    It is foreseen that cementitious products will be utilized in the construction of the final repository. The use of conventional cementitious material creates pulses in the magnitude of pH 12.13 in the leachates and release alkalis. Such a high pH is detrimental mainly to impairment of bentonite functioning, but also to possibly enhanced dissolution of spent fuel and alteration of fracture filling materials. It also complicates the safety analysis of the repository, as the effect of a high pH-plume should be considered in the evaluation. As no reliable pH-plume models exist, the use of products giving a pH below 11 in the leachates facilitates the safety analysis, although limiting the amount of low-pH cement is recommended. In earlier studies it was found that shotcreting, standard casting and rock bolting with low-pH cement (pH . 11 in the leachate) should be possible without any major development work. This report summarizes the results of development work done during 2008 and 2009 in the fields of low-pH rock bolt grout, low-pH shotcrete and steel corrosion in low-pH concrete. Development of low-pH rock bolt grout mixes and laboratory testing of the selected grout was followed by installation of twenty rock bolts for rock support at Aspo HRL using the chosen low-pH grout. The operation was successful and the bolts and grout are subject to follow up the next ten years. Low-pH shotcrete for rock support was initially developed within the ESDRED project, which was an Integrated Project within the European Commission sixth framework for research and technological development. ESDRED is an abbreviation for Engineering Studies and Demonstrations of Repository Designs. ESDRED was executed from 1st February 2004 to 31st January 2009. The development of the mix design described in this report was based on the results from ESDRED. After laboratory testing of the chosen mix, it was field tested in niche NASA 0408A at Aspo HRL. Further, some areas in the TASS-tunnel were

  17. Development of rock bolt grout and shotcrete for rock support and corrosion of steel in low-pH cementitious materials

    International Nuclear Information System (INIS)

    Boden, Anders; Pettersson, Stig

    2011-04-01

    It is foreseen that cementitious products will be utilized in the construction of the final repository. The use of conventional cementitious material creates pulses in the magnitude of pH 12.13 in the leachates and release alkalis. Such a high pH is detrimental mainly to impairment of bentonite functioning, but also to possibly enhanced dissolution of spent fuel and alteration of fracture filling materials. It also complicates the safety analysis of the repository, as the effect of a high pH-plume should be considered in the evaluation. As no reliable pH-plume models exist, the use of products giving a pH below 11 in the leachates facilitates the safety analysis, although limiting the amount of low-pH cement is recommended. In earlier studies it was found that shotcreting, standard casting and rock bolting with low-pH cement (pH . 11 in the leachate) should be possible without any major development work. This report summarizes the results of development work done during 2008 and 2009 in the fields of low-pH rock bolt grout, low-pH shotcrete and steel corrosion in low-pH concrete. Development of low-pH rock bolt grout mixes and laboratory testing of the selected grout was followed by installation of twenty rock bolts for rock support at Aspo HRL using the chosen low-pH grout. The operation was successful and the bolts and grout are subject to follow up the next ten years. Low-pH shotcrete for rock support was initially developed within the ESDRED project, which was an Integrated Project within the European Commission sixth framework for research and technological development. ESDRED is an abbreviation for Engineering Studies and Demonstrations of Repository Designs. ESDRED was executed from 1st February 2004 to 31st January 2009. The development of the mix design described in this report was based on the results from ESDRED. After laboratory testing of the chosen mix, it was field tested in niche NASA 0408A at Aspo HRL. Further, some areas in the TASS-tunnel were

  18. Probabilistic methods as a tool aiding dimensioning drift and shaft seals for a repository in rock salt

    Energy Technology Data Exchange (ETDEWEB)

    Roehlig, Klaus-Juergen; Plischke, Elmar; Li, Xiaoshuo [TU Clausthal, Clausthal-Zellerfeld (Germany). Inst. of Disposal Research (IELF)

    2015-07-01

    For repositories in rock salt, demonstrating the integrity of drift and shaft seals is an indispensable part of the long-term safety case. In this study, probabilistic methods are applied to assess the fictitious abutment length for a shaft seal and the effective permeability of a drift seal (dam), i.e. the integral entity for the whole structure including contact zone and damaged salt zone. For the seal permeability, the question arises how to derive it based on permeability measurements with a limited number of samples due to cost restrictions. Furthermore, it is of interest which conclusions can be derived regarding the minimum length of drift seals if the failure probability should be smaller than e.g. 10{sup -4}. Based on numerical experiments it was demonstrated that small-scale measurements can be upscale using known averaging methods. This suggests that dimensioning can be carried out based on cautions average estimates and the required reliability statement (e.g. about a failure probability smaller than e.g. 10{sup -4}) can be derived for realistic dam lengths. However, due to the limited amount of data available there are remaining uncertainties concerning the underlying model assumptions.

  19. Probabilistic methods as a tool aiding dimensioning drift and shaft seals for a repository in rock salt

    International Nuclear Information System (INIS)

    Roehlig, Klaus-Juergen; Plischke, Elmar; Li, Xiaoshuo

    2015-01-01

    For repositories in rock salt, demonstrating the integrity of drift and shaft seals is an indispensable part of the long-term safety case. In this study, probabilistic methods are applied to assess the fictitious abutment length for a shaft seal and the effective permeability of a drift seal (dam), i.e. the integral entity for the whole structure including contact zone and damaged salt zone. For the seal permeability, the question arises how to derive it based on permeability measurements with a limited number of samples due to cost restrictions. Furthermore, it is of interest which conclusions can be derived regarding the minimum length of drift seals if the failure probability should be smaller than e.g. 10 -4 . Based on numerical experiments it was demonstrated that small-scale measurements can be upscale using known averaging methods. This suggests that dimensioning can be carried out based on cautions average estimates and the required reliability statement (e.g. about a failure probability smaller than e.g. 10 -4 ) can be derived for realistic dam lengths. However, due to the limited amount of data available there are remaining uncertainties concerning the underlying model assumptions.

  20. Sandstone uranium deposits of Meghalaya: natural analogues for radionuclide migration and backfill material in geological repository for high level radioactive waste disposal

    International Nuclear Information System (INIS)

    Bajpai, R.K.; Narayan, P.K.

    2008-01-01

    Sandstone uranium deposits serve as potential natural analogue to demonstrate safety offered by geological media against possible release of nuclear waste from their confinement and migration towards biosphere. In this study, available database on geochemical aspects of Domisiat uranium deposit of Meghalaya has been evaluated to highlight the behavior of radionuclides of concern over long term in a geological repository. Constituents like actinides (U and Th), fission products and RE elements are adequately retained in clays and organic matters associated with these sandstone deposits. The study also highlights the possibility of utilization of lean ore discarded during mining and milling as backfill material in far field areas and optimizing near field buffers/backfills in a geological repository located in granitic rocks in depth range of 400-500m. (author)

  1. Modeling of Coupled Thermo-Hydro-Mechanical-Chemical Processes for Bentonite in a Clay-rock Repository for Heat-generating Nuclear Waste

    Science.gov (United States)

    Xu, H.; Rutqvist, J.; Zheng, L.; Birkholzer, J. T.

    2016-12-01

    Engineered Barrier Systems (EBS) that include a bentonite-based buffer are designed to isolate the high-level radioactive waste emplaced in tunnels in deep geological formations. The heat emanated from the waste can drive the moisture flow transport and induce strongly coupled Thermal (T), Hydrological (H), Mechanical (M) and Chemical (C) processes within the bentonite buffer and may also impact the evolution of the excavation disturbed zone and the sealing between the buffer and walls of an emplacement tunnel The flow and contaminant transport potential along the disturbed zone can be minimized by backfilling the tunnels with bentonite, if it provides enough swelling stress when hydrated by the host rock. The swelling capability of clay minerals within the bentonite is important for sealing gaps between bentonite block, and between the EBS and the surrounding host rock. However, a high temperature could result in chemical alteration of bentonite-based buffer and backfill materials through illitization, which may compromise the function of these EBS components by reducing their plasticity and capability to swell under wetting. Therefore, an adequate THMC coupling scheme is required to understand and to predict the changes of bentonite for identifying whether EBS bentonite can sustain higher temperatures. More comprehensive links between chemistry and mechanics, taking advantage of the framework provided by a dual-structure model, named Barcelona Expansive Model (BExM), was implemented in TOUGHREACT-FLAC3D and is used to simulate the response of EBS bentonite in in clay formation for a generic case. The current work is to evaluate the chemical changes in EBS bentonite and the effects on the bentonite swelling stress under high temperature. This work sheds light on the interaction between THMC processes, evaluates the potential deterioration of EBS bentonite and supports the decision making in the design of a nuclear waste repository in light of the maximum allowance

  2. The JRC Nanomaterials Repository: A unique facility providing representative test materials for nanoEHS research.

    Science.gov (United States)

    Totaro, Sara; Cotogno, Giulio; Rasmussen, Kirsten; Pianella, Francesca; Roncaglia, Marco; Olsson, Heidi; Riego Sintes, Juan M; Crutzen, Hugues P

    2016-11-01

    The European Commission has established a Nanomaterials Repository that hosts industrially manufactured nanomaterials that are distributed world-wide for safety testing of nanomaterials. In a first instance these materials were tested in the OECD Testing Programme. They have then also been tested in several EU funded research projects. The JRC Repository of Nanomaterials has thus developed into serving the global scientific community active in the nanoEHS (regulatory) research. The unique Repository facility is a state-of-the-art installation that allows customised sub-sampling under the safest possible conditions, with traceable final sample vials distributed world-wide for research purposes. This paper describes the design of the Repository to perform a semi-automated subsampling procedure, offering high degree of flexibility and precision in the preparation of NM vials for customers, while guaranteeing the safety of the operators, and environmental protection. The JRC nanomaterials are representative for part of the world NMs market. Their wide use world-wide facilitates the generation of comparable and reliable experimental results and datasets in (regulatory) research by the scientific community, ultimately supporting the further development of the OECD regulatory test guidelines. Copyright © 2016 The Authors. Published by Elsevier Inc. All rights reserved.

  3. A comparison study of single and double layer repositories for high level radioactive wastes within a saturated and discontinuous granitic rock mass

    International Nuclear Information System (INIS)

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

    2004-02-01

    The present study is to analyze and compare a long term thermohydro mechanical interaction behavior of a single layer and a double layer repository for high level radioactive wastes within a saturated and discontinuous granitic rock mass, and then to contribute this understanding to the development of a Korean disposal concept. The model includes a saturated and discontinuous granitic rock mass, PWR spent nuclear fuel in a disposal canister surrounded by compacted bentonite inside a deposition hole, and mixed bentonite backfilled in the rest of the space within a repository cavern. It is assumed that two joint sets exist within the model. Joint set 1 includes joints of 56 .deg. dip angle, spaced at 20 m, and joint set 2 is in the perpendicular direction to joint set 1 and includes joints of .deg. dip angle, spaced at 20 m. The two dimensional distinct element code, UDEC is used for the analysis. To understand the joint behavior adjacent to the repository cavern, Barton-Bandis joint model is used. Effect of the decay heat from PWR spent fuels on the repository model has been analyzed, and a steady state flow algorithm is used for the hydraulic analysis

  4. Stress corrosion cracking tests on high-level-waste container materials in simulated tuff repository environments

    International Nuclear Information System (INIS)

    Abraham, T.; Jain, H.; Soo, P.

    1986-06-01

    Types 304L, 316L, and 321 austenitic stainless steel and Incoloy 825 are being considered as candidate container materials for emplacing high-level waste in a tuff repository. The stress corrosion cracking susceptibility of these materials under simulated tuff repository conditions was evaluated by using the notched C-ring method. The tests were conducted in boiling synthetic groundwater as well as in the steam/air phase above the boiling solutions. All specimens were in contact with crushed Topopah Spring tuff. The investigation showed that microcracks are frequently observed after testing as a result of stress corrosion cracking or intergranular attack. Results showing changes in water chemistry during test are also presented

  5. Requirements of actual final repository concepts for different host rock formations. Final report; Anforderungen an aktuelle Endlagerkonzepte fuer unterschiedliche Wirtsgesteinsformationen. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Fass, Thorsten; Hartwig-Thurat, Eva; Krischer, Angelika; Lambers, Ludger; Larue, Juergen; Uhlmann, Stephan; Weyand, Torben

    2017-08-15

    In the frame of the research project the basic requirements and technical safety specifications with respect to the retrievability of stored radioactive wastes for the different final repository concepts based on the host rock formations occurring in Germany are presented. Existing international disposal concepts for clay/claystone, granite and salt are described and compared to the actual German regulatory requirements. The safety engineering relations between stock piling and possible retrieval are described and evaluated.

  6. 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

  7. Development of low-pH cementitious materials for HLRW repositories. Resistance against ground waters aggression

    OpenAIRE

    Garcia Calvo, Jose Luis; Hidalgo, A.; Fernandez Luco, L.; Alonso Alonso, Maria Cruz

    2010-01-01

    One of the most accepted engineering construction concepts of underground repositories for high radioactive waste considers the use of low-pH cementitious materials. This paper deals with the design of those based on Ordinary Portland Cements with high contents of silica fume and/or fly ashes that modify most of the concrete “standard” properties, the pore fluid composition and the microstructure of the hydrated products. Their resistance to long-term groundwater aggression is also evaluated....

  8. Assessing microbiologically induced corrosion of waste package materials in the Yucca Mountain repository

    Energy Technology Data Exchange (ETDEWEB)

    Horn, J. M., LLNL

    1998-01-01

    The contribution of bacterial activities to corrosion of nuclear waste package materials must be determined to predict the adequacy of containment for a potential nuclear waste repository at Yucca Mountain (YM), NV. The program to evaluate potential microbially induced corrosion (MIC) of candidate waste container materials includes characterization of bacteria in the post-construction YM environment, determination of their required growth conditions and growth rates, quantitative assessment of the biochemical contribution to metal corrosion, and evaluation of overall MIC rates on candidate waste package materials.

  9. Research on swelling clays and bitumen as sealing materials for radioactive waste repositories

    International Nuclear Information System (INIS)

    Allison, J.A.; Wilson, J.; Mawditt, J.M.; Hurt, J.C.

    1991-01-01

    This report describes a programme of research to investigate the performance of composite seals incorporating adjacent blocks of swelling clay and bitumen. It is shown that the interaction of the materials can promote a self-sealing mechanism which prevents water penetration, even when defects are present in the bitumen layer. A review of the swelling properties of highly compacted bentonite and magnesium oxide is presented, and the characteristic sealing properties of bituminous materials are described. On the basis of this review, it is concluded that bentonite is the preferred candidate material for use in composite clay/bitumen seals for intermediate-level radioactive waste repositories. However, it is thought that magnesium oxide may have other sealing applications for high-level waste repositories. A programme of laboratory experiments is described in which relevant swelling and intrusion properties of highly compacted bentonite blocks and the annealing characteristics of oxidised and hard-grade industrial bitumens are examined. The results of composite sealing experiments involving different water penetration routes are reported, and factors governing the mechanism of self-sealing are described. The validation of the sealing concept at a laboratory scale indicates that composite bentonite/bitumen seals could form highly effective barriers for the containment of radioactive wastes. Accordingly, recommendations are made concerning the development of the research, including the implementation of full-scale demonstration experiments to simulate conditions in an underground repository. 13 tabs., 41 figs., 62 refs

  10. Repository design

    Energy Technology Data Exchange (ETDEWEB)

    John, C M

    1982-01-01

    Various technical issues of radioactive waste design are addressed in this paper. Two approaches to repository design considered herein are: (1) design to minimize the disturbance of the hot rock; and (2) designs that intentionally modify the hot rock to insure better containment of the wastes. The latter designs range from construction of a highly impermeable barrier around a spherical cavern to creating a matrix of tunnels and boreholes to form a cage within which the hydraulic pressure is nearly constant. Examples of these design alternatives are described in some detail. It is concluded that proposed designs for repositories illustrate that performance criteria considered acceptable for such facilities can be met by appropriate site selection and repository engineering. With these technically feasible design concepts, it is also felt that socioeconomic and institutional issues can be better resolved. (BLM)

  11. Projected environmental impacts of radioactive material transportation to the first US repository site

    International Nuclear Information System (INIS)

    Neuhauser, K.S.; Cashwell, J.W.; Reardon, P.C.; Ostmeyer, R.M.; McNair, G.W.

    1986-01-01

    This paper discusses the relative national environmental impacts of transporting nuclear wastes to each of the nine candidate repository sites in the United States. Several of the potential sites are closely clustered and, for the purpose of distance and routing calculations, are treated as a single location. These are: Cypress Creek Dome and Richton Dome in Mississippi (Gulf Interior Region), Deaf Smith County and Swisher County sites in Texas (Permian Basin), and Davis Canyon and Lavender Canyon site in Utah (Paradox Basin). The remaining sites are: Vacherie Dome, Louisiana; Yucca Mountain, Nevada; and Hanford Reservation, Washington. For compatibility with both the repository system authorized by the NWPA and with the MRS option, two separate scenarios were analyzed. In belief, they are (1) shipment of spent fuel and high-level wastes (HLW) directly from waste generators to a repository (Reference Case) and (2) shipment of spent fuel to a Monitored Retrievable Storage (MRS) facility and then to a repository. Between 17 and 38 truck accident fatalities, between 1.4 and 7.7 rail accident fatalities, and between 0.22 and 12 radiological health effects can be expected to occur as a result of radioactive material transportation during the 26-year operating period of the first repository. During the same period in the United States, about 65,000 total deaths from truck accidents and about 32,000 total deaths from rail accidents would occur; also an estimated 58,300 cancer fatalities are predicted to occur in the United States during a 26-year period from exposure to background radiation alone (not including medical and other manmade sources). The risks reported here are upper limits and are small by comparison with the ''natural background'' of risks of the same type. 3 refs., 6 tabs

  12. Rock-welding materials for deep borehole nuclear waste disposal.

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Pin [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Wang, Yifeng [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Rodriguez, Mark A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Brady, Patrick Vane [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Swift, Peter N. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-11-01

    The concept of deep borehole nuclear waste disposal has recently been proposed. Effective sealing of a borehole after waste emplacement is generally required. In a high temperature disposal mode, the sealing function will be fulfilled by melting the ambient granitic rock with waste decay heat or an external heating source, creating a melt that will encapsulate waste containers or plug a portion of the borehole above a stack of the containers. However, there are certain drawbacks associated with natural materials, such as high melting temperatures, slow crystallization kinetics, the resulting sealing materials generally being porous with low mechanical strength, insufficient adhesion to waste container surface, and lack of flexibility for engineering controls. Here we show that natural granitic materials can be purposefully engineered through chemical modifications to enhance the sealing capability of the materials for deep borehole disposal. This work systematically explores the effect of chemical modification and crystallinity (amorphous vs. crystalline) on the melting and crystallization processes of a granitic rock system. A number of engineered granitic materials have been obtained that have decreased melting points, enhanced viscous densification, and accelerated recrystallization rates without compromising the mechanical integrity of the materials.

  13. Elaboration of protocol for characterization of clay as a filling material and coverage for surface repository

    International Nuclear Information System (INIS)

    Santos, Daisy Mary Marchezini dos

    2017-01-01

    The nuclear energy in its various applications generates wastes that must be properly treated. The Radioactive Waste Management covers operations since generation of the waste until to its storage in repository ensuring the protection of man and of environment of the possible negative impacts. The radioactive waste are segregated, treated, conditioned in suitable packaging and posteriorly are stored or disposal in repository. The “RBMN Project” is a priority project of CNEN to implementation the repository for the deposition of low and intermediate level radioactive waste generated by nuclear energy activities in Brazil, proposing a definitive solution for its storage. Engineered and natural barriers as the filling layer and coverage layer will compose the disposal system of a near surface repository, concept proposed by the “RBMN Project”. The use these barriers views to avoid or restrict the release of radionuclides present in waste for the humans beings and environment. The waterproofing barriers are composed of clays. Certainly, for the national repository, will be used those clays existing in the place where it will be implanted it. So some fundamentals tests will have to be carried for to verify the suitability of these clays as barriers. These tests were determined and realized with a reference clay, a brazilian bentonite constituted of 67.2% montmorillonite. The results were compared with national and international literature of materials with similar mineralogical features. The values found with 95% reliance were 9.73±0,35 μm for granulometric size. For the moisture content were 13,3±0,6% and for capacity of cationic exchange were , 816±9 mmol.kg"-"1. For the hydraulic conductivity, without the use of internal pressure, it was obtained maximum value of 59.0% saturated. In addition, during the observed period, there was no percolation in the test specimen submitted to internal pressure of 200 kPa. This result leads to the conclusion that the

  14. The stability of candidate buffer materials for a low-level radioactive waste repository

    International Nuclear Information System (INIS)

    Torok, J.; Buckley, L.P.; Burton, G.R.; Tosello, N.B.; Maves, S.R.; Blimkie, M.E.; Donaldson, J.R.

    1989-11-01

    Inorganic ion-exchangers, clinoptilolite and clay, will be placed on the floor of a low-level radioactive waste repository to be built at Chalk River Nuclear Laboratories. The stability of these ion-exchange materials for a range of potential chemical environments in the repository was investigated. The leaching of waste forms and concrete and biological activity may create acidic or basic environment. The dissolution mechanisms of the ion exchangers for both acid and alkali conditions were established. Changes in distribution coefficients occurred shortly after the commencement of the treatment and were due to changes in the counter-ion content of the ion exchangers. No evidence was found to suggest gradual selective destruction of exchange sites responsible for the high distribution coefficients observed

  15. Geochemical performance of earthen and cementitious sealing materials for radioactive waste repositories

    International Nuclear Information System (INIS)

    Melchoir, D.; Glazier, R.; Marton, R.

    1988-01-01

    Earthen and cementitious materials are proposed as part of the sealing system for radioactive waste repositories. Compacted clay-bearing earthen materials could be used in sealing shafts and shaft entryways; and in the waste emplacement boundary areas in some repository designs. Earthen material mixtures are being considered because they can be engineered and emplaced to achieve low permeabilities, appropriate swelling characteristics, and adequate strength with little tendency to degrade during changing environmental conditions. The proposed earthen sealing materials include sodium and calcium mont-morillonites, illites, and mixtures with graded aggregates of sand. To assess the relative advantages and disadvantages of various pure and mixed materials, important geochemical processes (e.g., ion-exchange, phase transformation, dissolution, and precipitation of secondary minerals) need to be evaluated. These processes could impact seal integrity by changing permeability and/or mineral swell potential. Hydrous calcium-silicate-based cementitious materials such as grouts or concrete might also be used in some proposed sealing systems

  16. Research on swelling clays and bitumen as sealing materials for radioactive waste repositories

    International Nuclear Information System (INIS)

    Allison, J.A.; Wilson, J.; Mawditt, J.M.; Hurt, J.C.

    1990-10-01

    This report describes a programme of research to investigate the performance of composite seals comprising juxtaposed blocks of highly compacted bentonite clay and bitumen. It is shown that interaction of the materials can promote a self-sealing mechanism which prevents weather penetration, even when defects are present in the bitumen layer. Factors affecting seal performance are examined by means of laboratory experiments, and implications for the design of repository backfilling and sealing systems are discussed. It is concluded that design principles and material specifications should be further developed on the basis of large scale experiments. (author)

  17. New scenario for the accumulation and release of radiation damage in rock salt and related materials

    NARCIS (Netherlands)

    Hartog, H.W. den; Vainshtein, D.I.; Dubinko, V.I.; Turkin, A.A.

    2002-01-01

    Rock salt might be a promising geological medium for a radioactive waste repository. However, we have observed that even a basically stable compound such as NaCl may become unstable after heavy irradiation. As a result of the irradiation, dislocations, Na and Cl2 precipitates and large voids are

  18. 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.

  19. Site investigations, design, construction, operation, shutdown and surveillance of repositories for low- and intermediate-level radioactive wastes in rock cavities

    International Nuclear Information System (INIS)

    1984-01-01

    The report provides an overview and technical guidelines for considerations and for activities to be undertaken for safety assessment, site investigations, design, construction, operation, shutdown and surveillance of repositories for the disposal of low- and intermediate-level radioactive wastes in rock cavities. A generalized sequence of investigations is introduced which proceeds through region and site selection to the stage where the site is confirmed by detailed geoscientific investigations as being suitable for a waste repository. The different procedures and somewhat specific investigative needs with respect to existing mines are dealt with separately. General design, as well as specific requirements with respect to the different stages of design and construction, are dealt with. A review of activities related to the operational and post-operational stages of repositories in rock cavities is presented. The report describes in general terms the procedures related to different stages of disposal operation; also the conditions for shutdown together with essential shutdown and sealing activities and the related safety assessment requirements. Guidance is also given on the surveillance programme which will allow for inspection, testing, maintenance and security of a disposal facility during the operational phase, as well as for the post-operational stage for periods determined as necessary by the national authorities

  20. Sealing properties of cement-based grout materials used in the rock sealing project

    International Nuclear Information System (INIS)

    Onofrei, M.; Gray, M.N.; Pusch, R.; Boergesson, L.; Karnland, O.; Shenton, B.; Walker, B.

    1993-12-01

    The Task Force on Sealing Materials and Techniques of the Stripa Project recommended that work be undertaken to study the sealing properties of cement-based grout materials. A new class of cement-based grouts (high-performance grouts) with the ability to penetrate and seal fine fractures in granite was investigated. The materials were selected for their small mean particle size and the ability to be made fluid by a superplasticizer at low water/cementitious-materials ratios. The fundamental physical and chemical properties (such as the particle size and chemical composition) of the materials were evaluated. The rheological properties of freshly mixed grouts, which control the workability of the grouts, were determined together with the properties of hardened materials, which largely control the long-term performance (longevity) of the materials in repository settings. The materials selected were shown to remain gel-like during the setting period, and so the grouts may be expected to remain largely homogenous during and after injection into the rock without separating into solid and liquid phases. The hydraulic conductivity and strength of hardened grouts were determined. The microstructure of the bulk grouts was characterized by a high degree of homogeneity with extremely fine porosity. The low hydraulic conductivity and good mechanical properties are consistent with the extremely fine porosity. The ability of the fractured grouts to self-seal was also observed in tests in which the hydraulic conductivity of recompacted granulated grouts was determined. The laboratory tests were carried out in parallel with investigations of the in situ performance of the materials and with the development of geochemical and theoretical models for cement-based grout longevity. (author). 56 refs., 15 tabs., 98 figs

  1. Sealing properties of cement-based grout materials used in the rock sealing project

    Energy Technology Data Exchange (ETDEWEB)

    Onofrei, M; Gray, M N; Pusch, R; Boergesson, L; Karnland, O; Shenton, B; Walker, B

    1993-12-01

    The Task Force on Sealing Materials and Techniques of the Stripa Project recommended that work be undertaken to study the sealing properties of cement-based grout materials. A new class of cement-based grouts (high-performance grouts) with the ability to penetrate and seal fine fractures in granite was investigated. The materials were selected for their small mean particle size and the ability to be made fluid by a superplasticizer at low water/cementitious-materials ratios. The fundamental physical and chemical properties (such as the particle size and chemical composition) of the materials were evaluated. The rheological properties of freshly mixed grouts, which control the workability of the grouts, were determined together with the properties of hardened materials, which largely control the long-term performance (longevity) of the materials in repository settings. The materials selected were shown to remain gel-like during the setting period, and so the grouts may be expected to remain largely homogenous during and after injection into the rock without separating into solid and liquid phases. The hydraulic conductivity and strength of hardened grouts were determined. The microstructure of the bulk grouts was characterized by a high degree of homogeneity with extremely fine porosity. The low hydraulic conductivity and good mechanical properties are consistent with the extremely fine porosity. The ability of the fractured grouts to self-seal was also observed in tests in which the hydraulic conductivity of recompacted granulated grouts was determined. The laboratory tests were carried out in parallel with investigations of the in situ performance of the materials and with the development of geochemical and theoretical models for cement-based grout longevity. (author). 56 refs., 15 tabs., 98 figs.

  2. Thorium and Uranium in the Rock Raw Materials Used For the Production of Building Materials

    Science.gov (United States)

    Pękala, Agnieszka

    2017-10-01

    Thorium and uranium are constant components of all soils and most minerals thereby rock raw materials. They belong to the particularly dangerous elements because of their natural radioactivity. Evaluation of the content of the radioactive elements in the rock raw materials seems to be necessary in the early stage of the raw material evaluation. The rock formations operated from deposits often are accumulated in landfills and slag heaps where the concentration of the radioactive elements can be many times higher than under natural conditions. In addition, this phenomenon may refer to buildings where rock raw materials are often the main components of the construction materials. The global control system of construction products draws particular attention to the elimination of used construction products containing excessive quantities of the natural radioactive elements. In the presented study were determined the content of thorium and uranium in rock raw materials coming from the Bełachatów lignite deposit. The Bełchatów lignite deposit extracts mainly lignite and secondary numerous accompanying minerals with the raw material importance. In the course of the field works within the framework of the carried out work has been tested 92 samples of rocks of varied petrographic composition. There were carried out analyses of the content of the radioactive elements for 50 samples of limestone of the Jurassic age, 18 samples of kaolinite clays, and 24 samples of siliceous raw materials, represented by opoka-rocks, diatomites, gaizes and clastic rocks. The measurement of content of the natural radioactive elements thorium and uranium based on measuring the frequency counts of gamma quantum, recorded separately in measuring channels. At the same time performed measurements on volume patterns radioactive: thorium and uranium. The studies were carried out in Mazar spectrometer on the powdered material. Standardly performed ten measuring cycles, after which were calculated

  3. Selection criteria for container materials at the proposed Yucca Mountain high level nuclear waste repository

    International Nuclear Information System (INIS)

    Halsey, W.G.

    1989-11-01

    A geological repository has been proposed for the permanent disposal of the nation's high level nuclear waste at Yucca Mountain in the Nevada desert. The containers for this waste must remain intact for the unprecedented service lifetime of 1000 years. A combination of engineering, regulatory, and licensing requirements complicate the container material selection. In parallel to gathering information regarding the Yucca Mountain service environment and material performance data, a set of selection criteria have been established which compare candidate materials to the performance requirements, and allow a quantitative comparison of candidates. These criteria assign relative weighting to varied topic areas such as mechanical properties, corrosion resistance, fabricability, and cost. Considering the long service life of the waste containers, it is not surprising that the corrosion behavior of the material is a dominant factor. 7 refs

  4. Waste Package and Material Testing for the Proposed Yucca Mountain High Level Waste Repository

    International Nuclear Information System (INIS)

    Doering, Thomas; Pasupathi, V.

    2002-01-01

    Over the repository lifetime, the waste package containment barriers will perform various functions that will change with time. During the operational period, the barriers will function as vessels for handling, emplacement, and waste retrieval (if necessary). During the years following repository closure, the containment barriers will be relied upon to provide substantially complete containment, through 10,000 years and beyond. Following the substantially complete containment phase, the barriers and the waste package internal structures help minimize release of radionuclides by aqueous- and gaseous-phase transport. These requirements have lead to a defense-in-depth design philosophy. A multi-barrier design will result in a lower breach rate distributed over a longer period of time, thereby ensuring the regulatory requirements are met. The design of the Engineered Barrier System (EBS) has evolved. The initial waste package design was a thin walled package, 3/8 inch of stainless steel 304, that had very limited capacity, (3 PWR and 4 BWR assemblies) and performance characteristics, 300 to 1,000 years. This design required over 35,000 waste packages compared to today's design of just over 10,000 waste packages. The waste package designs are now based on a defense-in-depth/multi-barrier philosophy and have a capacity similar to the standard storage and rail transported spent nuclear fuel casks. Concurrent with the development of the design of the waste packages, a comprehensive waste package materials testing program has been undertaken to support the selection of containment barrier materials and to develop predictive models for the long-term behavior of these materials under expected repository conditions. The testing program includes both long-term and short-term tests and the results from these tests combination with the data published in the open literature are being used to develop models for predicting performance of the waste packages

  5. Investigations of possibilities to dispose of spent nuclear fuel in Lithuania: a model case. Volume 3, Generic Safety Assessment of Repository in Crystalline Rocks

    International Nuclear Information System (INIS)

    Motiejunas, S.; Poskas, P.

    2005-01-01

    In this Volume a generic safety assessment of the repository for spent nuclear fuel in crystalline rock in Lithuania is presented. Modeling of safety relevant radionuclide release from the defected canister and their transport through the near field and far field was performed. Doses to humans due to released radionuclides in the well water were calculated and compared with the dose restrictions existing in Lithuania. For this stage of generic safety assessment only two scenarios were chosen: base scenario and canister defect scenario. KBS-3 concept developed by SKB for disposal of spent nuclear fuel in Sweden was chosen as prototype for repository in crystalline basement in Lithuania. The KBS-3H design with horizontal canister emplacement is proposed as a reference design for Lithuania

  6. Analysis of Technical Status on the Application of Cementitious Materials for Radwaste Repository

    International Nuclear Information System (INIS)

    Kim, Jin Seop; Kwon, Sang Ki; Cho, Won Jin

    2008-12-01

    In this report, technical status on the application of cementitious materials and related research trends in Sweden, Switzerland and Japan etc. is listed based on the example of ONKALO in Finland. SKB and POSIVA have defined a pH limit ≤ 11 for cement grout leachates. To attain this pH, blending agents must comprise at least 50 wt % of dry materials. Because low pH cement has little, or no free portlandite, the cement consists predominantly of calcium silicate hydrate(CSH) gel with a Ca/Si ratio ≤ 0.8(Savage D. 2007). Silica fume as a blending agent is considered to be most promising for repository low-pH grouts. When adding silica fume to enhance cement quality, it demands high water content in cement paste. Then it is necessary to use additives such as superplasticiser to improve the workability of low-pH cement. Posiva, SKB and NUMO co-operated in developing low-pH grouts for deep repositories 2002-2005. Additionally, it is needed to study more about long-term performance characteristics, interaction of bentonite buffer material with high pH plume, influence on the migration/sorption of radionuclides and their performance numerical modeling. In this regards, international co-research projects such as ESDRED and IAEA CRP are being actively performed

  7. Some observations on the mechanism of corrosion to be encountered in nuclear waste repositories located in tuffaceous rock

    International Nuclear Information System (INIS)

    Wilde, M.H.; Wilde, B.E.

    1993-01-01

    Potentiostatic anodic polarization studies have been conducted in a J-13 simulated nuclear waste repository environment, which was allowed to evaporate to dryness followed by rehydration prior to polarization. The behavior of Type 316L stainless steel, AISI 1020 carbon steel, Hastelloy C22 and platinum was compared with that noted previously for a non-baked simulate. The anodic dissolution characteristics of Type 316L stainless steel in environments containing 1000X Cl - J-13 depend markedly on whether the solution is merely a mixture of virgin chemicals or a mixture that has been evaporated to dryness, baked and rehydrated to the same volume. In the non-evaporated environment Type 316L stainless steel pitted severely, and in the evaporated/rehydrated environment a non-corroding type of behavior was observed along with the precipitation of a dense scale. Similar behavior was observed for Hastelloy C22. The polarization curves for carbon steel and platinum were the same as those noted for 316L and Hastelloy C22, when conducted in the evaporated/rehydrated environment. X-ray diffraction studies indicated that the scale produced in all tests conducted on evaporated/rehydrated solutions was calcium carbonate. Based on the qualitatively similar polarization characteristics of materials having such widely differing corrosion properties, it is concluded that the major factor controlling the anodic charge transfer reaction under these conditions is the formation of a calcium carbonate scale. (Author)

  8. Planned investigations for packing materials for a waste package in a salt repository: [Final report

    International Nuclear Information System (INIS)

    Shade, J.W.; Bunnell, L.R.; Thornton, T.A.

    1987-10-01

    A considerable number of materials have been either proposed or investigated as packing materials for nuclear waste package systems. Almost always the expandable clays, such as the smectites contained in commercial bentonites, have received the most attention when their primary function is to retard groundwater flow. Other materials including zeolites, metals, and dessicants are considered as special-purpose additives. Materials that tend to hydrolyze and lead to porosity reduction, such as silicates, oxides, and sulfates, have also been suggested as packing materials. All these types of materials are also considered as components of tailored mixtures to achieve a broad range of packing material performance. Some of these materials are reviewed, along with proposed candidate materials, with respect to the properties required to function in a salt repository. The investigation of packing materials is composed of five studies which are discussed below. Initial candidates will consist of calcium hydroxide, a sodium silicate, and a cement-gypsum mixture in addition to the reference crushed salt. Consequently these tests will be necessary to determine properties of individual components and to optimize properties of mixtures. 13 refs., 7 figs., 1 tab

  9. DECOVALEX III/BENCHPAR PROJECTS. Implications of Thermal-Hydro-Mechanical Coupling on the Near-Field Safety of a Nuclear Waste Repository in a Homogeneous Rock Mass. Report of BMT1B/WP2

    International Nuclear Information System (INIS)

    Jing, L.

    2005-02-01

    This report presents the works performed for the second phase (BMT1B) of BMT1 of the DECOVALEX III project for the period of 1999-2002. The works of BMT1 is divided into three phases: BMT1A, BMT1B and BMT1C. The BMT1A concerns with calibration of the computer codes with a reference T-H-M experiment at Kamaishi Mine, Japan. The objective is to validate the numerical approaches, computer codes and material models, so that the teams simulating tools are at a comparable level of maturity and sophistication. The BMT1B uses the calibrated codes to perform scoping calculations, considering varying degrees of THM coupling and varying permeability values of the surrounding rock for a reference generic repository design without fractures. The aim is to identify the coupling mechanisms of importance for construction, performance and safety of the repository. The chosen measures for evaluating the long term safety and performance of the repository are the maximal temperature created by the thermal loading from the emplaced wastes, the time for re-saturation of the buffer, the maximal swelling stress developed in the buffer, the structural integrity of the rock mass and the permeability evolution in the rock mass. Six teams participated in BMT1B: IRSN/CEA (France), CNSC (Canada), ANDRA/INERIS (France), JNC (Japan), BGR/ISEB-ZAG (Germany) and SKI/KTH (Sweden). All teams used FEM approach except the ANDRA/INERIS team who used the FDM approach, with different codes. All research teams except ISEB/ZAG used models with full THM coupling capabilities. The governing equations in these models were derived within the framework of Biot's theory of consolidation and have for primary unknown variables: temperature, pore fluid pressure and displacements of the solid skeleton. Since the original Biot's theory of consolidation is applicable to saturated materials and isothermal conditions, the research teams have to extend Biot's theory in order to deal with thermal effects and the variably

  10. DECOVALEX III/BENCHPAR PROJECTS. Implications of Thermal-Hydro-Mechanical Coupling on the Near-Field Safety of a Nuclear Waste Repository in a Homogeneous Rock Mass. Report of BMT1B/WP2

    Energy Technology Data Exchange (ETDEWEB)

    Jing, L. [Royal Inst. of Technology, Stockholm (Sweden). Engineering Geology; Nguyen, T.S. [Canadian Nuclear Safety Commission, Ottawa, ON (Canada)] (eds.)

    2005-02-15

    This report presents the works performed for the second phase (BMT1B) of BMT1 of the DECOVALEX III project for the period of 1999-2002. The works of BMT1 is divided into three phases: BMT1A, BMT1B and BMT1C. The BMT1A concerns with calibration of the computer codes with a reference T-H-M experiment at Kamaishi Mine, Japan. The objective is to validate the numerical approaches, computer codes and material models, so that the teams simulating tools are at a comparable level of maturity and sophistication. The BMT1B uses the calibrated codes to perform scoping calculations, considering varying degrees of THM coupling and varying permeability values of the surrounding rock for a reference generic repository design without fractures. The aim is to identify the coupling mechanisms of importance for construction, performance and safety of the repository. The chosen measures for evaluating the long term safety and performance of the repository are the maximal temperature created by the thermal loading from the emplaced wastes, the time for re-saturation of the buffer, the maximal swelling stress developed in the buffer, the structural integrity of the rock mass and the permeability evolution in the rock mass. Six teams participated in BMT1B: IRSN/CEA (France), CNSC (Canada), ANDRA/INERIS (France), JNC (Japan), BGR/ISEB-ZAG (Germany) and SKI/KTH (Sweden). All teams used FEM approach except the ANDRA/INERIS team who used the FDM approach, with different codes. All research teams except ISEB/ZAG used models with full THM coupling capabilities. The governing equations in these models were derived within the framework of Biot's theory of consolidation and have for primary unknown variables: temperature, pore fluid pressure and displacements of the solid skeleton. Since the original Biot's theory of consolidation is applicable to saturated materials and isothermal conditions, the research teams have to extend Biot's theory in order to deal with thermal effects and

  11. 2012 best practices for repositories collection, storage, retrieval, and distribution of biological materials for research international society for biological and environmental repositories.

    Science.gov (United States)

    2012-04-01

    Third Edition [Formula: see text] [Box: see text] Printed with permission from the International Society for Biological and Environmental Repositories (ISBER) © 2011 ISBER All Rights Reserved Editor-in-Chief Lori D. Campbell, PhD Associate Editors Fay Betsou, PhD Debra Leiolani Garcia, MPA Judith G. Giri, PhD Karen E. Pitt, PhD Rebecca S. Pugh, MS Katherine C. Sexton, MBA Amy P.N. Skubitz, PhD Stella B. Somiari, PhD Individual Contributors to the Third Edition Jonas Astrin, Susan Baker, Thomas J. Barr, Erica Benson, Mark Cada, Lori Campbell, Antonio Hugo Jose Froes Marques Campos, David Carpentieri, Omoshile Clement, Domenico Coppola, Yvonne De Souza, Paul Fearn, Kelly Feil, Debra Garcia, Judith Giri, William E. Grizzle, Kathleen Groover, Keith Harding, Edward Kaercher, Joseph Kessler, Sarah Loud, Hannah Maynor, Kevin McCluskey, Kevin Meagher, Cheryl Michels, Lisa Miranda, Judy Muller-Cohn, Rolf Muller, James O'Sullivan, Karen Pitt, Rebecca Pugh, Rivka Ravid, Katherine Sexton, Ricardo Luis A. Silva, Frank Simione, Amy Skubitz, Stella Somiari, Frans van der Horst, Gavin Welch, Andy Zaayenga 2012 Best Practices for Repositories: Collection, Storage, Retrieval and Distribution of Biological Materials for Research INTERNATIONAL SOCIETY FOR BIOLOGICAL AND ENVIRONMENTAL REPOSITORIES (ISBER) INTRODUCTION T he availability of high quality biological and environmental specimens for research purposes requires the development of standardized methods for collection, long-term storage, retrieval and distribution of specimens that will enable their future use. Sharing successful strategies for accomplishing this goal is one of the driving forces for the International Society for Biological and Environmental Repositories (ISBER). For more information about ISBER see www.isber.org . ISBER's Best Practices for Repositories (Best Practices) reflect the collective experience of its members and has received broad input from other repository professionals. Throughout this document

  12. Aespoe Hard Rock Laboratory. Prototype repository. Analyses of microorganisms, gases, and water chemistry in buffer and backfill, 2010

    International Nuclear Information System (INIS)

    Lydmark, Sara

    2011-06-01

    The prototype repository (hereafter, 'Prototype') is an international project to build and study a fullscale model of the planned Swedish final repository for spent nuclear fuel. However, the Prototype differs from a real storage in that it is drained, which makes the swelling pressure lower in the Prototype than in a real storage facility. The heat from the radioactive decay is simulated by electrical heaters. The project is being conducted at the Aespoe Hard Rock Laboratory (HRL) in crystalline rock at a depth of approximately 450 m. A monitoring programme is investigating the evolution of the water chemistry, gas, and microbial activity at the site, and a specific aim is to monitor the microbial consumption of oxygen in situ in the Prototype. This document describes the results of the analyses of microbes, gases, and chemistry inside the Prototype in 2010. Hydrogen, helium, nitrogen, oxygen, carbon monoxide, carbon dioxide, methane, ethane, and ethene were analysed at the following sampling points in the Prototype: KBU10001, KBU10002, KBU10004, KBU10008, and KFA04. Where the sampling points in the Prototype delivered pore water, the water was analysed for amount of ATP (i.e. the biovolume), culturable heterotrophic aerobic bacteria (CHAB), sulphate-reducing bacteria (SRB), methane-oxidizing bacteria (MOB), and iron-reducing bacteria (IRB). The pore water collected from the Prototype was subject to as many chemical analyses as the amount of water allowed. Chemical analyses were also performed on pore water from two additional sampling points, KBU10005 and KBU10006. Chemical data from a previous investigation of the groundwater outside the Prototype were compared with the pore water chemistry. The improved sampling and analysis protocols introduced in 2007 worked very well. The International Progress Report (IPR) 08-01 (Eriksson 2008) revealed that many of the hydrochemical sampling points differ greatly from each other. The 16 sampling points were therefore

  13. Aespoe Hard Rock Laboratory. Prototype repository. Analyses of microorganisms, gases, and water chemistry in buffer and backfill, 2010

    Energy Technology Data Exchange (ETDEWEB)

    Lydmark, Sara [Microbial Analytics Sweden AB, Moelnlycke (Sweden)

    2011-06-15

    The prototype repository (hereafter, 'Prototype') is an international project to build and study a fullscale model of the planned Swedish final repository for spent nuclear fuel. However, the Prototype differs from a real storage in that it is drained, which makes the swelling pressure lower in the Prototype than in a real storage facility. The heat from the radioactive decay is simulated by electrical heaters. The project is being conducted at the Aespoe Hard Rock Laboratory (HRL) in crystalline rock at a depth of approximately 450 m. A monitoring programme is investigating the evolution of the water chemistry, gas, and microbial activity at the site, and a specific aim is to monitor the microbial consumption of oxygen in situ in the Prototype. This document describes the results of the analyses of microbes, gases, and chemistry inside the Prototype in 2010. Hydrogen, helium, nitrogen, oxygen, carbon monoxide, carbon dioxide, methane, ethane, and ethene were analysed at the following sampling points in the Prototype: KBU10001, KBU10002, KBU10004, KBU10008, and KFA04. Where the sampling points in the Prototype delivered pore water, the water was analysed for amount of ATP (i.e. the biovolume), culturable heterotrophic aerobic bacteria (CHAB), sulphate-reducing bacteria (SRB), methane-oxidizing bacteria (MOB), and iron-reducing bacteria (IRB). The pore water collected from the Prototype was subject to as many chemical analyses as the amount of water allowed. Chemical analyses were also performed on pore water from two additional sampling points, KBU10005 and KBU10006. Chemical data from a previous investigation of the groundwater outside the Prototype were compared with the pore water chemistry. The improved sampling and analysis protocols introduced in 2007 worked very well. The International Progress Report (IPR) 08-01 (Eriksson 2008) revealed that many of the hydrochemical sampling points differ greatly from each other. The 16 sampling points were

  14. Analysis of Academic Attitudes and Existing Processes to Inform the Design of Teaching and Learning Material Repositories: A User-Centred Approach

    Science.gov (United States)

    King, Melanie; Loddington, Steve; Manuel, Sue; Oppenheim, Charles

    2008-01-01

    The last couple of years have brought a rise in the number of institutional repositories throughout the world and within UK Higher Education institutions, with the majority of these repositories being devoted to research output. Repositories containing teaching and learning material are less common and the workflows and business processes…

  15. Sorption of plutonium and americium on repository, backfill and geological materials relevant to the JNFL low-level radioactive waste repository at Rokkasho-Mura

    International Nuclear Information System (INIS)

    Baston, G.M.N.; Berry, J.A.; Brownsword, M.; Heath, T.G.; Tweed, C.J.; Williams, S.J.

    1995-01-01

    An integrated program of batch sorption experiments and mathematical modeling has been carried out to study the sorption of plutonium and americium on a series of repository, backfill and geological materials relevant to the JNFL low-level radioactive waste repository at Rokkasho-Mura. The sorption of plutonium and americium on samples of concrete, mortar, sand/bentonite, tuff, sandstone and cover soil has been investigated. In addition, specimens of bitumen, cation and anion exchange resins, and polyester were chemically degraded. The resulting degradation product solutions, alongside solutions of humic and isosaccharinic acids were used to study the effects on plutonium sorption onto concrete, sand/bentonite and sandstone. The sorption behavior of plutonium and americium has been modeled using the geochemical speciation program HARPHRQ in conjunction with the HATCHES database

  16. Interaction of Water with Cement Based Repository Materials - Application of Neutron Imaging

    International Nuclear Information System (INIS)

    Mcglinn, P.J.; Brew, D.R.M.; Beer, F.C. De; Radebe, M.J.; Nshimirimana, R.

    2013-01-01

    Cementitious materials are conventionally used in conditioning intermediate and low level radioactive waste. In this study, a candidate cement-based wasteform and a series of barrier materials have been investigated using neutron imaging to: 1) characterise the wasteform for disposal in a repository for radioactive materials, and 2) characterise the compositon of the barrier materials in assessing their potential to transmit water. Imaging showed both the pore size distribution and the extent of the cracking that had occurred in the wasteform samples. The rate of the water penetration measured both by conventional sorptivity measurements and neutron imaging was greater than in pastes made from Ordinary Portland Cement. The ability of the cracks to distribute the water through the sample in a very short time was also evident. Macro-pore volume distributions of barrier samples, also acquired using neutron tomography, are shown to relate to water/cement ratio, composition and sorptivity data. The study highlights the significant potential of neutron imaging in the investigation of cementitious materials. The technique has the advantage of visualising and measuring, non-destructively, material distribution within macroscopic samples and is particularly useful in defining movement of water through the cementitious materials. (author)

  17. PSU/WES interlaboratory comparative methodology study of an experimental cementitious repository seal material

    International Nuclear Information System (INIS)

    Roy, D.M.; Grutzeck, M.W.; Mather, K.

    1980-09-01

    A study is underway in two separate laboratories to investigate possible use of portland cement grout as repository sealing material for underground isolation of nuclear waste. The labs involved are the Materials Research Laboratory of the Pennsylvania State University (PSU) and the Structures Laboratory (SL) of the US Army Engineer Waterways Experiment Station. The same cementitious (grout) mixture was prepared in each laboratory in September 1980, and tests were started. Testing included characterization of cement and fly ash by chemical, physical, and petrographic procedures. Tests of hardened specimens included restrained expansion, compressive strength, modulus of elasticity, density, permeability, x-ray diffraction, and scanning electron microscopy. Each laboratory made many of the same tests and some that were not directly comparable. This document (Report 1) contains largely 3- and 7-day results and none beyond 28-day ages

  18. Swelling pressures of a potential buffer material for high-level waste repository

    International Nuclear Information System (INIS)

    Lee, Jae Owan; Cho, Won Jin; Chun, Kwan Sik

    1999-01-01

    The swelling pressure of a potential buffer material was measured and the effect of dry density, bentonite content and initial water content on the swelling pressure was investigated to provide the information for the selection of buffer material in a high-level waste repository. Swelling tests were carried out according to Box-Behnken's experimental design. Measured swelling pressures were in the wide range of 0.7 Kg/cm 2 to 190.2 Kg/cm 2 under given experimental conditions. Based upon the experimental data, a 3-factor polynomial swelling model was suggested to analyze the effect of dry density, bentonite content and initial water content on the swelling pressure. The swelling pressure increased with an increase in the dry density and bentonite content, while it decreased with increasing the initial water content and, beyond about 12 wt.% of the initial water content, levelled to nearly constant value. (author). 21 refs., 10 figs., 4 tabs

  19. Long Term Behaviour of Cementitious Materials in the Korean Repository Environment

    International Nuclear Information System (INIS)

    Park, J.-W.; Kim, C.-L.

    2013-01-01

    The safe management of radioactive waste is a national task required for sustainable generation of nuclear power and for energy self-reliance in Korea. After the selection of the final candidate site for low- and intermediate-level waste (LILW) disposal in Korea, a construction and operation license was issued for the Wolsong LILW Disposal Center (WLDC) for the first stage of disposal. Underground silo type disposal has been determined for the initial phase. The engineered barrier system of the disposal silo consists of waste packages, disposal containers, backfills, and a concrete lining. Main objective of our study in this IAEA-CRP is to investigate closure concepts and cementitious backfill materials for the closure of silos. For this purpose, characterisation of cementitious materials, development of silo closure concept, and evaluation of long-term behaviour of cementitious materials, including concrete degradation in repository environment, have been carried out. The overall implementation plan for the CRP comprises performance testing for the physic-chemical properties of cementitious materials, degradation modelling of concrete structures, comparisons of performance for silo closure options, radionuclide transport modelling (considering concrete degradation in repository conditions), and the implementation of an input parameter database and quality assurance for safety/performance assessment. In particular, the concrete degradation modelling study has been focused on the corrosion of reinforcement steel induced by chloride attack, which was of primary concern in the safety assessment of the WLDC. A series of electrochemical experiments were conducted to investigate the effect of dissolved oxygen, pH, and Cl on the corrosion rate of reinforcing steel in a concrete structure saturated with groundwater. Laboratory-scale experiments and a thermodynamic modelling were performed to understand the porosity change of cement pastes, which were prepared using

  20. Crevice Corrosion Behavior of Candidate Nuclear Waste Container Materials in Repository Environment Paper Number 02529

    International Nuclear Information System (INIS)

    Hua, F.; Sarver, J.; Mohn, W.

    2001-01-01

    Alloy 22 (UNS N06022) and Ti Grade 7 (UNS R52400) have been proposed as the corrosion resistant materials for fabricating the waste package outer barrier and the drip shield, respectively for the proposed nuclear waste repository Yucca Mountain Project. In this work, the susceptibility of welded and annealed Alloy 22 (N06022) and Ti Grade 7 (UNS R52400) to crevice corrosion was studied by the Multiple Crevice Assembly (ASTM G78) method combined with surface morphological observation after four and eight weeks of exposure to the Basic Saturated Water (BSW-12) in a temperature range from 60 to 105 C. The susceptibility of the materials to crevice corrosion was evaluated based on the appearance of crevice attack underneath the crevice formers and the weight loss data. The results showed that, after exposed to BSW-12 for four and eight weeks, no obvious crevice attack was observed on these materials. The descaled weight loss increased with the increase in temperature for all materials. The weight loss, however, is believed to be caused by general corrosion, rather than crevice corrosion. There was no significant difference between the annealed and welded materials either. On the other hand, to conclude that these materials are immune to crevice corrosion in BSW-12 will require longer term testing

  1. Far-field sorption data bases for performance assessment of a L/ILW repository in an undisturbed Palfris marl host rock

    International Nuclear Information System (INIS)

    Bradbury, M.H.; Baeyens, B.

    1997-12-01

    A Palfris marl formation at Wellenberg (Gemeinde Wolfenschiessen, NW) has been chosen by NAGRA as a potential repository site for low- and intermediate-level radioactive waste, L/ILW. In the coming years a series of performance assessment studies will be performed for this site. One set of key data required for such safety analysis calculations is sorption data bases (SDB) for safety relevant radionuclides in the far-field. The purpose of this report is to describe the procedures used to generate sorption data bases appropriate for the in situ conditions existing along the different potential flow paths in an undisturbed marl host rock formation. An important aim was to document the sources of sorption data used and, in particular, the processes by which data selections were mad.e. The main guiding principles here were 'transparency' and 'traceability'. Inherent within this whole process is also the justification for, and defensibility of, the selected values. Much of the sorption data used to generate the SDB for marl came from the open literature. A major part of this report is concerned with describing the procedures whereby these initial literature values are modified so that they apply to the actual marl mineralogies and groundwater chemistries. The resulting 'reference R d values' are then further modified using so called Lab -> Field transfer factors to produce sorption values which are appropriate to the in situ bulk rock conditions. The Lab -> Field transfer factors attempt to correct for the differences in sorption site availability between the crushed rock state used in batch tests and the intact rock state existing in reality in the host rock. There are two main groundwater chemistries and five characteristic mineralogical compositions which cover the three broad types of flow paths which have been identified in the Palfris marl formation. In principle the methodology described here to construct sorption data bases for marl is applicable to any type of

  2. Design aspects of the Alpha Repository: III. Uniaxial quasi-static and creep properties of the site rock. Technical memorandum report RSI-0029

    International Nuclear Information System (INIS)

    Hansen, F.D.; Gnirk, P.F.

    1975-01-01

    Candidate mining horizons for the Alpha Repository have been tentatively selected at depths of 1,900, 2,100, and 2,700 ft in the massive salt formations underlying Eddy and Lea counties in New Mexico. The rock salt in the mining horizon at 1,900 ft exhibits average tensile and uniaxial compressive strengths of 200 and 2,445 psi, while the rock salt in the 2,700 ft horizon is 20 to 35 percent stronger. The elastic constants were essentially identical for the two horizons, with an average Young's modulus of 1.94 x 10 6 psi and a Poisson's ratio of 0.33 to 0.34. The anhydrite exhibits tensile and uniaxial compressive strengths of 830 and 13,085 psi, and its Poisson's ratio is 0.35, essentially the same as for rock salt, but its Young's modulus is 10.2 x 10 6 psi, five times greater than that of rock salt. In general, rock salt exhibits a type of bilinear stress-strain curve, with a discontinuity in slope occurring at about 750 psi. Rock salt appears to fail by crushing, rather than in an abrupt ''brittle fracture'' fashion. Anhydrite exhibits a linear stress-strain relationship, with abrupt and distinct failure at the level required for rupture. Uniaxial creep tests were performed on specimens from the 1,900 ft and 2,700 ft horizons using stress levels of 750 and 1,500 psi from 30 to over 200 hours. Results indicate that, for a constant stress level, strain is a function of time to the power of 0.20 to 0.24 and strain appears to be a nonlinear function of the deviatoric stress. Neither steady-state nor tertiary creep was observed

  3. Far-field sorption data bases for performance assessment of a L/ILW repository in an undisturbed Palfris marl host rock

    Energy Technology Data Exchange (ETDEWEB)

    Bradbury, M.H.; Baeyens, B. [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

    1997-12-01

    A Palfris marl formation at Wellenberg (Gemeinde Wolfenschiessen, NW) has been chosen by NAGRA as a potential repository site for low- and intermediate-level radioactive waste, L/ILW. In the coming years a series of performance assessment studies will be performed for this site. One set of key data required for such safety analysis calculations is sorption data bases (SDB) for safety relevant radionuclides in the far-field. The purpose of this report is to describe the procedures used to generate sorption data bases appropriate for the in situ conditions existing along the different potential flow paths in an undisturbed marl host rock formation. An important aim was to document the sources of sorption data used and, in particular, the processes by which data selections were mad.e. The main guiding principles here were `transparency` and `traceability`. Inherent within this whole process is also the justification for, and defensibility of, the selected values. Much of the sorption data used to generate the SDB for marl came from the open literature. A major part of this report is concerned with describing the procedures whereby these initial literature values are modified so that they apply to the actual marl mineralogies and groundwater chemistries. The resulting `reference R{sub d} values` are then further modified using so called Lab -> Field transfer factors to produce sorption values which are appropriate to the in situ bulk rock conditions. The Lab -> Field transfer factors attempt to correct for the differences in sorption site availability between the crushed rock state used in batch tests and the intact rock state existing in reality in the host rock. There are two main groundwater chemistries and five characteristic mineralogical compositions which cover the three broad types of flow paths which have been identified in the Palfris marl formation. In principle the methodology described here to construct sorption data bases for marl is applicable to any

  4. Crushed aggregate-betonite mixtures as backfill material for the Finnish repositories of low- and intermediate-level radioactive wastes

    International Nuclear Information System (INIS)

    Holopainen, P.; Pirhonen, V.; Snellman, M.

    1984-03-01

    Backfill materials consisting of three components: crushed rock aggregate, finely ground rock aggregate and bentonite (3 to 2 per cent of weight) were studied. The production and installation procedures of the material were evaluated. Laboratory tests were made to determine the hydraulic conductivity and swelling potential of the materials. Chemical tests were made on the different materials and groundwaters. Mineralogical changes of the clay fraction were estimated. (author)

  5. Development of low-pH cementitious materials for HLRW repositories

    International Nuclear Information System (INIS)

    Garcia Calvo, J.L.; Hidalgo, A.; Alonso, C.; Fernandez Luco, L.

    2010-01-01

    One of the most accepted engineering construction concepts of underground repositories for high radioactive waste considers the use of low-pH cementitious materials. This paper deals with the design of those based on Ordinary Portland Cements with high contents of silica fume and/or fly ashes that modify most of the concrete 'standard' properties, the pore fluid composition and the microstructure of the hydrated products. Their resistance to long-term groundwater aggression is also evaluated. The results show that the use of OPC cement binders with high silica content produces low-pH pore waters and the microstructure of these cement pastes is different from the conventional OPC ones, generating C-S-H gels with lower CaO/SiO 2 ratios that possibly bind alkali ions. Leaching tests show a good resistance of low-pH concretes against groundwater aggression although an altered front can be observed.

  6. The influence of introduced micro-organisms on corrosion of repository construction materials

    International Nuclear Information System (INIS)

    West, J.M.

    1985-01-01

    The work described in this report forms part of a wider project on the role of geomicrobiology in radioactive waste containment. This has established the presence of microbes in relevant geological formations including several groups of significance to waste containment. Microbial groups demonstrated have included those which could influence deterioration of repository structural materials, eg. sulphate reducing bacteria (SRB). This report describes work carried out to assess this role. More specifically the objectives of this phase of the project are: identification of suitable microbial isolates; to ascertain the growth characteristics of the isolates; to develop and construct experimental cells for use in corrosion rate tests; and to conduct preliminary short term experiments in static conditions designed to assess corrosion rates of mild steel in an ideal growth environment for SRB. Using information gained from these experiments to initiate long term corrosion experiments of steel in an SRB inoculated bentonite simulating near-field conditions in a backfill/canister system. (author)

  7. 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

  8. Radioactive waste repository study

    International Nuclear Information System (INIS)

    1978-11-01

    This is the second part of a report of a preliminary study for AECL. It considers the requirements for an underground waste repository for the disposal of wastes produced by the Canadian Nuclear Fuel Program. The following topics are discussed with reference to the repository: 1) geotechnical assessment, 2) hydrogeology and waste containment, 3) thermal loading and 4) rock mechanics. (author)

  9. Study of inorganic sorbents as materials for underground repositories in China

    International Nuclear Information System (INIS)

    Zhixiong, W.

    1989-01-01

    Since 1983, the construction of nuclear power plants has been taking place in Zhejiang and Guangdong provinces of China. The project is a part of the radwaste disposal plan of China. The project under the contract with IAEA studies absorption kinetics and mechanism of backfill material and selection of proper backfill material for the radwaste disposal plan. There are varieties of clay minerals as inorganic sorbents in China, such as zeolite, illite montomorillonite, kolinite, and so on. Bentonite is the first selected material for the research project. Bentonite is a common montonorillonite clay with good mechanical properties and chemical stability under certain conditions in a repository capacity. There are many huge bentonite deposits in China. China's LILW disposal will be possibly selected in the bentonite district. The investigation of China's bentonite will include the properties of China's sites, the study of migration of radionuclides and the geochemistry of actinides elements. Various bentonites of China have been studied to select one of good quality. The project is significant to assess the barrier ability of bentonite. The project also made the primary work for zeolite as a sorbent which has been used for the disposal of LILW liquid in China. Clinopliotite has been used in China's hydraulic fracture test of the radwaste liquid

  10. Research and development activities at INE concerning corrosion of final repository container materials

    International Nuclear Information System (INIS)

    Kienzler, Bernhard

    2017-01-01

    The present work provides a historical overview of the research and development activities carried out at the (Nuclear) Research Center Karlsruhe (today KIT) since the beginning of the 1980s on the corrosion of materials which might be suitable for construction of containers for highly radioactive wastes. The report relates almost exclusively to the work performed by Dr. Emmanuel Smailos, who elaborated the corrosion of various materials at the Institute for Nuclear Waste Disposal (INE). The requirements for the containers and materials, which were subject to changes in time, are presented. The changes were strongly influenced by the changed perception of the use of nuclear energy. The selection of the materials under investigations, the boundary conditions for the corrosion experiments and the analytical methods are described. Results of the corrosion of the materials such as finegrained steel, Hastelloy C4, nodular cast iron, titanium-palladium and copper or copper-nickel alloys in typical salt solutions are summarized. The findings of special investigations, e.g. corrosion under irradiation or the influence of sulfide on the corrosion rates are shown. For construction of disposal canisters, experiments were conducted to determine the contact corrosion, the influence of the hydrogen embrittlement of Ti-Pd and fine-grained steels on the corrosion behavior as well as the corrosion behavior of welding and the influence of different welding processes with the resulting heat-affected zones on the corrosion behavior. The work was contributed to several European research programs and was well recognized in the USA. Investigations on the corrosion of steels in non-saline solutions and corrosion under interim storage conditions as well as under the expected conditions of the Konrad repository for low-level radioactive wastes are also described. In addition, the experiments on ceramic materials are presented and the results of the corrosion of Al 2 O 3 and ZrO 2 ceramics

  11. Evolution of cement based materials in a repository for radioactive waste and their chemical barrier function

    International Nuclear Information System (INIS)

    Kienzler, Bernhard; Metz, Volker; Schlieker, Martina; Bohnert, Elke

    2015-01-01

    The use of cementitious materials in nuclear waste management is quite widespread. It covers the solidification of low/intermediate-level liquid as well as solid wastes (e.g. laboratory wastes) and serves as shielding. For both high-level and intermediate-low level activity repositories, cement/concrete likewise plays an important role. It is used as construction material for underground and surface disposals, but more importantly it serves as barrier or sealing material. For the requirements of waste conditioning, special cement mixtures have been developed. These include special mixtures for the solidification of evaporator concentrates, borate binding additives and for spilling solid wastes. In recent years, low-pH cements were strongly discussed especially for repository applications, e.g. (Celine CAU DIT COUMES 2008; Garcia-Sineriz, et al. 2008). Examples for relevant systems are Calcium Silicate Cements (ordinary Portland cement (OPC) based) or Calcium Aluminates Cements (CAC). Low-pH pore solutions are achieved by reduction of the portlandite content by partial substitution of OPC by mineral admixtures with high silica content. The blends follow the pozzolanic reaction consuming Ca(OH) 2 . Potential admixtures are silica fume (SF) and fly ashes (FA). In these mixtures, super plasticizers are required, consisting of polycarboxilate or naphthalene formaldehyde as well as various accelerating admixtures (Garcia-Sineriz, et al. 2008). The pH regime of concrete/cement materials may stabilize radionuclides in solution. Newly formed alteration products retain or release radionuclides. An important degradation product of celluloses in cement is iso-saccharin acid. According to Glaus 2004 (Glaus and van Loon 2004), it reacts with radionuclides forming dissolved complexes. Apart from potentially impacting radionuclide solubility limitations, concrete additives, radionuclides or other strong complexants compete for surface sites for sorbing onto cement phases. In

  12. The influence of organic materials on the near field of an intermediate level waste radioactive waste repository

    International Nuclear Information System (INIS)

    Wilkins, J.D.

    1988-02-01

    The influence of organic materials, which are present in some intermediate level wastes, on the chemistry of the near field of a radioactive waste repository is discussed. Particular attention is given to the possible formation of water soluble complexing agents formed as a result of the radiation field and chemical conditions. The present state of the research is reviewed. (author)

  13. Developing institutional repository at National Institute for Materials Science : Researchers directory service “SAMURAI” and Research Collection Library

    Science.gov (United States)

    Takaku, Masao; Tanifuji, Mikiko

    National Institute for Materials Science (NIMS) has developed an institutional repository “NIMS eSciDoc” since 2008. eSciDoc is an open source repository software made in Germany, and provides E-Science infrastructures through its flexible data model and rich Web APIs. NIMS eScidoc makes use of eSciDoc functions to benefit for NIMS situations. This article also focuses on researchers directory service “SAMURAI” in addition to NIMS eSciDoc. Successfully launched in October 2010, SAMURAI provides approximately 500 researchers' profile and publication information.

  14. The separation of radionuclide migration by solution and particle transport in LLRW repository buffer material

    International Nuclear Information System (INIS)

    Torok, J.; Buckley, L.P.; Woods, B.L.

    1989-01-01

    Laboratory-scale lysimeter experiments were performed with simulated waste forms placed in candidate buffer materials which have been chosen for a low-level radioactive waste repository. Radionuclide releases into the effluent water and radionuclide capture by the buffer material were determined. The results could not be explained by traditional solution transport mechanisms, and transport by particles released from the waste form and/or transport by buffer particles were suspected as the dominant mechanism for radionuclide release from the lysimeters. To elucidate the relative contribution of particle and solution transport, the waste forms were replaced by a wafer of neutron-activated buffer soaked with selected soluble isotopes. Particle transport was determined by the movement of gamma-emitting neutron-activation products through the lysimeter. Solution transport was quantified by comparing the migration of soluble radionuclides relative to the transport of neutron activation products. The new approach for monitoring radionuclide migration in soil is presented. It facilitates the determination of most of the fundamental coefficients required to model the transport process

  15. Experimental studies on the migration of radionuclides of the elements I, Sr, Cs, Co and Pd in the roof rock of the projected waste repository at Gorleben

    International Nuclear Information System (INIS)

    Klotz, D.; Lang, H.; Moser, H.

    1985-07-01

    The studies were intended to provide information on the sorptive properties of 15 samples of fine-grain and medium-grain sands with regard to the radionuclides of I, Sr, Cs, Co, and Pd, and on their hydraulic properties. The samples were taken from the geologic formations in the area surrounding the projected waste repository in the Gorleben salt mine, at depth of up to 250 m down from terrain surface, and were analysed by means of column and batch experiments. Further goals were to determine the radionuclide migration as a function of flow velocity of the groundwater, and of sand compactness, as well as the effects of carrier ions and main groundwater contituents. The margins of retardation factors for the various radionuclides are given. One important result of the studies is that it could be expeimentally verified that there is the process of quasi irreversible sorption, i.e. it could be shown that desorption of radionuclides from natural, unconsolidated rock proceeds very much slowlier than sorption, so that this finding is of great significance to the safety assessment of a radioactive waste repository in geologic formations. (orig./HP) [de

  16. Near Field sorption Data Bases for Compacted MX-80 Bentonite for Performance Assessment of a High-Level Radioactive Waste Repository in Opalinus Clay Host Rock

    Energy Technology Data Exchange (ETDEWEB)

    Bradbury, M.; Baeyens, B

    2003-08-01

    Bentonites of various types and compacted forms are being investigated in many countries as backfill materials in high-level radioactive waste disposal concepts. Nagra is currently considering an Opalinus clay (OPA) formation in the Zuercher Weinland as a potential location for a high-level radioactive waste repository. A compacted MX-80 bentonite is foreseen as a potential backfill material. Performance assessment studies will be performed for this site and one of the requirements for such an assessment are sorption data bases (SDB) for the bentonite near-field. The purpose of this report is to describe the procedures used to develop the SDB. One of the pre-requisites for developing a SDB is a water chemistry for the compacted bentonite porewater. For a number of reasons mentioned in the report, and discussed in more detail elsewhere, this is not a straightforward task. There are considerable uncertainties associated with the major ion concentrations and in particular with the system pH and Eh. The MX-80 SDB was developed for a reference bentonite porewater (pH = 7.25) which was calculated using the reference OPA porewater. In addition, two further SDBs are presented for porewaters calculated at pH values of 6.9 and 7.9 corresponding to lower and upper bound values calculated for the range of groundwater compositions anticipated for the OPA host rock. 'In house' sorption isotherm data were measured for Cs(I), Ni(II), Eu(III), Th(IV), Se(IV) and 1(-1) on the 'as received' MX-80 material equilibrated with a simulated porewater composition. Complementary 'in house' sorption edge and isotherm measurements on conditioned Na/Ca montmorillonites were also available for many of these radionuclides. These data formed the core of the SDB. Nevertheless, some of the required sorption data still had to be obtained from the open literature. An important part of this report is concerned with describing selection procedures and the modifications

  17. DECOVALEX III/BENCHPAR PROJECTS. Evaluation of the Impact of Thermal-Hydro-Mechanical Couplings in Bentonite and Near-Field Rock Barriers on a Nuclear Waste Repository in a Sparsely Fractured Hard Rock. Report of BMT1C/WP2

    International Nuclear Information System (INIS)

    Jing, L.

    2005-02-01

    This report presents the works performed for the third, also the last, phase (BMT1C) of BMT1 of the DECOVALEX III project for the period of 1999-2002. The works of BMT1 is divided into three phases: BMT1A, BMT1B and BMT1C. The BMT1A concerns with calibration of the computer codes with a reference Thermal (T), Hydrological (H) and Mechanical (M) experiment at Kamaishi Mine, Japan. The objective is to validate the numerical approaches, computer codes and material models, so that the teams simulating tools are at a comparable level of maturity and sophistication. The BMT1B uses the calibrated codes to perform scoping calculations, considering varying degrees of THM coupling and varying permeability values of the surrounding rock for a reference generic repository design without fractures. The aim is to identify the coupling mechanisms of importance for construction, performance and safety of the repository. BMT1C concerns with scoping calculations with different coupling combinations for the case where a horizontal fracture intersects the deposition hole and a vertical fracture zone divides two adjacent deposition tunnel/hole system. A hydrostatic condition is applied along the vertical fracture as a hydraulic boundary condition. In addition, the SKI/KTH team performed an additional calculation case of a highly fractured rock mass with two orthogonal sets of fractures with a spacing of 0.5 m. The chosen measures for evaluating the long term safety and performance of the repository are the maximal temperature created by the thermal loading from the emplaced wastes, the time for resaturation of the buffer, the maximal swelling stress developed in the buffer, the structural integrity of the rock mass and the permeability evolution in the rock mass. The analyses fro BMT1C were conducted by four research teams: SKI/KTH (Sweden), CNSC (Canada), IRSN/CEA(France) and JNC (Japan), using FEM approach with different computer codes. From the results, it is clear that the

  18. DECOVALEX III/BENCHPAR PROJECTS. Evaluation of the Impact of Thermal-Hydro-Mechanical Couplings in Bentonite and Near-Field Rock Barriers on a Nuclear Waste Repository in a Sparsely Fractured Hard Rock. Report of BMT1C/WP2

    Energy Technology Data Exchange (ETDEWEB)

    Jing, L. [Royal Inst. of Technology, Stockholm (Sweden). Engineering Geology; Nguyen, T.S. [Canadian Nuclear Safety Commission, Ottawa, ON (Canada)] (eds.)

    2005-02-15

    This report presents the works performed for the third, also the last, phase (BMT1C) of BMT1 of the DECOVALEX III project for the period of 1999-2002. The works of BMT1 is divided into three phases: BMT1A, BMT1B and BMT1C. The BMT1A concerns with calibration of the computer codes with a reference Thermal (T), Hydrological (H) and Mechanical (M) experiment at Kamaishi Mine, Japan. The objective is to validate the numerical approaches, computer codes and material models, so that the teams simulating tools are at a comparable level of maturity and sophistication. The BMT1B uses the calibrated codes to perform scoping calculations, considering varying degrees of THM coupling and varying permeability values of the surrounding rock for a reference generic repository design without fractures. The aim is to identify the coupling mechanisms of importance for construction, performance and safety of the repository. BMT1C concerns with scoping calculations with different coupling combinations for the case where a horizontal fracture intersects the deposition hole and a vertical fracture zone divides two adjacent deposition tunnel/hole system. A hydrostatic condition is applied along the vertical fracture as a hydraulic boundary condition. In addition, the SKI/KTH team performed an additional calculation case of a highly fractured rock mass with two orthogonal sets of fractures with a spacing of 0.5 m. The chosen measures for evaluating the long term safety and performance of the repository are the maximal temperature created by the thermal loading from the emplaced wastes, the time for resaturation of the buffer, the maximal swelling stress developed in the buffer, the structural integrity of the rock mass and the permeability evolution in the rock mass. The analyses fro BMT1C were conducted by four research teams: SKI/KTH (Sweden), CNSC (Canada), IRSN/CEA(France) and JNC (Japan), using FEM approach with different computer codes. From the results, it is clear that the

  19. Rock mechanics evaluation of potential repository sites in the Paradox, Permian, and Gulf Coast Basins: Volume 1

    International Nuclear Information System (INIS)

    1987-09-01

    Thermal and thermomechanical analyses of a conceptual radioactive waste repository containing commercial and defense high-level wastes and spent fuel have been performing using finite element models. The thermal and thermomechanical responses of the waste package, disposal room, and repository regions were evaluated. four bedded salt formations, in Davis and Lavender Canyons in the Paradox Basin of southeastern Utah and in Deaf Smith and Swisher counties in the Permian Basin of northwestern Texas, and three salt domes, Vacherie Dome in northwestern Louisiana and Richton and Cypress Creek Domes in southeastern Mississippi, located in the Gulf Coast Basin, were examined. In the Paradox Basin, the pressure exerted on the waste package overpack was much greater than the initial in situ stress. The disposal room closure was less than 10 percent after 5 years. Surface uplift was nominal, and no significant thermomechanical perturbation of the aquitards was observed. In the Permian Basin, the pressure exerted on the waste package overpack was greater than the initial in situ stress. The disposal room closures were greater than 10 percent in less than 5 years. Surface uplift was nominal, and no significant thermomechanical perturbation of the aquitards was observed. In the Gulf Coast Basin, the pressure exerted on the waste package overpack was greater than the initial in situ stress. The disposal room closures were greater than 10 percent in less than 5 years. No significant thermomechanical perturbation of the overlying geology was observed. 40 refs., 153 figs., 32 tabs

  20. 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

  1. The projected environmental impacts of transportation of radioactive material to the first United States repository site

    International Nuclear Information System (INIS)

    Cashwell, J.W.; Neuhauser, K.S.; Reardon, P.C.; McNair, G.W.

    1987-01-01

    The relative national environmental impacts of transporting spent fuel and other nuclear wastes to each of 9 candidate repository sites in the United States were analyzed for the 26-year period of repository operation. Two scenarios were examined for each repository: 1) shipment of 5-year-old spent fuel and Defence High-Level Waste (DHLW) directly from their points of origin to a repository (reference case); and 2) shipment of 5-year-old spent fuel to a Monitored Retrievable Storage (MRS) facility and shipment (by dedicated rail) of 10-year-old consolidated spent fuel from the MRS to a repository. Transport by either all truck or all rail from the points of origin were analyzed as bounding cases. The computational system used to analyze these impacts included the WASTES II logistics code and the RADTRAN III risk analysis code. The radiological risks for the reference case increased as the total shipment miles to a repository increased for truck; the risks also increased with mileage for rail but at a lower rate. For the MRS scenario the differences between repository sites were less pronounced for both modal options, because of the reduction in total shipment miles possible with the large dedicated rail casks. All the risks reported are small in comparison to the radiological risks due to 'natural background'

  2. Impact of cementitious materials decalcification on transfer properties: application to radioactive waste deep repository

    International Nuclear Information System (INIS)

    Perlot, C.

    2005-09-01

    Cementitious materials have been selected to compose the engineering barrier system (EBS) of the French radioactive waste deep repository, because of concrete physico-chemical properties: the hydrates of the cementitious matrix and the pH of the pore solution contribute to radionuclides retention; furthermore the compactness of these materials limits elements transport. The confinement capacity of the system has to be assessed while a period at least equivalent to waste activity (up to 100.000 years). His durability was sustained by the evolution of transfer properties in accordance with cementitious materials decalcification, alteration that expresses structure long-term behavior. Then, two degradation modes were carried out, taking into account the different physical and chemical solicitations imposed by the host formation. The first mode, a static one, was an accelerated decalcification test using nitrate ammonium solution. It replicates the EBS alteration dues to underground water. Degradation kinetic was estimated by the amount of calcium leached and the measurement of the calcium hydroxide dissolution front. To evaluate the decalcification impact, samples were characterized before and after degradation in term of microstructure (porosity, pores size distribution) and of transfer properties (diffusivity, gas and water permeability). The influence of cement nature (ordinary Portland cement, blended cement) and aggregates type (lime or siliceous) was observed: experiments were repeated on different mortars mixes. On this occasion, an essential reflection on this test metrology was led. The second mode, a dynamical degradation, was performed with an environmental permeameter. It recreates the EBS solicitations ensured during the re-saturation period, distinguished by the hydraulic pressure imposed by the geologic layer and the waste exothermicity. This apparatus, based on triaxial cell functioning, allows applying on samples pressure drop between 2 and 10 MPa and

  3. An assessment of strontium sorption onto bentonite buffer material in waste repository.

    Science.gov (United States)

    Pathak, Pankaj

    2017-03-01

    In the present study, changes occurring in sorption characteristics of a representative bentonite (WIn-BT) exposed to SrCl 2 (0.001-0.1 M) under the pH range of 1-13 were investigated. Such interaction revealed a significant variation in surface charge density and binding energy of ions with respect to bentonite, and alteration in their physicochemical properties viz., specific surface area, cation exchange capacity, thermal and mechanical behaviour were observed. The distribution coefficients (k d ) calculated for sorption onto virgin (UCBT) and contaminated bentonite (CBT) indicated a greater influence of mineralogical changes occurred with variance of pH and strontium concentration. Notably, the sorption mechanism clearly elucidates the effect of structural negative charge and existence of anionic metal species onto CBT, and depicted the reason behind significant k d values at highly acidic and alkaline pH. The maximum k d of UCBT and CBT (0.001M SrCl2) were 8.99 and 2.92 L/kg, respectively, at the soil pH 8.5; whereas it was 2.37 and 1.23 L/kg at pH 1 for the CBT (0.1M SrCl2) and CBT (0.01M SrCl2) , respectively. The findings of this study can be useful to identify the physicochemical parameters of candidate buffer material and sorption reversibility in waste repository.

  4. Mechanical properties of buffer materials for repositories of high-level nuclear waste, 2

    International Nuclear Information System (INIS)

    Komine, Hideo; Ogata, Nobuhide

    1993-01-01

    Compacted bentonites have attracted increasing attention as back filling (buffer) materials for repositories of high-level nuclear waste. However, since little has been known concerning the swelling characteristics of compacted bentonites, it is necessary to clarify the fundamental swelling characteristics and the quantitative evaluation on this characteristics is required. For this purpose, a theoretical model concerning the swelling characteristics (swelling deformation and swelling pressure) of compacted bentonites were developed. The following conclusions were drawn from this theoretical study; (1) The evaluation formula of the swelling characteristics of compacted bentonites based on the diffuse double layer theory has been proposed by combining the theoretical model and the theoretical equation to estimate the swelling characteristics of a crystal. (2) The applicability of the evaluation formula proposed in this study has been confirmed by the comparison of the experimental results with calculated results. The sensitivity of this evaluation formula has also been investigated to find that the swelling characteristics is strongly dependent on the ion concentration of pore water and on the montmorillonite content of bentonite. (author)

  5. Performance of phosphoric acid activated montmorillonite as buffer materials for radioactive waste repository

    International Nuclear Information System (INIS)

    Wang, Tsing-Hai; Liu, Tsung-Ying; Wu, Ding-Chiang; Li, Ming-Hsu; Chen, Jiann-Ruey; Teng, Shi-Ping

    2010-01-01

    In this study, the performance of phosphoric acid activated montmorillonite (PAmmt) was evaluated by cesium ions adsorption experiments. The PAmmt samples were obtained by activating with 1, 3 and 5 mol L -1 of phosphoric acid, respectively under reflux for 3, 12, and 24 h. Experimental results demonstrated that the treatment of raw K-10 montmorillonite with phosphoric acid increased the materials' affinity for Cs uptake and no significant amount of suspension solids were produced. A relatively insignificant variation in the CEC value was observed. Furthermore, PAmmt also showed high adsorption selectivity toward Cs ions. The improved sorptive properties were mainly related to the increased surface area and the relatively higher surface charge density. Increased specific surface area was the resulted from partial decomposition of lamellar structure of mmt; while the higher surface charge density was caused by the protonation of octahedral Al-OH sites during the acid activation. Generally speaking, stronger acid concentration and longer activation times would produce relatively more decomposed PAmmt particles. However, as the activation exceeds 3 h, the precipitation of Si 4+ would passivate PAmmt against further acid attacks. Based upon our results, acid activation by phosphoric acid could produce PAmmt samples with high sorption capacity and selectivity, and good structural integrity, which are beneficial to be used at radioactive waste repository.

  6. Thermal dimensioning of the deep repository. Influence of canister spacing, canister power, rock thermal properties and nearfield design on the maximum canister surface temperature

    International Nuclear Information System (INIS)

    Hoekmark, Harald; Faelth, Billy

    2003-12-01

    The report addresses the problem of the minimum spacing required between neighbouring canisters in the deep repository. That spacing is calculated for a number of assumptions regarding the conditions that govern the temperature in the nearfield and at the surfaces of the canisters. The spacing criterion is that the temperature at the canister surfaces must not exceed 100 deg C .The results are given in the form of nomographic charts, such that it is in principle possible to determine the spacing as soon as site data, i.e. the initial undisturbed rock temperature and the host rock heat transport properties, are available. Results of canister spacing calculations are given for the KBS-3V concept as well as for the KBS-3H concept. A combination of numerical and analytical methods is used for the KBS-3H calculations, while the KBS-3V calculations are purely analytical. Both methods are described in detail. Open gaps are assigned equivalent heat conductivities, calculated such that the conduction across the gaps will include also the heat transferred by radiation. The equivalent heat conductivities are based on the emissivities of the different gap surfaces. For the canister copper surface, the emissivity is determined by back-calculation of temperatures measured in the Prototype experiment at Aespoe HRL. The size of the different gaps and the emissivity values are of great importance for the results and will be investigated further in the future

  7. Hydrologic issues in repository siting

    International Nuclear Information System (INIS)

    Remson, I.; Gorelick, S.M.

    1982-01-01

    Extrapolation of Darcy's law to the transport of water an solutes in unfractured poorly permeable rocks being studied for nuclear waste disposal is questioned. The hydrologic literature includes numerous references to both non-Darcian flow in dense materials devoid of macrofractures and microfractures and to threshold gradients below which no flow occurs. For such situations to occur, the pore-size range must be small enough so that all pore water is sufficiently close to mineral surfaces to be affected by the surficial forces. Then the flow will be non-Newtonian and non-Darcian, and solute transport will be by molecular diffusion. If fluid transport in very dense unfractured rocks is non-Darcian, useful methods of testing candidate host rocks become apparent. In situ nondestructive pressure testing of canister waste emplacement boreholes in a mined repository can verify the absence of both fracture flow and Darcian flow. 18 references

  8. Deep repository - engineered barrier systems. Assessment of backfill materials and methods for deposition tunnels

    International Nuclear Information System (INIS)

    Gunnarsson, David; Moren, Lena; Sellin, Patrik; Keto, Paula

    2006-09-01

    The main objectives of this report are to: 1) present density criteria considering deposition tunnels for the investigated backfill materials, 2) evaluate what densities can be achieved with the suggested backfill methods, 3) compare the density criteria to achievable densities, 4) based on this comparison evaluate the safety margin for the combinations of backfill materials and methods and, 5) make recommendations for further investigations and development work. The backfilling methods considered in this report are compaction of backfill material in situ in the tunnel and placement of pre-compacted blocks and pellets. The materials investigated in the second phase of the SKB-Posiva backfilling project can be divided into three main categories: 1. Bentonite clays: two high-grade Na-bentonites from Wyoming (MX-80 and SPV200), one low-grade bentonite from Kutch (India Asha 230), and one high and one low-grade Ca-bentonite from Milos (Deponite CA-N and Milos backfill). The high-grade bentonites are used in different bentonite-ballast mixtures. 2. Smectite-rich mixed-layer clays: one from Dnesice-Plzensko Jih (DPJ) located in the Czech Republic and one from Northern Germany (Friedland clay). Mixtures of bentonite and ballast: Mixtures consisting of high-grade bentonite (0, 40 and 50 w-%) and crushed rock with different type of grain size distribution or sand. The relationships between dry densities and hydraulic conductivity, swelling pressure and compressibility in saturated state for these materials were investigated. Most of the tests were performed with a groundwater salinity of 3.5%. This salinity is comparable to sea water and can be expected to be at the high end of salinities occurring during the assessment period. The purpose of the investigations was to determine the dry densities required to meet the function indicator criteria. These densities are referred to as the density criteria. However throughout the assessment period a loss of material and thus

  9. Deep repository - engineered barrier systems. Assessment of backfill materials and methods for deposition tunnels

    Energy Technology Data Exchange (ETDEWEB)

    Gunnarsson, David; Moren, Lena; Sellin, Patrik [Swedish Nuclear Fuel and Waste Management Co., Stockholm (Sweden); Keto, Paula [Saanio and Riekkola Oy, Helsinki (Finland)

    2006-09-15

    The main objectives of this report are to: 1) present density criteria considering deposition tunnels for the investigated backfill materials, 2) evaluate what densities can be achieved with the suggested backfill methods, 3) compare the density criteria to achievable densities, 4) based on this comparison evaluate the safety margin for the combinations of backfill materials and methods and, 5) make recommendations for further investigations and development work. The backfilling methods considered in this report are compaction of backfill material in situ in the tunnel and placement of pre-compacted blocks and pellets. The materials investigated in the second phase of the SKB-Posiva backfilling project can be divided into three main categories: 1. Bentonite clays: two high-grade Na-bentonites from Wyoming (MX-80 and SPV200), one low-grade bentonite from Kutch (India Asha 230), and one high and one low-grade Ca-bentonite from Milos (Deponite CA-N and Milos backfill). The high-grade bentonites are used in different bentonite-ballast mixtures. 2. Smectite-rich mixed-layer clays: one from Dnesice-Plzensko Jih (DPJ) located in the Czech Republic and one from Northern Germany (Friedland clay). Mixtures of bentonite and ballast: Mixtures consisting of high-grade bentonite (0, 40 and 50 w-%) and crushed rock with different type of grain size distribution or sand. The relationships between dry densities and hydraulic conductivity, swelling pressure and compressibility in saturated state for these materials were investigated. Most of the tests were performed with a groundwater salinity of 3.5%. This salinity is comparable to sea water and can be expected to be at the high end of salinities occurring during the assessment period. The purpose of the investigations was to determine the dry densities required to meet the function indicator criteria. These densities are referred to as the density criteria. However throughout the assessment period a loss of material and thus

  10. 30 CFR 717.15 - Disposal of excess rock and earth materials on surface areas.

    Science.gov (United States)

    2010-07-01

    ... 30 Mineral Resources 3 2010-07-01 2010-07-01 false Disposal of excess rock and earth materials on surface areas. 717.15 Section 717.15 Mineral Resources OFFICE OF SURFACE MINING RECLAMATION AND ENFORCEMENT, DEPARTMENT OF THE INTERIOR INITIAL PROGRAM REGULATIONS UNDERGROUND MINING GENERAL PERFORMANCE STANDARDS § 717.15 Disposal of excess rock and...

  11. Chemical buffering capacity of clay rock

    International Nuclear Information System (INIS)

    Beaucaire, C.; Pearson, F.J.; Gautschi, A.

    2004-01-01

    The long-term performance of a nuclear waste repository is strongly dependent on the chemical properties of the host rock. The host rock establishes the chemical environment that determines such important performance attributes as radionuclide solubilities from the waste and the transport rates from the repository to the accessible environment. Clay-rich rocks are especially favourable host rocks because they provide a strong buffering capacity to resist chemical changes prompted either internally, by reactions of the waste itself and emplacement materials, or externally, by changes in the hydrologic systems surrounding the host rock. This paper will focus on three aspects of the stability of clay-rich host rocks: their ability to provide pCO 2 and redox buffering, and to resist chemical changes imposed by changes in regional hydrology and hydro-chemistry. (authors)

  12. Influence of supplementary cementitious materials on the properties of concrete for secondary protection barrier in radioactive waste repositories

    Czech Academy of Sciences Publication Activity Database

    Koťátková, J.; Čáchová, M.; Bezdička, Petr; Vejmelková, E.; Konvalinka, P.; Zemanová, L.; Černý, R.

    2018-01-01

    Roč. 760, SI (2018), s. 96-101 ISSN 1662-9795. [Special Concrete and Composites 2017 /14./. Lísek, 10.10.2017-11.10.2017] R&D Projects: GA ČR(CZ) GA17-11635S Institutional support: RVO:61388980 Keywords : Basic physical properties * Mechanical properties * Repository * Secondary protection barrier * Supplementary cementitious materials * Thermal properties Subject RIV: CA - Inorganic Chemistry OBOR OECD: Inorganic and nuclear chemistry

  13. Experimental study of polyurethane foam reinforced soil used as a rock-like material

    Directory of Open Access Journals (Sweden)

    Eren Komurlu

    2015-10-01

    Full Text Available In this study, polyurethane foam type thermoset polymerizing, due to chemical reaction between its liquid ingredients, was tested as binder after solidifying and then a rock-like material mixing with a sandy silt type soil was prepared. The uniaxial compressive strengths (UCSs of polyurethane foam reinforced soil specimens were determined for different polyurethane ratios in the mixture. Additionally, a series of tests on slake durability, impact value, freezing–thawing resistance, and abrasion resistance of polyurethane reinforced soil (PRS mixture was conducted. The UCS values over 3 MPa were measured from the PRS specimens. The testing results showed that treated soil can economically become a desirable rock-like material in terms of slake durability and resistances against freezing–thawing, impact effect and abrasion. As another characteristic of the rock-like material made with polyurethane foam, unit volume weight was found to be quite lower than those of natural rock materials.

  14. Deep repository - engineered barrier systems. Assessment of backfill materials and methods for deposition tunnels

    International Nuclear Information System (INIS)

    Gunnarsson, D.; Moren, L.; Sellin, P; Keto, P.

    2007-09-01

    The main objectives of this report are to: (1) present density criteria considering deposition tunnels for the investigated backfill materials, (2) evaluate what densities can be achieved with the suggested backfill methods, (3) compare the density criteria to achievable densities, (4) based on this comparison evaluate the safety margin for the combinations of backfill materials and methods and, (5) make recommendations for further investigations and development work. The backfilling methods considered in this report are compaction of backfill material in situ in the tunnel and placement of pre-compacted blocks and pellets. The materials investigated in the second phase of the SKB-Posiva backfilling programme can be divided into three main categories: (1) Bentonite clays: two high-grade Na-bentonites from Wyoming (MX-80 and SPV200), one low-grade bentonite from Kutch (India Asha 2 0), and one high- and one low-grade Ca-bentonite from Milos (Deponite CA-N and Milos backfill). The highgrade bentonites are used in different bentonite-ballast mixtures. (2) Smectite-rich mixed-layer clays: one from Dnesice-Plzensko Jih (DPJ) located in the Czech Republic and one from Northern Germany (Friedland clay). (3) Mixtures of bentonite and ballast: Mixtures consisting of high-grade bentonite (30, 40 and 50 w-%) and crushed rock with different type of grain size distribution or sand. The general conclusion from the comparison between estimated achievable densities and the density criteria is that placing pre-compacted blocks of swelling clay or 50/50 mixture and pellets in the tunnel results in the highest safety margin. (orig.)

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  16. Characterization of the material produced using marble waste and reagents aiminig production of rock wool

    International Nuclear Information System (INIS)

    Rodrigues, Girley Ferreira; Espinosa, Denise Crocce Romano; Tenorio, Jorge Alberto Soares; Alves, Joner Oliveira

    2010-01-01

    The aim of this work was to characterize materials produced from the mixture of marble waste and chemical reagents. The materials were homogenized, melted and cooled in order to obtain materials with similar characteristics of rock wools. The batch was poured in a water-filled recipient and also in a Herty viscometer at three temperatures. Samples of produced materials were characterized by X-ray diffraction, scanning electron microscopy and differential thermal analysis. Results of this study indicate that it is possible the incorporation of marble waste in the production process of rock wool, replacing approximately 15% of the raw material used to fabricate this material. This process represents a technological breakthrough since it allows the reuse of marble waste, and also represents a possible decrease in rock wool production cost, which is a material with a growing market as thermo acoustic insulator. (author)

  17. The Evaluation of Material Properties of Low-pH Cement Grout for the Application of Cementitious Materials to Deep Radioactive Waste Repository Tunnels

    International Nuclear Information System (INIS)

    Kim, Jin Seop; Kwon, S. K.; Cho, W. J.; Kim, G. W.

    2009-12-01

    Considering the current construction technology and research status of deep repository tunnels for radioactive waste disposal, it is inevitable to use cementitious materials in spite of serious concern about their long-term environmental stability. Thus, it is an emerging task to develop low pH cementitious materials. This study reviews the state of the technology on low pH cements developed in Sweden, Switzerland, France, and Japan as well as in Finland which is constructing a real deep repository site for high-level radioactive waste disposal. Considering the physical and chemical stability of bentonite which acts as a buffer material, a low pH cement limits to pH ≤11 and pozzolan-type admixtures are used to lower the pH of cement. To attain this pH requirement, silica fume, which is one of the most promising admixtures, should occupy at least 40 wt% of total dry materials in cement and the Ca/Si ratio should be maintained below 0.8 in cement. Additionally, selective super-plasticizer needs to be used because a high amount of water is demanded from the use of a large amount of silica fume. In this report, the state of the technology on application of cementitious materials to deep repository tunnels for radioactive waste disposal was analysed. And the material properties of low-pH and high-pH cement grouts were evaluated base on the grout recipes of ONKALO in Finlan

  18. Stress Wave Propagation in Viscoelastic-Plastic Rock-Like Materials

    Directory of Open Access Journals (Sweden)

    Liu Lang

    2016-05-01

    Full Text Available Rock-like materials are composites that can be regarded as a mixture composed of elastic, plastic, and viscous components. They exhibit viscoelastic-plastic behavior under a high-strain-rate loading according to element model theory. This paper presents an analytical solution for stress wave propagation in viscoelastic-plastic rock-like materials under a high-strain-rate loading and verifies the solution through an experimental test. A constitutive equation of viscoelastic-plastic rock-like materials was first established, and then kinematic and kinetic equations were then solved to derive the analytic solution for stress wave propagation in viscoelastic-plastic rock-like materials. An experimental test using the SHPB (Split Hopkinson Pressure Bar for a concrete specimen was conducted to obtain a stress-strain curve under a high-strain-rate loading. Inverse analysis based on differential evolution was conducted to estimate undetermined variables for constitutive equations. Finally, the relationship between the attenuation factor and the strain rate in viscoelastic-plastic rock-like materials was investigated. According to the results, the frequency of the stress wave, viscosity coefficient, modulus of elasticity, and density play dominant roles in the attenuation of the stress wave. The attenuation decreases with increasing strain rate, demonstrating strongly strain-dependent attenuation in viscoelastic-plastic rock-like materials.

  19. DISPERSION AND SORPTION CHARACTERISTICS OF URANIUM IN THE ZEOLITE-QUARTZ MIXTURE AS BACKFILL MATERIAL IN THE RADIOACTIVE WASTE REPOSITORY

    Directory of Open Access Journals (Sweden)

    Herry Poernomo

    2010-06-01

    Full Text Available The experiment of sorption and dispersion characteristics of uranium in the zeolite-quartz mixture as candidate of raw material of backfill material in the radioactive waste repository has been performed. The objective is to know the effect of zeolite and quartz grain size on the zeolite-to-quartz weight ratio that gives porosity (ε, permeability (K, and dispersivity (α of uranium in the zeolite-quartz mixture as backfill material. The experiment was carried out by fixed bed method in the column filled by the zeolite-quartz mixture with zeolite-to-quartz weight percent ratio of 100/0, 80/20, 60/40, 40/60, 20/80, 0/100 wt. % in the water saturated condition flowed by uranyl nitrate solution of 500 ppm concentration (Co as uranium simulation which was leached from immobilized radioactive waste in the repository. The concentration of uranium in the effluents represented as Ct were analyzed by spectrophotometer Corning Colorimeter 253 every 15 minutes, then using Co and Ct uranium dispersivity (α in the backfill material was determined. The experiment data shown that 0.196 mm particle size of zeolite and 0.116 mm particle size of quartz on the zeolite-to-quartz weight ratio of 60/40 wt. % with ε = 0.678, K = 3.345x10-4 cm/second, and α = 0.759 cm can be proposed as candidate of raw material of backfill material in the radioactive waste repository.   Keywords: backfill material, quartz, radioactive waste, zeolite

  20. Researching radioactive waste disposal. [Underground repository

    Energy Technology Data Exchange (ETDEWEB)

    Feates, F; Keen, N [UKAEA Research Group, Harwell. Atomic Energy Research Establishment

    1976-02-16

    At present it is planned to use the vitrification process to convert highly radioactive liquid wastes, arising from nuclear power programme, into glass which will be contained in steel cylinders for storage. The UKAEA in collaboration with other European countries is currently assessing the relative suitability of various natural geological structures as final repositories for the vitrified material. The Institute of Geological Sciences has been commissioned to specify the geological criteria that should be met by a rock structure if it is to be used for the construction of a repository though at this stage disposal sites are not being sought. The current research programme aims to obtain basic geological data about the structure of the rocks well below the surface and is expected to continue for at least three years. The results in all the European countries will then be considered so that the United Kingdom can choose a preferred method for isolating their wastes. It is only at that stage that a firm commitment may be made to select a site for a potential repository, when a far more detailed scientific research study will be instituted. Heat transfer problems and chemical effects which may occur within and around repositories are being investigated and a conceptual design study for an underground repository is being prepared.

  1. Assessment of rock wool as support material for on-site sanitation: hydrodynamic and mechanical characterization.

    Science.gov (United States)

    Wanko, Adrien; Laurent, Julien; Bois, Paul; Mosé, Robert; Wagner-Kocher, Christiane; Bahlouli, Nadia; Tiffay, Serge; Braun, Bouke; Provo kluit, Pieter-Willem

    2016-01-01

    This study proposes mechanical and hydrodynamic characterization of rock wool used as support material in compact filter. A double-pronged approach, based on experimental simulation of various physical states of this material was done. First of all a scanning electron microscopy observation allows to highlight the fibrous network structure, the fibres sizing distribution and the atomic absorption spectrum. The material was essentially lacunar with 97 ± 2% of void space. Static compression tests on variably saturated rock wool samples provide the fact that the strain/stress behaviours depend on both the sample conditioning and the saturation level. Results showed that water exerts plastifying effect on mechanical behaviour of rock wool. The load-displacement curves and drainage evolution under different water saturation levels allowed exhibiting hydraulic retention capacities under stress. Finally, several tracer experiments on rock wool column considering continuous and batch feeding flow regime allowed: (i) to determine the flow model for each test case and the implications for water dynamic in rock wool medium, (ii) to assess the rock wool double porosity and discuss its advantages for wastewater treatment, (iii) to analyse the benefits effect for water treatment when the high level of rock wool hydric retention was associated with the plug-flow effect, and (iv) to discuss the practical contributions for compact filter conception and management.

  2. Cs sorption to potential host rock of low-level radioactive waste repository in Taiwan: experiments and numerical fitting study.

    Science.gov (United States)

    Wang, Tsing-Hai; Chen, Chin-Lung; Ou, Lu-Yen; Wei, Yuan-Yaw; Chang, Fu-Lin; Teng, Shi-Ping

    2011-09-15

    A reliable performance assessment of radioactive waste repository depends on better knowledge of interactions between nuclides and geological substances. Numerical fitting of acquired experimental results by the surface complexation model enables us to interpret sorption behavior at molecular scale and thus to build a solid basis for simulation study. A lack of consensus on a standard set of assessment criteria (such as determination of sorption site concentration, reaction formula) during numerical fitting, on the other hand, makes lower case comparison between various studies difficult. In this study we explored the sorption of cesium to argillite by conducting experiments under different pH and solid/liquid ratio (s/l) with two specific initial Cs concentrations (100mg/L, 7.5 × 10(-4)mol/L and 0.01 mg/L, 7.5 × 10(-8)mol/L). After this, numerical fitting was performed, focusing on assessment criteria and their consequences. It was found that both ion exchange and electrostatic interactions governed Cs sorption on argillite. At higher initial Cs concentration the Cs sorption showed an increasing dependence on pH as the solid/liquid ratio was lowered. In contrast at trace Cs levels, the Cs sorption was neither s/l dependent nor pH sensitive. It is therefore proposed that ion exchange mechanism dominates Cs sorption when the concentration of surface sorption site exceeds that of Cs, whereas surface complexation is attributed to Cs uptake under alkaline environments. Numerical fitting was conducted using two different strategies to determine concentration of surface sorption sites: the clay model (based on the cation exchange capacity plus surface titration results) and the iron oxide model (where the concentration of sorption sites is proportional to the surface area of argillite). It was found that the clay model led to better fitting than the iron oxide model, which is attributed to more amenable sorption sites (two specific sorption sites along with larger site

  3. Proposed format and content of license applications for deep geologic terminal repositories for radioactive material

    International Nuclear Information System (INIS)

    1978-01-01

    Chapters are devoted to the following: introduction and general description; summary safety analysis; site characteristics; principal design criteria; repository design; operations systems; management of onsite generated waste; radiation protection; accident safety analysis; conduct of operations; operating controls and limits; and quality assurance

  4. FEBEX project: full-scale engineered barriers experiment for a deep geological repository for high level radioactive waste in crystalline host rock. Final report

    International Nuclear Information System (INIS)

    Alberdi, J.; Barcala, J. M.; Campos, R.; Cuevas, A. M.; Fernandez, E.

    2000-01-01

    FEBEX has the multiple objective of demonstrating the feasibility of manufacturing, handling and constructing the engineered barriers and of developing codes for the thermo-hydro-mechanical and thermo-hydro-geochemical performance assessment of a deep geological repository for high level radioactive wastes. These objectives require integrated theoretical and experimental development work. The experimental work consists of three parts: an in situ test, a mock-up test and a series of laboratory tests. The experiments is based on the Spanish reference concept for crystalline rock, in which the waste capsules are placed horizontally in drifts surround by high density compacted bentonite blocks. In the two large-scale tests, the thermal effects of the wastes were simulated by means of heaters; hydration was natural in the in situ test and controlled in the mock-up test. The large-scale tests, with their monitoring systems, have been in operation for more than two years. the demonstration has been achieved in the in situ test and there are great expectation that numerical models sufficiently validated for the near-field performance assessment will be achieved. (Author)

  5. Deep ground water microbiology in Swedish granite rock and it's relevance for radio-nuclide migration from a Swedish high level nuclear waste repository

    International Nuclear Information System (INIS)

    Pedersen, Karsten

    1989-03-01

    Data on numbers, species and activity of deep ground water microbial populations in Swedish granite rock have been collected. Specific studies are performed on radio-nuclid uptake on bacteria judge to be probable inhabitants in Swedish nuclear waste repositories. An integrated mobile field laboratory was used for water sampling and for the immediate counting and inoculation of the samples from boreholes at levels between 129 and 860 m. A sampler adapted for the collection of undisturbed samples for gas analysis was used to collect samples for bacterial enumerations and enrichments. The sampler can be opened and closed from the surface at the actual sampling depth. The samples can subsequently be brought to the surface without contact with air and with the pressure at the actual sampling depth. The number of bacteria were determined in samples from the gas sampler when this was possible. Else numbers are determined in the water that is pumped up to the field lab. The average total number of bacteria is 3 x 10 5 bacterial ml -1 . The number of bacteria possible to recover with plate count arrays from 0.10 to 21.9%. (author)

  6. FEBEX project: full-scale engineered barriers experiment for a deep geological repository for high level radioactive waste in crystalline host rock

    Energy Technology Data Exchange (ETDEWEB)

    Alberid, J; Barcala, J M; Campos, R; Cuevas, A M; Fernandez, E [Ciemat. Madrid (Spain)

    2000-07-01

    FEBEX has the multiple objective of demonstrating the feasibility of manufacturing, handling and constructing the engineered barriers and of developing codes for the thermo-hydro-mechanical and thermo-hydro-geochemical performance assessment of a deep geological repository for high level radioactive wastes. These objectives require integrated theoretical and experimental development work. The experimental work consists of three parts: an in situ test, a mock-up test and a series of laboratory tests. The experiments is based on the Spanish reference concept for crystalline rock, in which the waste capsules are placed horizontally in drifts surround by high density compacted bentonite blocks. In the two large-scale tests, the thermal effects of the wastes were simulated by means of heaters; hydration was natural in the in situ test and controlled in the mock-up test. The large-scale tests, with their monitoring systems, have been in operation for more than two years. the demonstration has been achieved in the in situ test and there are great expectation that numerical models sufficiently validated for the near-field performance assessment will be achieved. (Author)

  7. Migration of fluids as a tool to evaluate the feasibility of the implantation of geological radioactive wastes repositories (RARN) in granitoid rocks: tests on granites submitted to natural deformation vs. not deformed

    International Nuclear Information System (INIS)

    Lopes, Nilo Henrique Balzani; Barbosa, Pedro Henrique Silva; Santos, Alanna Leite dos; Amorim, Lucas Eustáquio Dias; Freitas, Mônica Elizetti de; Rios, Francisco Javier

    2017-01-01

    Fluid composition and migration studies in granitoid rocks subjected to deformation events are a factor that should be considered in the selection of geologically favorable areas for RANR construction, and may be an excellent complement to engineering barrier designs. The research objective was to develop an academic approach, comparing the behavior of deformed and non-deformed granites, not being related to any CNEN project of deploying repositories. It is concluded that in the choice of suitable sites for the construction of repositories, granite bodies that are submitted to metamorphic / deformation / hydrothermal events or that are very fractured should be disregarded. The domes of granite batholith that have undergone hydraulic billing should also be discarded. It has been found that, because of the warming caused by radioactive decay reactions, there is a real possibility that the release of potentially abrasive fluids contained in the minerals can reach and corrode the walls of the repositories and / or packaging

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

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  9. Localized corrosion of metallic materials and γ radiation effects in passive layers under simulated radwaste repository conditions. Final report

    International Nuclear Information System (INIS)

    Schultze, J.W.; Kudelka, S.; Michaelis, A.; Schweinsberg, M.; Thies, A.

    1996-02-01

    The task of the project was to simulate the conditions in a radwaste repository and to perform local analyses in order to detect the critical conditions and material susceptibilities leading to localized corrosion of materials. The information thus obtained was to yield more precise data on the long-term stability of materials for the intended purpose, in order to be able to appropriately select or optimize the materials (Ti, TiO.2Pd, Hastelloy C4, fine-grained structural steel). A major aspect to be examined was natural inhomogeneities of the electrode surfaces, as determined by the grain structure of the selected materials. Thus a laterally inhomogeneous composition in the welded zone induces an inhomogeneous current distribution, and hence strong susceptibility to localized corrosion. This effect was to be quantified, and the localized corrosion processes had to be identified by means of novel, electrochemical methods with a resolution power of μm. The investigations were to be made under conditions as near to practice as possible, for instance by simulating radwaste repository conditions and performing measurements at elevated temperatures (170 C) in an autoclave. Another task was to examine the radiation effects of γ radiation on passive layers, and describe the possible modifications induced by recrystallisation, photocorrosion, or oxide formation. (orig./MM) [de

  10. Far Field Sorption Data Bases for Performance Assessment of a High-Level Radioactive Waste Repository in an Undisturbed Opalinus Clay Host Rock

    International Nuclear Information System (INIS)

    Bradburry, M.; Baeyens, B.

    2003-08-01

    An Opalinus Clay formation in the Zuercher Weinland is under consideration by Nagra as a potential location for a high-level and long-Iived intermediate-level radioactive waste repository. Performance assessment studies will be performed for this site and the purpose of this report is to describe the procedures used to develop sorption data bases appropriate for an undisturbed Opalinus Clay host rock which are required for such safety analysis calculations. In tight, low water content argillaceous rock formations such as Opalinus Clay, there is uncertainty concerning the in situ pH/P CO 2 . In order to take this intrinsic uncertainty into account porewater chemistries were calculated for a reference case, pH = 7.24, and for two other pH values, 6.3 and 7.8. Sorption data bases are given for the three cases. The basis for the sorption data bases is 'in-house' sorption measurements for Cs(I), Sr(II), Ni(II), Eu(III), Sn(IV), Se(IV), Th(IV) and I(-I) carried out on Opalinus Clay samples from Mont Terri (Canton Jura) since at the time the experiments were performed no core samples from the Benken borehole (Zuercher Weinland) were available. The Opalinus Clay at Mont Terri and Benken are part of the same geological formation . Despite having directly measured data for the above key radionuclides, some of the required distribution ratios (Rd) used to generate the sorption data bases still came from the open literature. An important part of this report is concerned with describing the procedures whereby these selected literature Rd values were modified so as to apply to the Benken Opalinus Clay mineralogy and groundwater chemistries calculated at the three pH values given above. The resulting Rd values were then further modified using so-called Lab→Field transfer factors to produce sorption values which were appropriate to the in situ bulk rock for the selected range of water chemistry conditions. Finally, it is important to have some appreciation of the uncertainties

  11. Far Field Sorption Data Bases for Performance Assessment of a High-Level Radioactive Waste Repository in an Undisturbed Opalinus Clay Host Rock

    Energy Technology Data Exchange (ETDEWEB)

    Bradburry, M.; Baeyens, B

    2003-08-01

    An Opalinus Clay formation in the Zuercher Weinland is under consideration by Nagra as a potential location for a high-level and long-Iived intermediate-level radioactive waste repository. Performance assessment studies will be performed for this site and the purpose of this report is to describe the procedures used to develop sorption data bases appropriate for an undisturbed Opalinus Clay host rock which are required for such safety analysis calculations. In tight, low water content argillaceous rock formations such as Opalinus Clay, there is uncertainty concerning the in situ pH/P{sub CO{sub 2}}. In order to take this intrinsic uncertainty into account porewater chemistries were calculated for a reference case, pH = 7.24, and for two other pH values, 6.3 and 7.8. Sorption data bases are given for the three cases. The basis for the sorption data bases is 'in-house' sorption measurements for Cs(I), Sr(II), Ni(II), Eu(III), Sn(IV), Se(IV), Th(IV) and I(-I) carried out on Opalinus Clay samples from Mont Terri (Canton Jura) since at the time the experiments were performed no core samples from the Benken borehole (Zuercher Weinland) were available. The Opalinus Clay at Mont Terri and Benken are part of the same geological formation . Despite having directly measured data for the above key radionuclides, some of the required distribution ratios (Rd) used to generate the sorption data bases still came from the open literature. An important part of this report is concerned with describing the procedures whereby these selected literature Rd values were modified so as to apply to the Benken Opalinus Clay mineralogy and groundwater chemistries calculated at the three pH values given above. The resulting Rd values were then further modified using so-called Lab{yields}Field transfer factors to produce sorption values which were appropriate to the in situ bulk rock for the selected range of water chemistry conditions. Finally, it is important to have some

  12. Comparison of Crack Initiation, Propagation and Coalescence Behavior of Concrete and Rock Materials

    Science.gov (United States)

    Zengin, Enes; Abiddin Erguler, Zeynal

    2017-04-01

    There are many previously studies carried out to identify crack initiation, propagation and coalescence behavior of different type of rocks. Most of these studies aimed to understand and predict the probable instabilities on different engineering structures such as mining galleries or tunnels. For this purpose, in these studies relatively smaller natural rock and synthetic rock-like models were prepared and then the required laboratory tests were performed to obtain their strength parameters. By using results provided from these models, researchers predicted the rock mass behavior under different conditions. However, in the most of these studies, rock materials and models were considered as contains none or very few discontinuities and structural flaws. It is well known that rock masses naturally are extremely complex with respect to their discontinuities conditions and thus it is sometimes very difficult to understand and model their physical and mechanical behavior. In addition, some vuggy rock materials such as basalts and limestones also contain voids and gaps having various geometric properties. Providing that the failure behavior of these type of rocks controlled by the crack initiation, propagation and coalescence formed from their natural voids and gaps, the effect of these voids and gaps over failure behavior of rocks should be investigated. Intact rocks are generally preferred due to relatively easy side of their homogeneous characteristics in numerical modelling phases. However, it is very hard to extract intact samples from vuggy rocks because of their complex pore sizes and distributions. In this study, the feasibility of concrete samples to model and mimic the failure behavior vuggy rocks was investigated. For this purpose, concrete samples were prepared at a mixture of %65 cement dust and %35 water and their physical and mechanical properties were determined by laboratory experiments. The obtained physical and mechanical properties were used to

  13. Investigation on natural radioactive nuclide contents of rock products in Xi'an construction materials market

    International Nuclear Information System (INIS)

    Zhou Chunlin; Han Feng; Shang Aiguo; Li Tiantuo; Guo Huiping; Yie Lichao; Li Guifang

    2001-01-01

    The author reports the investigation results on natural radioactive nuclide contents of rock products from Xi'an construction materials market. The products were classified according to the national standard. The results show that natural radioactive nuclide contents in sampled rock products are in normal radioactive background levels. The radio-activity ranges of 238 U, 226 Ra, 232 Th and 40 K are 2.7 - 181.8, 0.92 - 271.0, 0.63 - 148.0, 1.8 - 1245 Bq·kg -1 , respectively. According to the national standard (JC 518-93), the application of some rock products must be limited

  14. A study on nuclide migration in buffer materials and rocks for geological disposal of radioactive waste

    International Nuclear Information System (INIS)

    Sato, Haruo

    1998-01-01

    This thesis summarizes the results investigated in order to establish a basic theory on the predictive method of diffusion coefficients of nuclides in compacted sodium bentonite which is a candidate buffer material and in representative rocks for the geological disposal of radioactive waste by measuring the pore structural factors of the compacted bentonite and rocks such as porosity and tortuosity, measuring diffusion coefficients of nuclides in the bentonite and rocks, acquiring basic data on diffusion and developing diffusion models which can quantitatively predict nuclide migration in long-term. (J.P.N.). 117 refs

  15. Effect of leachate of cementitious materials on the geological media. Experimental study of the influence of high pH plume on rock

    International Nuclear Information System (INIS)

    Kato, Hiroshige; Sato, Mitsuyoshi; Owada, Hitoshi; Mihara, Morihiro; Ohi, Takao

    2000-05-01

    Cementitious materials will be used in TRU waste disposal repository. In such cases, it is considered that the migration of alkaline leachates from cementitious materials, so called high pH plume, will cause dissolution of rock and precipitation of secondary minerals. In addition, the high pH plume will move along the flow of groundwater, so it is predicted that rock formation and components of high pH groundwater vary with time and space. However, time and spatial dependence of the variations of secondary minerals and groundwater components has not been clarified. In order to acquire the data of variations of secondary minerals and groundwater components, we carried out the rock alteration experiments with column method. The crushed granodiorite was filled into 4 meters length column (φ 3.7 cm) and artificial cement leachate (pH=13.3; Na=0.1mol/l, K=0.1mol/l, Ca=0.002mol/l) was streamed at flow rates of 0.1 ml/min for 7 months at 80degC. As the result, secondary minerals confirmed on the rock were calcite and C-S-H at upstream of column and C-S-H at mid-downstream. The pH value of the fluid dominated by Na and K did not be decreased by reaction with the rock. In this study, the data relating to the effect of high pH plume on rock over the long term was acquired. (author)

  16. The importance of stress percolation patterns in rocks and other polycrystalline materials.

    Science.gov (United States)

    Burnley, P C

    2013-01-01

    A new framework for thinking about the deformation behavior of rocks and other heterogeneous polycrystalline materials is proposed, based on understanding the patterns of stress transmission through these materials. Here, using finite element models, I show that stress percolates through polycrystalline materials that have heterogeneous elastic and plastic properties of the same order as those found in rocks. The pattern of stress percolation is related to the degree of heterogeneity in and statistical distribution of the elastic and plastic properties of the constituent grains in the aggregate. The development of these stress patterns leads directly to shear localization, and their existence provides insight into the formation of rhythmic features such as compositional banding and foliation in rocks that are reacting or dissolving while being deformed. In addition, this framework provides a foundation for understanding and predicting the macroscopic rheology of polycrystalline materials based on single-crystal elastic and plastic mechanical properties.

  17. Selection and Basic Properties of the Buffer Material for High-Level Radioactive Waste Repository in China

    Institute of Scientific and Technical Information of China (English)

    WEN Zhijian

    2008-01-01

    Radioactive wastes arising from a wide range of human activities are in many different physical and chemical forms, contaminated with varying radioactivity. Their common features are the potential hazard associated with their radioactivity and the need to manage them in such a way as to protect the human environment. The geological disposal is regarded as the most reasonable and effective way to safely disposing high-level radioactive wastes in the world. The conceptual model of geological disposal in China is based on a multi-barrier system that combines an isolating geological environment with an engineered barrier system. The buffer is one of the main engineered barriers for HLW repository. It is expected to maintain its low water permeability, self-sealing property, radio nuclides adsorption and retardation properties, thermal conductivity, chemical buffering property,canister supporting property, and stress buffering property over a long period of time. Bentonite is selected as the main content of buffer material that can satisfy the above requirements. The Gaomiaozi deposit is selected as the candidate supplier for China's buffer material of high level radioactive waste repository. This paper presents the geological features of the GMZ deposit and basic properties of the GMZ Na-bentonite. It is a super-large deposit with a high content of montmorillonite (about 75%), and GMZ-1, which is Na-bentonite produced from GMZ deposit is selected as the reference material for China's buffer material study.

  18. Reexamination of the source material of acid igneous rocks, based on the selected Sr isotopic data

    International Nuclear Information System (INIS)

    Kagami, Hiroo; Shuto, Kenji; Gorai, Masao

    1975-01-01

    The relation between the ages and the initial strontium isotopic compositions obtained from acid igneous rocks by the whole-rock isochron method is re-examined, on the basis of the selected data. The points based on the data having high values of standard deviation (on the isochrons) show considerable scattering. This is probably ascribed to admixture of sialic materials, or secondary alteration and other geologic causes. The points based on the data having lower values of standard deviation (sigma value: 0.0001 - 0.0019), on the other hand, are evidently plotted within a narrow region just above the presumed Sr evolutional region of the source material of oceanic tholeiites. It is noteworthy that the former region meets the latter region at an earlier stage of the evolutional history of the earth (about 40 x 10 8 yrs. ago or older). It may be conceivable that the former region is the Sr evolutional region of the source material of acid igneous rocks. Considering from the inclination of the above Sr evolutional region, the source material of most of acid igneous rocks may possibly be a certain basic material, chemically similar to the continental tholeiitic basalts or basaltic andesites. On the other hand, the source material of a few acid igneous rocks with low initial strontium isotopic ratios may be a certain basic material resembling the oceanic tholeiites. Another possibility is that these acid igneous rocks and oceanic tholeiites may have been formed, under different physical conditions, directly from a certain common source material presumably of peridotitic composition. (auth.)

  19. Thermo-mechanical effects from a KBS-3 type repository. Performance of pillars between repository tunnels

    International Nuclear Information System (INIS)

    Hakami, E.; Olofsson, Stig-Olof

    2000-03-01

    The aim of this study has been to investigate how the rock mass, in the near field of a KBS-3 type repository, will be affected by the excavation of tunnels and deposition holes and the thermal load from the deposited waste. The three-dimensional finite difference program FLAC 3D was used to perform numerical simulation of the rock mass behaviour. The rock mass was modelled as a homogeneous and isotropic continuum. The initial area heat intensity of the repository was assumed to be 10 W/m 2 in all models. The results show that in the middle of the pillar between the repository tunnels the temperature reaches a maximum of about 70 deg C after 55 years of deposition. The extent of areas where the rock is predicted to yield depends on the assumed quality of the rock mass and the initial in-situ stress field. The volume of yielded rock reaches a maximum after about 200 years after deposition. For a rock mass with internal friction angle of 45 deg and cohesion of 5 MPa (using a Mohr-Coulomb material model), the extent of yielded rock is limited to about 1.5 m behind the excavation periphery. The largest rock displacements are found in the tunnel floor at the upper part of the deposition holes. Tension and shear failure in the periphery of the excavations is predicted to occur during the rock excavation, with a depth extension depending on the magnitude and orientation of the in-situ stresses, as well as on the rock mass quality. Both the excavation effects and the then-no-mechanical effects are smallest when the major principal stress is oriented parallel with the deposition tunnels. The maximum convergence between tunnel walls was calculated to occur after 200 years and be about 9 mm, in the model assuming a rock mass with 5 MPa cohesion, 45 deg internal friction angle and maximum horizontal stress perpendicular to the tunnel. In this study confining effects from the buffer and backfill material was neglected. The effective stress concept was used in most of the models

  20. Constructability analysis for a deep repository - some thoughts on possibilities and limitations

    International Nuclear Information System (INIS)

    Baeckblom, G.; Leijon, B.; Stille, H.

    1995-01-01

    Potential site characterization for construction of a repository for geologic disposal of spent fuel in crystalline rock in Sweden is discussed. The present plan requires that the fuel be encapsulated in a composite steel-copper canister, that the repository be situated somewhere in Sweden with municipal approval, and that licensing be preceded by extensive studies and investigations. Important factors are mechanical stability, hydrology, and the suitability of construction materials. Site investigation will require a lot of surface and borehole information regarding rock types, zones, structures, fractures, hydraulic conductivity, stresses, rock strength, and groundwater chemistry. 6 refs., 4 tabs., 1 fig

  1. Preliminary design of the repository, stage 2

    International Nuclear Information System (INIS)

    Saanio, T.; Kirkkomaeki, T.; Keto, P.; Kukkola, T.; Raiko, H.

    2007-01-01

    , investigations in ONKALO and investigations during the excavation and operation of the repository. The repository is planned so that technical development can be flexibly utilized. The total volume of the repository is approximately 1.3 million m3. The maximum open volume at any one time is around 0.6 million m3, because the repository is excavated and backfilled in stages. The repository is divided into the controlled area and the uncontrolled area. Canisters are always handled and lowered to the deposition hole in the controlled area. The excavation and construction of new tunnels and the backfilling of the tunnels is carried out in the uncontrolled area. Extensive material transfers, such as transfers of broken rock and backfilling materials are conducted in the access tunnel. Separate ventilation systems are provided for the controlled and the uncontrolled area. (orig.)

  2. Preliminary design of the repository. Stage 2

    International Nuclear Information System (INIS)

    Saanio, T.; Kirkkomaeki, T.; Keto, P.; Kukkola, T.; Raiko, H.

    2007-04-01

    , investigations in ONKALO and investigations during the excavation and operation of the repository. The repository is planned so that technical development can be flexibly utilized. The total volume of the repository is approximately 1.3 million m 3 . The maximum open volume at any one time is around 0.6 million m 3 , because the repository is excavated and backfilled in stages. The repository is divided into the controlled area and the uncontrolled area. Canisters are always handled and lowered to the deposition hole in the controlled area. The excavation and construction of new tunnels and the backfilling of the tunnels is carried out in the uncontrolled area. Extensive material transfers, such as transfers of broken rock and backfilling materials are conducted in the access tunnel. Separate ventilation systems are provided for the controlled and the uncontrolled area. (orig.)

  3. Protein Structure Initiative Material Repository: an open shared public resource of structural genomics plasmids for the biological community

    Science.gov (United States)

    Cormier, Catherine Y.; Mohr, Stephanie E.; Zuo, Dongmei; Hu, Yanhui; Rolfs, Andreas; Kramer, Jason; Taycher, Elena; Kelley, Fontina; Fiacco, Michael; Turnbull, Greggory; LaBaer, Joshua

    2010-01-01

    The Protein Structure Initiative Material Repository (PSI-MR; http://psimr.asu.edu) provides centralized storage and distribution for the protein expression plasmids created by PSI researchers. These plasmids are a resource that allows the research community to dissect the biological function of proteins whose structures have been identified by the PSI. The plasmid annotation, which includes the full length sequence, vector information and associated publications, is stored in a freely available, searchable database called DNASU (http://dnasu.asu.edu). Each PSI plasmid is also linked to a variety of additional resources, which facilitates cross-referencing of a particular plasmid to protein annotations and experimental data. Plasmid samples can be requested directly through the website. We have also developed a novel strategy to avoid the most common concern encountered when distributing plasmids namely, the complexity of material transfer agreement (MTA) processing and the resulting delays this causes. The Expedited Process MTA, in which we created a network of institutions that agree to the terms of transfer in advance of a material request, eliminates these delays. Our hope is that by creating a repository of expression-ready plasmids and expediting the process for receiving these plasmids, we will help accelerate the accessibility and pace of scientific discovery. PMID:19906724

  4. Modeling transient heat transfer in nuclear waste repositories.

    Science.gov (United States)

    Yang, Shaw-Yang; Yeh, Hund-Der

    2009-09-30

    The heat of high-level nuclear waste may be generated and released from a canister at final disposal sites. The waste heat may affect the engineering properties of waste canisters, buffers, and backfill material in the emplacement tunnel and the host rock. This study addresses the problem of the heat generated from the waste canister and analyzes the heat distribution between the buffer and the host rock, which is considered as a radial two-layer heat flux problem. A conceptual model is first constructed for the heat conduction in a nuclear waste repository and then mathematical equations are formulated for modeling heat flow distribution at repository sites. The Laplace transforms are employed to develop a solution for the temperature distributions in the buffer and the host rock in the Laplace domain, which is numerically inverted to the time-domain solution using the modified Crump method. The transient temperature distributions for both the single- and multi-borehole cases are simulated in the hypothetical geological repositories of nuclear waste. The results show that the temperature distributions in the thermal field are significantly affected by the decay heat of the waste canister, the thermal properties of the buffer and the host rock, the disposal spacing, and the thickness of the host rock at a nuclear waste repository.

  5. On the source material of magmas - with special reference to Nd isotopic ratios of igneous rocks

    International Nuclear Information System (INIS)

    Shuto, Kenji

    1980-01-01

    In 1973, the Sm-Nd method was first used for the measurement of the absolute age of igneous rocks and meteorites. Subsequently in the following years, the research works by means of the Nd isotopic ratio in igneous rocks have been made strenuously in order to reveal the chemistry of the source materials of magma giving rise to the igneous rocks and further the evolution process of mantle and earth's crust. The fundamental items for the Sm-Nd method are explained. Then, the research results more important in the above connection are given. Finally, the ideas by the author concerning the source materials of magma are presented from the data available on the Nd isotopes in meteorites and igneous rocks. The following matters are described: the fundamentals of Sm-Nd method, the Nd content in seawater, the negative correlation between Nd and Sr isotopic ratios in igneous rocks, magma source materials and Nd isotopes, and considerations on magma source materials. (J.P.N.)

  6. Forecasts and restrictions on vibrations from rock excavation and transportation. Encapsulation Plant and Repository for spent nuclear fuel, Laxemar; Prognoser och restriktioner foer vibrationer fraan bergschaktning och transporter. Inkapslingsanlaeggning och slutfoervar foer anvaent kaernbraensle, Laxemar

    Energy Technology Data Exchange (ETDEWEB)

    Lind, Carl; Johansson, Sven-Erik (Nitro Consult AB (Sweden))

    2010-12-15

    This study describes the impact on the surroundings that may occur during rock excavation activities for the final repository for spent nuclear fuel in Laxemar and the encapsulation facility in Simpevarp. The study also includes vibrations created by heavy shipments related to activities at the final repository. The study will provide input to the environmental impact assessment and future design work. The survey area for buildings and facilities covered by the study extends approximately 1,000 metres from the proposed location of the final repository. For the encapsulation facility the survey area has been limited to residential buildings and summer houses within 1,000 metres of the proposed location. In addition, residential buildings along road 743 have been surveyed with regard to the impact of heavy shipments between Laxemar and Faarbo. The results of the surveys and information on planned rock excavation activities have been used to formulate preliminary restrictions and predictions of vibrations and air shock waves from blasting, as well as noise from rock drilling. Predictions have also been made of vibrations from heavy shipments, and a reference survey has been carried out in a residential building near road 743. The predictions of vibrations from blasting rounds reveal low or very low levels. No risk of damage to buildings or equipment is expected. Vibrations from blasting may, however, be perceptible within large parts of the study area, since the human perception threshold for vibration is very low. They will hardly be regarded as disturbing, however. When the accesses to the final repository have been built and rock excavation continues at repository level, the impact on the surroundings is expected to be minimal. The main reason for this is that the blasting will then occur at a depth of about 500 metres, at an ample distance to buildings at surface level. Predictions of air shock waves from blasting rounds indicate low levels. There is no risk of

  7. Characterisation and modelling of mixing processes in groundwaters of a potential geological repository for nuclear wastes in crystalline rocks of Sweden

    International Nuclear Information System (INIS)

    Gómez, Javier B.; Gimeno, María J.; Auqué, Luis F.; Acero, Patricia

    2014-01-01

    This paper presents the mixing modelling results for the hydrogeochemical characterisation of groundwaters in the Laxemar area (Sweden). This area is one of the two sites that have been investigated, under the financial patronage of the Swedish Nuclear Waste and Management Co. (SKB), as possible candidates for hosting the proposed repository for the long-term storage of spent nuclear fuel. The classical geochemical modelling, interpreted in the light of the palaeohydrogeological history of the system, has shown that the driving process in the geochemical evolution of this groundwater system is the mixing between four end-member waters: a deep and old saline water, a glacial meltwater, an old marine water, and a meteoric water. In this paper we put the focus on mixing and its effects on the final chemical composition of the groundwaters using a comprehensive methodology that combines principal component analysis with mass balance calculations. This methodology allows us to test several combinations of end member waters and several combinations of compositional variables in order to find optimal solutions in terms of mixing proportions. We have applied this methodology to a dataset of 287 groundwater samples from the Laxemar area collected and analysed by SKB. The best model found uses four conservative elements (Cl, Br, oxygen-18 and deuterium), and computes mixing proportions with respect to three end member waters (saline, glacial and meteoric). Once the first order effect of mixing has been taken into account, water–rock interaction can be used to explain the remaining variability. In this way, the chemistry of each water sample can be obtained by using the mixing proportions for the conservative elements, only affected by mixing, or combining the mixing proportions and the chemical reactions for the non-conservative elements in the system, establishing the basis for predictive calculations. - Highlights: • Laxemar (Sweden) groundwater is the combined result

  8. Characterisation and modelling of mixing processes in groundwaters of a potential geological repository for nuclear wastes in crystalline rocks of Sweden

    Energy Technology Data Exchange (ETDEWEB)

    Gómez, Javier B., E-mail: jgomez@unizar.es; Gimeno, María J., E-mail: mjgimeno@unizar.es; Auqué, Luis F., E-mail: lauque@unizar.es; Acero, Patricia, E-mail: patriace@unizar.es

    2014-01-01

    This paper presents the mixing modelling results for the hydrogeochemical characterisation of groundwaters in the Laxemar area (Sweden). This area is one of the two sites that have been investigated, under the financial patronage of the Swedish Nuclear Waste and Management Co. (SKB), as possible candidates for hosting the proposed repository for the long-term storage of spent nuclear fuel. The classical geochemical modelling, interpreted in the light of the palaeohydrogeological history of the system, has shown that the driving process in the geochemical evolution of this groundwater system is the mixing between four end-member waters: a deep and old saline water, a glacial meltwater, an old marine water, and a meteoric water. In this paper we put the focus on mixing and its effects on the final chemical composition of the groundwaters using a comprehensive methodology that combines principal component analysis with mass balance calculations. This methodology allows us to test several combinations of end member waters and several combinations of compositional variables in order to find optimal solutions in terms of mixing proportions. We have applied this methodology to a dataset of 287 groundwater samples from the Laxemar area collected and analysed by SKB. The best model found uses four conservative elements (Cl, Br, oxygen-18 and deuterium), and computes mixing proportions with respect to three end member waters (saline, glacial and meteoric). Once the first order effect of mixing has been taken into account, water–rock interaction can be used to explain the remaining variability. In this way, the chemistry of each water sample can be obtained by using the mixing proportions for the conservative elements, only affected by mixing, or combining the mixing proportions and the chemical reactions for the non-conservative elements in the system, establishing the basis for predictive calculations. - Highlights: • Laxemar (Sweden) groundwater is the combined result

  9. Gas and water flow in an excavation-induced fracture network around an underground drift: A case study for a radioactive waste repository in clay rock

    Science.gov (United States)

    de La Vaissière, Rémi; Armand, Gilles; Talandier, Jean

    2015-02-01

    The Excavation Damaged Zone (EDZ) surrounding a drift, and in particular its evolution, is being studied for the performance assessment of a radioactive waste underground repository. A specific experiment (called CDZ) was designed and implemented in the Meuse/Haute-Marne Underground Research Laboratory (URL) in France to investigate the EDZ. This experiment is dedicated to study the evolution of the EDZ hydrogeological properties (conductivity and specific storage) of the Callovo-Oxfordian claystone under mechanical compression and artificial hydration. Firstly, a loading cycle applied on a drift wall was performed to simulate the compression effect from bentonite swelling in a repository drift (bentonite is a clay material to be used to seal drifts and shafts for repository closure purpose). Gas tests (permeability tests with nitrogen and tracer tests with helium) were conducted during the first phase of the experiment. The results showed that the fracture network within the EDZ was initially interconnected and opened for gas flow (particularly along the drift) and then progressively closed with the increasing mechanical stress applied on the drift wall. Moreover, the evolution of the EDZ after unloading indicated a self-sealing process. Secondly, the remaining fracture network was resaturated to demonstrate the ability to self-seal of the COx claystone without mechanical loading by conducting from 11 to 15 repetitive hydraulic tests with monitoring of the hydraulic parameters. During this hydration process, the EDZ effective transmissivity dropped due to the swelling of the clay materials near the fracture network. The hydraulic conductivity evolution was relatively fast during the first few days. Low conductivities ranging at 10-10 m/s were observed after four months. Conversely, the specific storage showed an erratic evolution during the first phase of hydration (up to 60 days). Some uncertainty remains on this parameter due to volumetric strain during the

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  11. Characterisation and modelling of mixing processes in groundwaters of a potential geological repository for nuclear wastes in crystalline rocks of Sweden.

    Science.gov (United States)

    Gómez, Javier B; Gimeno, María J; Auqué, Luis F; Acero, Patricia

    2014-01-15

    This paper presents the mixing modelling results for the hydrogeochemical characterisation of groundwaters in the Laxemar area (Sweden). This area is one of the two sites that have been investigated, under the financial patronage of the Swedish Nuclear Waste and Management Co. (SKB), as possible candidates for hosting the proposed repository for the long-term storage of spent nuclear fuel. The classical geochemical modelling, interpreted in the light of the palaeohydrogeological history of the system, has shown that the driving process in the geochemical evolution of this groundwater system is the mixing between four end-member waters: a deep and old saline water, a glacial meltwater, an old marine water, and a meteoric water. In this paper we put the focus on mixing and its effects on the final chemical composition of the groundwaters using a comprehensive methodology that combines principal component analysis with mass balance calculations. This methodology allows us to test several combinations of end member waters and several combinations of compositional variables in order to find optimal solutions in terms of mixing proportions. We have applied this methodology to a dataset of 287 groundwater samples from the Laxemar area collected and analysed by SKB. The best model found uses four conservative elements (Cl, Br, oxygen-18 and deuterium), and computes mixing proportions with respect to three end member waters (saline, glacial and meteoric). Once the first order effect of mixing has been taken into account, water-rock interaction can be used to explain the remaining variability. In this way, the chemistry of each water sample can be obtained by using the mixing proportions for the conservative elements, only affected by mixing, or combining the mixing proportions and the chemical reactions for the non-conservative elements in the system, establishing the basis for predictive calculations. © 2013 Elsevier B.V. All rights reserved.

  12. Foreign materials in a deep repository for spent nuclear fuels; Fraemmande material i ett djupfoervar foer anvaent kaernbraensle

    Energy Technology Data Exchange (ETDEWEB)

    Jones, C.; Christiansson, Aa.; Wiborgh, M. [Kemakta Konsult AB, Stockholm (Sweden)

    1999-12-01

    The effects of foreign substances introduced into a spent-fuel repository are reviewed. Possible impacts on processes and barrier-functions are examined, and the following areas are identified: Corrosion of the spent-fuel canister through the presence of sulfur and substances that favor microbial growth; impacts on the bentonite properties through the presence of cations as calcium, potassium and iron; radionuclide transport through the presence of complex-formers and surface-active substances.

  13. Relationship of material properties to seismic coupling. Part I. Shock wave studies of rock and rock-like materials

    International Nuclear Information System (INIS)

    Larson, D.B.; Rodean, H.C.

    1975-01-01

    Our research seeks an understanding of the relationship of material properties to explosive-energy coupling in various earth media by integrating experimental observations with computer calculational models to obtain a predictive capability. The procedure chosen consists of: first, selecting materials exhibiting interesting values of the properties that are believed to control coupling; second, experimentally determining material behavior under various types of loading and unloading; third, development of constitutive relationships; fourth, adapting these constitutive relationships to computer calculational models; and fifth, verifying the calculational models through comparison with small-scale and field high-strain-rate experiments. The object of this report is to present the shock-wave data and to make a preliminary evaluation of the results in terms of material properties, coupling, and their interactions. (U.S.)

  14. Granite-repository - geochemical environment

    International Nuclear Information System (INIS)

    1979-04-01

    Some geochemical data of importance for a radioactive waste repository in hard rock are reviewed. The ground water composition at depth is assessed. The ground water chemistry in the vicinity of uranium ores is discussed. The redox system in Swedish bedrock is described. Influences of extreme climatic changes and of repository mining and construction are also evaluated

  15. Redox front penetration in the fractured Toki Granite, central Japan: An analogue for redox reactions and redox buffering in fractured crystalline host rocks for repositories of long-lived radioactive waste

    International Nuclear Information System (INIS)

    Yamamoto, Koshi; Yoshida, Hidekazu; Akagawa, Fuminori; Nishimoto, Shoji; Metcalfe, Richard

    2013-01-01

    Highlights: • Deep redox front developed in orogenic granitic rock have been studied. • The process was controlled by the buffering capacity of minerals. • This is an analogue of redox front penetration into HLW repositories in Japan. - Abstract: Redox buffering is one important factor to be considered when assessing the barrier function of potential host rocks for a deep geological repository for long-lived radioactive waste. If such a repository is to be sited in fractured crystalline host rock it must be demonstrated that waste will be emplaced deeper than the maximum depth to which oxidizing waters can penetrate from the earth’s surface via fractures, during the assessment timeframe (typically 1 Ma). An analogue for penetration of such oxidizing water occurs in the Cretaceous Toki Granite of central Japan. Here, a deep redox front is developed along water-conducting fractures at a depth of 210 m below the ground surface. Detailed petrographical studies and geochemical analyses were carried out on drill core specimens of this redox front. The aim was to determine the buffering processes and behavior of major and minor elements, including rare earth elements (REEs), during redox front development. The results are compared with analytical data from an oxidized zone found along shallow fractures (up to 20 m from the surface) in the same granitic rock, in order to understand differences in elemental migration according to the depth below the ground surface of redox-front formation. Geochemical analyses by XRF and ICP-MS of the oxidized zone at 210 m depth reveal clear changes in Fe(III)/Fe(II) ratios and Ca depletion across the front, while Fe concentrations vary little. In contrast, the redox front identified along shallow fractures shows strong enrichments of Fe, Mn and trace elements in the oxidized zone compared with the fresh rock matrix. The difference can be ascribed to the changing Eh and pH of groundwater as it flows downwards in the granite, due to

  16. Development of mechanical-hydraulic models for the prediction of the long-term sealing capacity of concrete based sealing materials in rock salt. Project Titel LASA

    Energy Technology Data Exchange (ETDEWEB)

    Czaikowski, Oliver; Dittrich, Juergen; Hertes, Uwe; Jantschik, Kyra; Wieczorek, Klaus; Zehle, Bernd

    2016-08-15

    The research work leading to these results has received funding from the German Federal Ministry of Economic Affairs and Energy (BMWi) under contract no. 02E11132. This report presents the current state of laboratory investigations and modelling activities related to the LASA project. The work is related to the research and development of plugging and sealing for repositories in salt rock and is of fundamental importance for the salt option which represents one of the three European repository options in addition to the clay rock and the crystalline rock options.

  17. Repository for high level radioactive wastes in Brazil: the importance of geochemical (Micro thermometric) studies and fluid migration in potential host rocks; Repositorios para rejeitos radioativos de alto nivel (RANR) no Brasil: a importancia de estudos geoquimicos (microtermometricos) e de migracao de fluidos em rochas potenciamente hospedeiras

    Energy Technology Data Exchange (ETDEWEB)

    Rios, Francisco Javier; Fuzikawa, Kazuo; Alves, James Vieira; Neves, Jose Marques Correia [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN-CNEN-MG), Belo Horizonte, MG (Brazil). Lab. de Inclusoes Fluidas e Metalogenese]. E-mail: javier@cdtn.br

    2003-04-15

    A detailed fluid inclusion study of host rocks, is of fundamental importance in the selection of geologically suitable areas for high level nuclear waste repository constructions (HLRW). The LIFM-CDTN is enabled to develop studies that confirm: the presence or not, of corrosive fluid in minerals from host rocks of the repository and the possible presence of micro fractures (and fluid leakage) when these rocks are submitted to high temperatures. These fluid geochemistry studies, with permeability determinations by means of pressurized air injection must be carried out in rocks hosting nuclear waste. Micro fracture determination is of vital importance since many naturally corrosive solutions, present in the mineral rocks, could flow out through these plans affecting the walls of the repository. (author)

  18. Handling and final disposal of nuclear waste. Hard Rock Laboratory

    International Nuclear Information System (INIS)

    1989-09-01

    The purpose of the Hard Rock Laboratory is to provide an opportunity for research and development in a realistic and undisturbed underground rock environment down to the depth planned for the future repository. The R and D work in the underground laboratory has the following main goals: To test the quality and appropriateness of different methods for characterizing the bedrock with respect to conditions of importance for a final repository. To refine and demonstrate methods for how to adapt a repository to the local properties of the rock in connection with planning and construction. And, finally, to collect material and data of importance for the safety of the future repository and for confidence in the quality of the safety assessments 13 figs, 3 tabs

  19. Implications of the use of low-pH cementitious materials in high activity radioactive waste repositories

    International Nuclear Information System (INIS)

    Garcia Calvo, J.L.; Alonso, M.C.; Fernandez Luco, L.; Hidalgo, A.; Sanchez, M.

    2008-01-01

    One of the most accepted engineering construction concepts for high radioactive nuclear waste of underground repositories considers the use of low pH cementitious materials, in order to avoid the formation of an alkaline plume fluid which perturbs one of the engineered barriers of the repository, the bentonite. The accepted solution to maintain the bentonite stability, which is function of the pH, is to develop cementitious materials that generate pore waters with pH ≤ 11, because the corrosion velocity of the clay is significantly reduced below this value. The IETcc-CSIC has focused the research activity on low-pH cementitious materials using two cements: Ordinary Portland Cements (OPC) and Calcium Aluminates Cements (CAC). In both cases, the achievement of a low-pH environment implies the use of high content of mineral admixtures to prepare the binder. Obviously, the inclusion of high contents of mineral admixtures in the cement formulation modifies most of the concrete 'standard' properties and the microstructure of the obtained cement products. When designing a concrete based on low-pH binders, not only the functional requirements have to be reached but also the modifications of the basic properties of the concrete must be taken into account. Besides, due to the location and the long service life of this type of products, their durability properties must be also guaranteed. This paper deals with the procedure followed in the design of a specific application of low pH cements; for instance, the shotcrete plug fabrication. The challenge of this type of use (shotcreting) is more complex taking into account that requires the employment of additives that must be compatible with the concrete mixture. Furthermore, their effectiveness must be assured without increase the pH above the admissible levels. Therefore, their compatibility with admixtures is tested in the present work. The compliance of the requirements for a shotcrete plug was evaluated at laboratory scale

  20. SORPTION AND DISPERSION OF STRONTIUM RADIONUCLIDE IN THE BENTONITE-QUARTZ-CLAY AS BACKFILL MATERIAL CANDIDATE ON RADIOACTIVE WASTE REPOSITORY

    Directory of Open Access Journals (Sweden)

    Herry Poernomo

    2010-12-01

    Full Text Available The experiment of sorption and dispersion characteristics of strontium in the mixture of bentonite-quartz, clay-quartz, bentonite-clay-quartz as candidate of raw material for backfill material in the radioactive waste repository has been performed. The objective of this research is to know the grain size effect of bentonite, clay, and quartz on the weight percent ratio of bentonite to quartz, clay to quartz, bentonite to clay to-quartz can be gives physical characteristics of best such as bulk density (rb, effective porosity (e, permeability (K, best sorption characteristic such as distribution coefficient (Kd, and best dispersion characteristics such as dispersivity (a and effective dispersion coefficient (De of strontium in the backfill material candidate. The experiment was carried out in the column filled by the mixture of bentonite-quartz, clay-quartz, bentonite-clay-quartz with the weight percent ratio of bentonite to quartz, clay to quartz, bentonite to clay to quartz of 100/0, 80/20, 60/40, 40/60, 20/80, 0/100 respectively at saturated condition of water, then flowed 0.1 N Sr(NO32 as buffer solution with tracer of 0.05 Ci/cm3 90Sr as strontium radionuclide simulation was leached from immobilized radioactive waste in the radioactive waste repository. The concentration of 90Sr in the effluents represented as Ct were analyzed by Ortec b counter every 30 min, then by using profile concentration of Co and Ct, values of Kd, a and De of 90Sr in the backfill material was determined. The experiment data showed that the best results were -80+120 mesh grain size of bentonite, clay, quartz respectively on the weight percent ratio of bentonite to clay to quartz of 70/10/20 with physical characteristics of rb = 0.658 g/cm3, e = 0.666 cm3/cm3, and K = 1.680x10-2 cm/sec, sorption characteristic of Kd = 46.108 cm3/g, dispersion characteristics of a = 5.443 cm, and De = 1.808x10-03 cm2/sec can be proposed as candidate of raw material of backfill material

  1. Mechanical stability of repository tunnels and factors to be considered for determining tunnel spacing

    International Nuclear Information System (INIS)

    Takeuchi, Kunifumi

    1994-01-01

    Kristallin-1 organized by Nagra is currently advanced as a synthetic project regarding a high level radioactive waste (HLW) repository in Switzerland. Its host rock is granitic rocks, and the potential siting area is located in northern Switzerland. The objective of this project is to demonstrate the long term safety of a HLW repository under more site-specific conditions than before. As the detailed geological data were investigated, the average size of undisturbed crystalline rock blocks is limited horizontally to about several hundred meter, therefore, the HLW repository area must be divided into several panels to avoid fracture zones. It is necessary to make tunnel spacing as small as possible for the purpose of reasonably designing the entire layout of repository tunnels. The main factors to be considered for determining repository tunnel spacing are listed. Rock mass modeling, rock mass material properties, the analysis model and parameters, the numerical analysis of repository tunnel stability and its main conclusion are reported. The numerical analysis of the temperature distribution in near field was carried out. Tunnel spacing should be set more than 20 m in view of the maximum temperature. (K.I.)

  2. Rock as a construction material: durability, deterioration and conservation

    Directory of Open Access Journals (Sweden)

    Esbert, Rosa M.ª

    1991-03-01

    Full Text Available The different aspects related to the deterioration and conservation of stone, used as a construction material, are reviewed in this article. The petrographical characteristics and physical properties which control the durability of stone material are stated. The importance of the voids and the properties more directly linked to the up-taking and transfer of humidity through the stone are pointed out. Regarding to the deterioration processes, the role of water, soluble salts and atmospheric pollutants upon the different alteration mechanisms of the building stones is emphasized. Finally, the steps related to the stone conservation, and the methods and products more currently employed to that aim are revised.

    Se compendian los distintos aspectos relacionados con el deterioro y la conservación de la piedra utilizada como material de edificación. Se revisan las características petrográficas y propiedades físicas que controlan la durabilidad de los materiales pétreos, resaltando la importancia de los espacios vacíos y de aquellas propiedades más directamente relacionadas con la captación y transferencia de humedad por el interior de la piedra. En cuanto a los procesos de deterioro se destaca el papel del agua, de las sales solubles y de los contaminantes atmosféricos en los diversos mecanismos de alteración desarrollados en la piedra de edificación. Finalmente se plantean las diversas fases relacionadas con la conservación de la piedra, y se revisan los métodos y productos más empleados en la actualidad para tal fin.

  3. Chemical conditions in the repository for low- and intermediate-level reactor waste

    International Nuclear Information System (INIS)

    Snellman, M.; Uotila, H.

    1984-01-01

    The chemical conditions in the proposed repositories for low- and intermediate-level reactor waste at Haestholmen (IVO) and Olkiluoto (TVO) have been discussed with respect to materials introduced into the repository, their possible long-term changes and interaction with groundwater flowing into the repository. The main possible groundwater-rock interactions have been discussed, as well as the role of micro-organisms, organic acids and colloids in the estimation of the barrier integrity. Experimental and theoretical studies have been performed on the basis of the natural groundwater compositions expected at the repository sites. Main emphasis is put on the chemical parameters which might influence the integrity of the different barriers in the repository as well as on the parameters which might effect the release and transport of radionuclides from the repository

  4. Report on materials characterization center workshop on stress corrosion cracking for the Salt Repository Project, December 16-17, 1986, Seattle, Washington: Workshop summary

    International Nuclear Information System (INIS)

    Merz, M.D.; Shannon, D.W.

    1986-09-01

    The Materials Characterization Center (MCC) at Pacific Northwest Laboratory (PNL) conducted a Workshop on Stress Corrosion Cracking for the Salt Repository Project on December 16 and 17, 1986 in Seattle, Washington. The workshop was held to formulate recommendations for addressing stress corrosion cracking (SCC) in a salt repository. It was attended by 24 representatives from major laboratories, universities, and industry. This report presents the recommendations of the workshop, along with the agenda, list of participants, questions and comments, summaries of working groups on low-strength steel and alternate materials, and materials handed out by the speakers

  5. Emplacement of rock avalanche material across saturated sediments, Southern Alp, New Zealand

    Science.gov (United States)

    Dufresne, A.; Davies, T. R.; McSaveney, M. J.

    2012-04-01

    The spreading of material from slope failure events is not only influenced by the volume and nature of the source material and the local topography, but also by the materials encountered in the runout path. In this study, evidence of complex interactions between rock avalanche and sedimentary runout path material were investigated at the 45 x 106 m3 long-runout (L: 4.8 km) Round Top rock avalanche deposit, New Zealand. It was sourced within myolinitic schists of the active strike-slip Alpine Fault. The narrow and in-failure-direction elongate source scarp is deep-seated, indicating slope failure was triggered by strong seismic activity. The most striking morphological deposit features are longitudinal ridges aligned radially to source. Trenching and geophysical surveys show bulldozed and sheared substrate material at ridge termini and laterally displaced sedimentary strata. The substrate failed at a minimum depth of 3 m indicating a ploughing motion of the ridges into the saturated material below. Internal avalanche compression features suggest deceleration behind the bulldozed substrate obstacle. Contorted fabric in material ahead of the ridge document substrate disruption by the overriding avalanche material deposited as the next down-motion hummock. Comparison with rock avalanches of similar volume but different emplacement environments places Round Top between longer runout avalanches emplaced over e.g. playa lake sediments and those with shorter travel distances, whose runout was apparently retarded by topographic obstacles or that entrained high-friction debris. These empirical observations indicate the importance of runout path materials on tentative trends in rock avalanche emplacement dynamics and runout behaviour.

  6. Thermal conductivity of compacted bentonite as a buffer material for a high-level radioactive waste repository

    International Nuclear Information System (INIS)

    Lee, Jae Owan; Choi, Heuijoo; Lee, Jong Youl

    2016-01-01

    Highlights: • The thermal conductivities were measured under various disposal conditions. • They were significantly influenced by the water content and dry density. • They were not sensitive to the temperature and the anisotropic structure. • A new model of thermal conductivity was proposed for the thermal analysis. - Abstract: Bentonite buffer is one of the major barrier components of a high-level radioactive waste (HLW) repository, and the thermal conductivity of the bentonite buffer is a key parameter for the thermal performance assessment of the HLW repository. This study measured the thermal conductivity of compacted bentonite as a buffer material and investigated its dependence upon various disposal conditions: the dry density, water content, anisotropic structure of the compacted bentonite, and temperature. The measurement results showed that the thermal conductivity was significantly influenced by the water content and dry density of the compacted bentonite, while there was not a significant variation with respect to the temperature. The anisotropy of the thermal conductivity had a negligible variation for an increasing dry density. The present study also proposed a geometric mean model of thermal conductivity which best fits the experimental data.

  7. Mont Terri Project - Heater experiment : rock and bentonite thermo-hydro-mechanical (THM) processes in the near field of a thermal source for development of deep underground high level radioactive waste repositories

    Energy Technology Data Exchange (ETDEWEB)

    Goebel, I.; Alheid, H.-J.; Kaufhold, St.; Naumann, M.; Pletsch, Th.; Plischke, I.; Schnier, H.; Schuster, K.; Sprado, K. [Bundesanstalt fuer Geowissenschaften und Rohstoffe (BGR), Hannover (Germany); Meyer, T.; Miehe, R.; Wieczorek, K. [Gesellschaft fuer Anlagen und Reaktorsicherheit mbH (GRS), Braunschweig (Germany); Mayor, J.C. [Empresa Nacional de Residuos Radioactivos SA (ENRESA), Madrid (Spain); Garcia-Sineriz, J.; Rey, M. [Asociacion para la Investigacion y Desarollo Industrial de los Recursos Naturales (AITEMIN), Madrid (Spain); Alonso, E.; Lloret, A.; Munoz, J.J. [Centre Internacional de Metodos Numerics en Ingenyeria (CIMNE), Barcelona (Spain); Weber, H. [National Cooperative for the Disposal of Radioactive Waste (Nagra), Wettingen (Switzerland); Ploetze, M. [Eidgenoessische Technische Hochschule Zuerich, Institut fuer Geotechnik, Zuerich (Switzerland); Klubertanz, G. [Colenco Power Engineering Ltd, Baden (Switzerland); Ammon, Ch. [Rothpletz Lienhard und Cie AG, Aarau (Switzerland); Graf, A.; Nussbaum, Ch.; Zingg, A. [Goetechnical Institute Ltd, Saint-Ursanne (Switzerland); Bossart, P. [Federal Office of Topography (swisstopo), Wabern (Switzerland); Buehler, Ch.; Kech, M.; Trick, Th. [Solexperts AG, Moenchaltorf (Switzerland); Emmerich, K. [ITC-WGT, Karlsruhe (Germany); Fernandez, A. M. [Ciemat, Madrid (Spain)

    2007-07-01

    The long-term safety of underground permanent repositories for radioactive waste relies on a combination of several engineered and geological barriers. The interactions between a host rock formation of the type 'Opalinus Clay' and an engineered barrier of the type 'bentonite buffer' are observed in the Heater Experiment (HE) during a hydration and a heating phase. The objective of the experiment is an improved understanding of the coupled thermo-hydro-mechanical (THM) processes in a host rock-buffer system achieved by experimental observations as well as numerical modelling. The basic objectives are in detail: a) Long-term monitoring in the vicinity of the heater during hydration and heating; especially observation and study of coupled THM processes in the near field, i.e. continuous measurements of temperatures, pore pressures, displacements, electric conductivity, and analysis of the gases and water released into the rock by effect of heating; b) Determination of the properties of barrier and host rock done mainly by laboratory and in situ experiments, i.e. general mechanical and mineralogical properties, mechanical state in-situ, and changes induced by the experiment; c) Study of the interaction between host rock and bentonite buffer as well as validation and refinement of existing tools for modelling THM processes; d) Study of the behaviour and reliability of instrumentation and measuring techniques, i.e. inspection of sensors after dismantling the experimental setting. To achieve the objectives, the experiment was accompanied by an extensive programme of continuous monitoring, experimental investigations on-site as well as in laboratories, and numerical modelling of the coupled THM processes. Finally, the experiment was dismantled to provide laboratory specimens of post-heating buffer and host rock material. The continuous monitoring of the experiment by a multitude of sensors (for temperature, pore pressure, total pressure, relative

  8. Review of the sorption of radionuclides on the bedrock of Haestholmen and on construction and backfill materials of a final repository for reactor wastes

    International Nuclear Information System (INIS)

    Kulmala, S.; Hakanen, M.

    1992-10-01

    Imatran Voima Oy (IVO) has plans to build a final repository for reactor wastes in the bedrock of the nuclear power plant site at Haestholmen, Loviisa. This report summarizes the sorption studies of radionuclides in Finnish bedrock performed at the Department of Radiochemistry, University of Helsinki. The values of mass distribution ratios, K d , and surface distribution ratios, K a ; of carbon, calsium, Zirconium, niobium, cobalt, nickel, strontium, cesium, uranium, plutonium, americium, thorium, chlorine, iodine and technetium are surveyed. Special attention is paid to the sorption data for construction and backfill materials of rector waste repository and the bedrock of Haestholmen. Safety assessment of a repository includes calculations of migration of the waste element in construction materials and backfill in the nearfield and in bedrock. Retardation by sorption of waste nuclides compared to groundwater flow is described by using distribution ratios between solid materials and water. (orig.)

  9. Influence of temperature elevation on the sealing performance of a potential buffer material for a high-level radioactive waste repository

    International Nuclear Information System (INIS)

    Cho, W.-J.; Lee, J.-O.; Kang, C.-H.

    2000-01-01

    The sealing performance of buffer material in a high-level waste repository depends largely upon the hydraulic conductivity, the swelling pressure, and the dissolution of organic carbon in the buffer material. Temperature effects on these properties were evaluated. The hydraulic conductivity and the swelling pressure of compacted bentonite increase with increasing temperature, but the effect of temperature elevation is not large. The dissolution of organic carbon in bentonite also increases with increasing temperature, but the resultant aqueous concentrations of organic carbon in bentonite suspensions are less than those of deep groundwater in granite. Therefore, the organic carbon dissolved from the bentonite will not cause a significant increase in the organic carbon content of deep groundwater in the repository environment. Overall, temperature effects on the sealing performance of buffer material in a waste repository is not important, if the maximum temperature is maintained below 100 deg. C

  10. In-situ experiments on bentonite-based buffer and sealing materials at the Mont Terri rock laboratory (Switzerland)

    Energy Technology Data Exchange (ETDEWEB)

    Wieczorek, K. [Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) GmbH, Braunschweig (Germany); Gaus, I. [National Cooperative for the Disposal of Radioactive Waste (NAGRA), Wettingen (Switzerland); Mayor, J. C. [Empresa Nacional de Residuos Radiactivos SA (ENRESA), Madrid (Spain); and others

    2017-04-15

    Repository concepts in clay or crystalline rock involve bentonite-based buffer or seal systems to provide containment of the waste and limit advective flow. A thorough understanding of buffer and seal evolution is required to make sure the safety functions are fulfilled in the short and long term. Experiments at the real or near-real scale taking into account the interaction with the host rock help to make sure the safety-relevant processes are identified and understood and to show that laboratory-scale findings can be extrapolated to repository scale. Three large-scale experiments on buffer and seal properties performed in recent years at the Mont Terri rock laboratory are presented in this paper: The 1:2 scale HE-E heater experiment which is currently in operation, and the full-scale engineered barrier experiment and the Borehole Seal experiment which have been completed successfully in 2014 and 2012, respectively. All experiments faced considerable difficulties during installation, operation, evaluation or dismantling that required significant effort to overcome. The in situ experiments show that buffer and seal elements can be constructed meeting the expectations raised through small-scale testing. It was, however, also shown that interaction with the host rock caused additional effects in the buffer or seal that could not always be quantified or even anticipated from the experience of small-scale tests (such as re-saturation by pore-water from the rock, interaction with the excavation damaged zone in terms of preferential flow or mechanical effects). This led to the conclusion that testing of the integral system buffer/rock or seal/rock is needed. (authors)

  11. Brine and Gas Flow Patterns Between Excavated Areas and Disturbed Rock Zone in the 1996 Performance Assessment for the Waste Isolation Pilot Plant for a Single Drilling Intrusion that Penetrates Repository and Castile Brine Reservoir

    International Nuclear Information System (INIS)

    Economy, Kathleen M.; Helton, Jon Craig; Vaughn, Palmer

    1999-01-01

    The Waste Isolation Pilot Plant (WIPP), which is located in southeastern New Mexico, is being developed for the geologic disposal of transuranic (TRU) waste by the U.S. Department of Energy (DOE). Waste disposal will take place in panels excavated in a bedded salt formation approximately 2000 ft (610 m) below the land surface. The BRAGFLO computer program which solves a system of nonlinear partial differential equations for two-phase flow, was used to investigate brine and gas flow patterns in the vicinity of the repository for the 1996 WIPP performance assessment (PA). The present study examines the implications of modeling assumptions used in conjunction with BRAGFLO in the 1996 WIPP PA that affect brine and gas flow patterns involving two waste regions in the repository (i.e., a single waste panel and the remaining nine waste panels), a disturbed rock zone (DRZ) that lies just above and below these two regions, and a borehole that penetrates the single waste panel and a brine pocket below this panel. The two waste regions are separated by a panel closure. The following insights were obtained from this study. First, the impediment to flow between the two waste regions provided by the panel closure model is reduced due to the permeable and areally extensive nature of the DRZ adopted in the 1996 WIPP PA, which results in the DRZ becoming an effective pathway for gas and brine movement around the panel closures and thus between the two waste regions. Brine and gas flow between the two waste regions via the DRZ causes pressures between the two to equilibrate rapidly, with the result that processes in the intruded waste panel are not isolated from the rest of the repository. Second, the connection between intruded and unintruded waste panels provided by the DRZ increases the time required for repository pressures to equilibrate with the overlying and/or underlying units subsequent to a drilling intrusion. Third, the large and areally extensive DRZ void volumes is a

  12. Brine and Gas Flow Patterns Between Excavated Areas and Disturbed Rock Zone in the 1996 Performance Assessment for the Waste Isolation Pilot Plant for a Single Drilling Intrusion that Penetrates Repository and Castile Brine Reservoir

    Energy Technology Data Exchange (ETDEWEB)

    ECONOMY,KATHLEEN M.; HELTON,JON CRAIG; VAUGHN,PALMER

    1999-10-01

    The Waste Isolation Pilot Plant (WIPP), which is located in southeastern New Mexico, is being developed for the geologic disposal of transuranic (TRU) waste by the U.S. Department of Energy (DOE). Waste disposal will take place in panels excavated in a bedded salt formation approximately 2000 ft (610 m) below the land surface. The BRAGFLO computer program which solves a system of nonlinear partial differential equations for two-phase flow, was used to investigate brine and gas flow patterns in the vicinity of the repository for the 1996 WIPP performance assessment (PA). The present study examines the implications of modeling assumptions used in conjunction with BRAGFLO in the 1996 WIPP PA that affect brine and gas flow patterns involving two waste regions in the repository (i.e., a single waste panel and the remaining nine waste panels), a disturbed rock zone (DRZ) that lies just above and below these two regions, and a borehole that penetrates the single waste panel and a brine pocket below this panel. The two waste regions are separated by a panel closure. The following insights were obtained from this study. First, the impediment to flow between the two waste regions provided by the panel closure model is reduced due to the permeable and areally extensive nature of the DRZ adopted in the 1996 WIPP PA, which results in the DRZ becoming an effective pathway for gas and brine movement around the panel closures and thus between the two waste regions. Brine and gas flow between the two waste regions via the DRZ causes pressures between the two to equilibrate rapidly, with the result that processes in the intruded waste panel are not isolated from the rest of the repository. Second, the connection between intruded and unintruded waste panels provided by the DRZ increases the time required for repository pressures to equilibrate with the overlying and/or underlying units subsequent to a drilling intrusion. Third, the large and areally extensive DRZ void volumes is a

  13. Developing an Experimental Simulation Method for Rock Avalanches: Fragmentation Behavior of Brittle Analogue Material

    Science.gov (United States)

    Thordén Haug, Øystein; Rosenau, Matthias; Leever, Karen; Oncken, Onno

    2013-04-01

    Gravitational mass movement on earth and other planets show a scale dependent behavior, of which the physics is not fully understood. In particular, the runout distance for small to medium sized landslides (volume dynamics control small and large landslides/rock avalanches. Several mechanisms have been proposed to explain this scale dependent behavior, but no consensus has been reached. Experimental simulations of rock avalanches usually involve transport of loose granular material down a chute. Though such granular avalanche models provide important insights into avalanche dynamics, they imply that the material fully disintegrate instantaneously. Observations from nature, however, suggests that a transition from solid to "liquid" occurs over some finite distance downhill, critically controlling the mobility and energy budget of the avalanche. Few experimental studies simulated more realistically the material failing during sliding and those were realized in a labscale centrifuge, where the range of volumes/scales is limited. To develop a new modeling technique to study the scale dependent runout behavior of rock avalanches, we designed, tested and verified several brittle materials allowing fragmentation to occur under normal gravity conditions. According to the model similarity theory, the analogue material must behave dynamically similar to the rocks in natural rock avalanches. Ideally, the material should therefore deform in a brittle manner with limited elastic and ductile strains up to a certain critical stress, beyond which the material breaks and deforms irreversibly. According to scaling relations derived from dimensional analysis and for a model-to-prototype length ratio of 1/1000, the appropriate yield strength for an analogue material is in the order of 10 kPa, friction coefficient around 0.8 and stiffness in the order of MPa. We used different sand (garnet, quartz) in combination with different matrix materials (sugar, salt, starch, plaster) to cement

  14. The hydrothermal stability of cement sealing materials in the potential Yucca Mountain high level nuclear waste repository

    International Nuclear Information System (INIS)

    Krumhansl, J.L.; Hinkebein, T.E.; Myers, J.

    1991-01-01

    Cementitious materials, together with other materials, are being considered to seal a potential repository at Yucca Mountain. A concern with cementitious materials is the chemical and mineralogic changes that may occur as these materials age while in contact with local ground waters. A combined theoretical and experimental approach was taken to determine the ability to theoretically predict mineralogic changes. The cementitious material selected for study has a relatively low Ca:Si ratio approaching that of the mineral tobermorite. Samples were treated hydrothermally at 200 degrees C with water similar to that obtained from the J-13 well on the Nevada Test Site. Post-test solutions were analyzed for pH as well as dissolved K, Na, Ca, Al, and Si. Solid phases formed during these experiments were characterized by scanning electron microscopy and X- ray diffraction. These findings were compared with predictions made by the geochemical modeling code EQ3NR/E06. It was generally found that there was good agreement between predicted and experimental results

  15. Study of Experiment on Rock-like Material Consist of fly-ash, Cement and Mortar

    Science.gov (United States)

    Nan, Qin; Hongwei, Wang; Yongyan, Wang

    2018-03-01

    Study the uniaxial compression test of rock-like material consist of coal ash, cement and mortar by changing the sand cement ratio, replace of fine coal, grain diameter, water-binder ratio and height-diameter ratio. We get the law of four factors above to rock-like material’s uniaxial compression characteristics and the quantitative relation. The effect law can be sum up as below: sample’s uniaxial compressive strength and elasticity modulus tend to decrease with the increase of sand cement ratio, replace of fine coal and water-binder ratio, and it satisfies with power function relation. With high ratio increases gradually, the uniaxial compressive strength and elastic modulus is lower, and presents the inverse function curve; Specimen tensile strength decreases gradually with the increase of fly ash. By contrast, uniaxial compression failure phenomenon is consistent with the real rock common failure pattern.

  16. Test plan: Sealing of the Disturbed Rock Zone (DRZ), including Marker Bed 139 (MB139) and the overlying halite, below the repository horizon, at the Waste Isolation Pilot Plant

    International Nuclear Information System (INIS)

    Ahrens, E.H.

    1992-05-01

    This test plan describes activities intended to demonstrate equipment and techniques for producing, injecting, and evaluating microfine cementitious grout. The grout will be injected in fractured rock located below the repository horizon at the Waste Isolation Pilot Plant (WIPP). These data are intended to support the development of the Alcove Gas Barrier System (AGBS), the design of upcoming, large-scale seal tests, and ongoing laboratory evaluations of grouting efficacy. Degradation of the grout will be studied in experiments conducted in parallel with the underground grouting experiment

  17. Review of the potential effects of alkaline plume migration from a cementitious repository for radioactive waste

    International Nuclear Information System (INIS)

    Savage, D.

    1997-01-01

    Extensive use of cement and concrete is envisaged in the construction of geological repositories for low and intermediate-level radioactive wastes, both for structural, and encapsulation and backfilling purposes. Saturation of these materials with groundwater may occur in the post-closure period of disposal, producing a hyperalkaline pore fluid with a pH in the range 10-13.5. These pore fluids have the potential to migrate from the repository according to local groundwater flow conditions and react chemically with the host rock. These chemical reactions may affect the rock's capacity to retard the migration of radionuclides released from the repository after the degradation of the waste packages. The effects of these chemical reactions on the behaviour of the repository rock as a barrier to waste migration need to be investigated for the purposes of assessing the safety of the repository design (so-called 'safety assessment' or 'performance assessment'). The objectives of the work reported here were to: identify those processes influencing radionuclide mobility in the geosphere which could be affected by plume migration; review literature relevant to alkali-rock reaction; contact organisations carrying out relevant research and summarise their current and future activities; and make recommendations how the effects of plume migration can be incorporated into models of repository performance assessment. (author)

  18. Numerical analysis for long-term stability of disposal facility considering thermal and hydraulic effect. Uncoupled analysis for long-term deformation of rock and buffer material and for transport of heat and water

    International Nuclear Information System (INIS)

    Sawada, Masataka; Okada, Tetsumi; Hasegawa, Takuma

    2004-01-01

    For the early realization of HLW geological repository and for its rational and economical design and safety assessment, it is important to evaluate the stability of repository facilities in deep underground, where high temperature, earth pressure and underground water flow affect the stability. This report discusses the numerical approaches that are useful for attaining these objectives. One of the efficient approaches is to develop models capable of simulating coupled thermo-hydro-mechanical (T-H-M) processes. Several T-H-M coupled simulation codes have been proposed and the international cooperative research project DECOVALEX has been held from 1991 in order to develop and validate the T-H-M coupled simulations. But mechanical models adopted in these simulation codes are too simple to be applied to the evaluation of long-term interaction of materials that show nonlinear mechanical behavior (especially in the case that surrounding rock is soft sedimentary rock). Before simulating the long-term and coupled phenomena, uncoupled simulations for four phenomena (creep behavior of surrounding rock mass, consolidation and deformation behavior of buffer material, transport of water, and transport of heat) are conducted using various parameters and boundary condition sets. From the results of those simulations, following conclusions are obtained: (1) swelling property of buffer material is important to evaluate mechanical behavior of barrier materials, (2) hydraulic properties of natural barrier can be more important than that of buffer material because suction effect of buffer material is so strong that transport of water in the buffer material is fast, (3) change of thermal properties and filling of gaps caused by water saturation of buffer material have a strong effect on the temperature field. On the next stage, we will develop a T-H-M coupled simulation code to evaluate the mechanical interaction between barrier materials based on the above study. (author)

  19. An analysis of the factors affecting the hydraulic conductivity and swelling pressure of Kyungju ca-bentonite for use as a clay-based sealing material for a high level waste repository

    International Nuclear Information System (INIS)

    Cho, Won Jin; Lee, Jae Owen; Kwon, Sang Ki

    2012-01-01

    The buffer and backfill are important components of the engineered barrier system in a high-level waste repository, which should be constructed in a hard rock formation at a depth of several hundred meters below the ground surface. The primary function of the buffer and backfill is to seal the underground excavation as a preferred flow path for radionuclide migration from the deposited high-level waste. This study investigates the hydraulic conductivity and swelling pressure of Kyungju Ca-bentonite, which is the candidate material for the buffer and backfill in the Korean reference high-level waste disposal system. The factors that influence the hydraulic conductivity and swelling pressure of the buffer and backfill are analyzed. The factors considered are the dry density, the temperature, the sand content, the salinity and the organic carbon content. The possibility of deterioration in the sealing performance of the buffer and backfill is also assessed.

  20. Alkaline degradation of organic materials contained in TRU wastes under repository conditions

    International Nuclear Information System (INIS)

    Otsuka, Yoshiki; Banba, Tsunetaka

    2007-09-01

    Alkaline degradation tests for 9 organic materials were conducted under the conditions of TRU waste disposal: anaerobic alkaline conditions. The tests were carried out at 90degC for 91 days. The sample materials for the tests were selected from the standpoint of constituent organic materials of TRU wastes. It has been found that cellulose and plastic solidified products are degraded relatively easily and that rubbers are difficult to degrade. It could be presumed that the alkaline degradation of organic materials occurs starting from the functional group in the material. Therefore, the degree of degradation difficulty is expected to be dependent on the kinds of functional group contained in the organic material. (author)

  1. Some Materials Degradation Issues in the U.S. High-Level Nuclear Waste Repository Study (The Yucca Mountain Project)

    Energy Technology Data Exchange (ETDEWEB)

    F. Hua; P. Pasupathi; N. Brown; K. Mon

    2005-09-19

    The safe disposal of radioactive waste requires that the waste be isolated from the environment until radioactive decay has reduced its toxicity to innocuous levels for plants, animals, and humans. All of the countries currently studying the options for disposing of high-level nuclear waste (HLW) have selected deep geologic formations to be the primary barrier for accomplishing this isolation. In U.S.A., the Nuclear Waste Policy Act of 1982 (as amended in 1987) designated Yucca Mountain in Nevada as the potential site to be characterized for high-level nuclear waste (HLW) disposal. Long-term containment of waste and subsequent slow release of radionuclides into the geosphere will rely on a system of natural and engineered barriers including a robust waste containment design. The waste package design consists of a highly corrosion resistant Ni-based Alloy 22 cylindrical barrier surrounding a Type 316 stainless steel inner structural vessel. The waste package is covered by a mailbox-shaped drip shield composed primarily of Ti Grade 7 with Ti Grade 24 structural support members. The U.S. Yucca Mountain Project has been studying and modeling the degradation issues of the relevant materials for some 20 years. This paper reviews the state-of-the-art understanding of the degradation processes based on the past 20 years studies on Yucca Mountain Project (YMP) materials degradation issues with focus on interaction between the in-drift environmental conditions and long-term materials degradation of waste packages and drip shields within the repository system during the 10,000 years regulatory period. This paper provides an overview of the current understanding of the likely degradation behavior of the waste package and drip shield in the repository after the permanent closure of the facility. The degradation scenario discussed in this paper include aging and phase instability, dry oxidation, general and localized corrosion, stress corrosion cracking and hydrogen induced

  2. Some Materials Degradation Issues in the U.S. High-Level Nuclear Waste Repository Study (The Yucca Mountain Project)

    International Nuclear Information System (INIS)

    Hua, F.; Pasupathi, P.; Brown, N.; Mon, K.

    2005-01-01

    The safe disposal of radioactive waste requires that the waste be isolated from the environment until radioactive decay has reduced its toxicity to innocuous levels for plants, animals, and humans. All of the countries currently studying the options for disposing of high-level nuclear waste (HLW) have selected deep geologic formations to be the primary barrier for accomplishing this isolation. In U.S.A., the Nuclear Waste Policy Act of 1982 (as amended in 1987) designated Yucca Mountain in Nevada as the potential site to be characterized for high-level nuclear waste (HLW) disposal. Long-term containment of waste and subsequent slow release of radionuclides into the geosphere will rely on a system of natural and engineered barriers including a robust waste containment design. The waste package design consists of a highly corrosion resistant Ni-based Alloy 22 cylindrical barrier surrounding a Type 316 stainless steel inner structural vessel. The waste package is covered by a mailbox-shaped drip shield composed primarily of Ti Grade 7 with Ti Grade 24 structural support members. The U.S. Yucca Mountain Project has been studying and modeling the degradation issues of the relevant materials for some 20 years. This paper reviews the state-of-the-art understanding of the degradation processes based on the past 20 years studies on Yucca Mountain Project (YMP) materials degradation issues with focus on interaction between the in-drift environmental conditions and long-term materials degradation of waste packages and drip shields within the repository system during the 10,000 years regulatory period. This paper provides an overview of the current understanding of the likely degradation behavior of the waste package and drip shield in the repository after the permanent closure of the facility. The degradation scenario discussed in this paper include aging and phase instability, dry oxidation, general and localized corrosion, stress corrosion cracking and hydrogen induced

  3. Progress report on the results of testing advanced conceptual design metal barrier materials under relevant environmental conditions for a tuff repository

    International Nuclear Information System (INIS)

    McCright, R.D.; Halsey, W.G.; Van Konynenburg, R.A.

    1987-12-01

    This report discusses the performance of candidate metallic materials envisioned for fabricating waste package containers for long-term disposal at a possible geological repository at Yucca Mountain, Nevada. Candidate materials include austenitic iron-base to nickel-base alloy (AISI 304L, AISI 316L, and Alloy 825), high-purity copper (CDA 102), and copper-base alloys (CDA 613 and CDA 715). Possible degradation modes affecting these container materials are identified in the context of anticipated environmental conditions at the repository site. Low-temperature oxidation is the dominant degradation mode over most of the time period of concern (minimum of 300 yr to a maximum of 1000 yr after repository closure), but various forms of aqueous corrosion will occur when water infiltrates into the near-package environment. The results of three years of experimental work in different repository-relevant environments are presented. Much of the work was performed in water taken from Well J-13, located near the repository, and some of the experiments included gamma irradiation of the water or vapor environment. The influence of metallurgical effects on the corrosion and oxidation resistance of the material is reviewed; these effects result from container fabrication, welding, and long-term aging at moderately elevated temperatures in the repository. The report indicates the need for mechanisms to understand the physical/chemical reactions that determine the nature and rate of the different degradation modes, and the subsequent need for models based on these mechanisms for projecting the long-term performance of the container from comparatively short-term laboratory data. 91 refs., 17 figs., 16 tabs

  4. Aespoe Hard Rock Laboratory Annual Report 1999

    International Nuclear Information System (INIS)

    2000-08-01

    Plug Test includes tests of backfill materials and emplacement methods and a test of a full-scale plug. The backfill and rock has been instrumented with about 230 transducers for measuring the thermo-hydro-mechanical processes.Saturation is in progress and is expected to take 1-2 years. The Long Term Tests of Buffer Material aim to validate models of buffer performance at standard KBS-3 repository conditions,and at quantifying clay buffer alteration processes at adverse conditions.The 4 long term test parcels and the additional 1-year parcel have been installed. Nine organisations from eight countries are currently participating in the Aespoe Hard Rock Laboratory in addition to SKB

  5. Aespoe Hard Rock Laboratory Annual Report 1999

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-08-01

    Plug Test includes tests of backfill materials and emplacement methods and a test of a full-scale plug. The backfill and rock has been instrumented with about 230 transducers for measuring the thermo-hydro-mechanical processes.Saturation is in progress and is expected to take 1-2 years. The Long Term Tests of Buffer Material aim to validate models of buffer performance at standard KBS-3 repository conditions,and at quantifying clay buffer alteration processes at adverse conditions.The 4 long term test parcels and the additional 1-year parcel have been installed. Nine organisations from eight countries are currently participating in the Aespoe Hard Rock Laboratory in addition to SKB.

  6. Web-Based Learning Materials for Higher Education: The MERLOT Repository

    Science.gov (United States)

    Orhun, Emrah

    2004-01-01

    MERLOT (Multimedia Educational Resource for Learning and Online Teaching) is a web-based open resource designed primarily for faculty and students in higher education. The resources in MERLOT include over 8,000 learning materials and support materials from a wide variety of disciplines that can be integrated within the context of a larger course.…

  7. Materials interactions test methods to measure radionuclide release from waste forms under repository-relevant conditions

    International Nuclear Information System (INIS)

    Strickert, R.G.; Erikson, R.L.; Shade, J.W.

    1984-10-01

    At the request of the Basalt Waste Isolation Project, the Materials Characterization Center has collected and developed a set of procedures into a waste form compliance test method (MCC-14.4). The purpose of the test is to measure the steady-state concentrations of specified radionuclides in solutions contacting a waste form material. The test method uses a crushed waste form and basalt material suspended in a synthetic basalt groundwater and agitated for up to three months at 150 0 C under anoxic conditions. Elemental and radioisotopic analyses are made on filtered and unfiltered aliquots of the solution. Replicate experiments are performed and simultaneous tests are conducted with an approved test material (ATM) to help ensure precise and reliable data for the actual waste form material. Various features of the test method, equipment, and test conditions are reviewed. Experimental testing using actinide-doped borosilicate glasses are also discussed. 9 references, 2 tables

  8. Cementitious near-field sorption data bases for performance assessment of a L/ILW repository in a Palfris marl host rock. CEM-94: update I, June 1997

    International Nuclear Information System (INIS)

    Bradbury, M.H.; Loon, L.R. van

    1998-01-01

    This report is an update on an earlier cementitious sorption data base (SDB) prepared by Bradbury and Sarott (1994). The aim is to review any new information or data which have become available in the intervening time and modify the existing SDB appropriately. Discussions will be confined predominantly to areas which have led to significant changes to or reappraisals of the data/values or procedures for obtaining/modifying them. From this point of view this update and the previous SDB are closely related and belong together. The complexation of radionuclides with organic ligands from the chemical degradation of cellulose, and the subsequent negative effects on sorption properties, were identified as being processes of great importance. Since 1994 significant progress has been made in this field and a major part of this work is devoted to a reassessment of the impact of 'organics' on near-field sorption. In particular, the very conservative assumptions which had been made previously because of the general lack of good quality data available at that time, could be replaced by realistic parameter estimates based on new knowledge. For example, maximum likely concentrations of cellulose degradation products and cement additives in the cement pore waters could be calculated allowing the potential effects of these organic ligands on sorption to be bounded. Sorption values for safety relevant radionuclides corresponding to the three broad stages of cement/concrete degradation during the lifetime of the repository are presented in tabulated form. The influence of the wide variety of organic ligands existing in the different waste categories, SMA-1 to SMA-4, is quantified in terms of sorption reduction factors. In the compilation of this cement SDB update, radionuclide uptake onto the vast quantities of aggregate materials and corrosion products from iron/steel was not taken into account. (author) 10 figs., 8 tabs., refs

  9. Heat production / host rock compatibility; Waermeentwicklung / Gesteinsvertraeglichkeit

    Energy Technology Data Exchange (ETDEWEB)

    Meleshyn, A.; Weyand, T.; Bracke, G.; Kull, H.; Wieczorek, K.

    2016-05-15

    For the final high-level radioactive waste repository potential host rock formations are either rock salt or clays (Kristallin). Heat generating waste (decay heat of the radioactive materials) can be absorbed by the host rock. The effect of temperature increase on the thermal conductivity, the thermal expansion and the mechanical properties of salt, Kristallin, clays and argilliferous geotechnical barriers are described. Further issues of the report are the mineralogical behavior, phase transformations, hydrochemistry, microbial processes, gas formation, thermochemical processes and gas ingress. Recommendations for further research are summarized.

  10. Sealing a nuclear waste repository in Columbia river basalt: preliminary results

    International Nuclear Information System (INIS)

    Hodges, F.N.

    1980-01-01

    The long containment time required of repositories for nuclear waste (10 4 to 10 6 years) requires that materials used for repository seals be stable in the geologic environment of the repository and of proven longevity. A list of candidate materials for sealing a repository in Columbia River Basalts has been prepared and refined through laboratory testing. The most feasible techniques for emplacing preferred plug materials have been identified and the resultant plugs have been evaluated on the basis of design functions. Preconceptual designs for tunnel, shaft, and borehole seals consist of multiple zone plugs with each zone fulfilling one or more design functions. Zones of disturbed rock around tunnels and shafts, resulting from excavation and subsequent stress release, are zones of higher permeability and of possible fluid migration. In preliminary designs the disturbed zones are blocked by cut-off collars filled with low permeability materials

  11. Rheological characteristics of waste rock materials in abandoned mine deposit and debris flow hazards

    Science.gov (United States)

    Jeong, Sueng-Won; Lee, Choonoh; Cho, Yong-Chan; Wu, Ying-Hsin

    2015-04-01

    In Korea, approximately 5,000 metal mines are spread, but 50% of them are still abandoned without any proper remediation and cleanup. Summer heavy rainfall can result in the physicochemical modification of waste rock materials in the mountainous. From the geotechnical monitoring and field investigation, there are visible traces of mass movements every year. Soil erosion is one of severe phenomena in the study area. In particular, study area is located in the upper part of the Busan Metropolitan City and near the city's water supply. With respect to the supply of drinking water and maintenance of ecological balance, proper disposal of waste rock materials is required. For this reason, we examine the rheological properties of waste rock materials as a function of solid content using a ball- and vane-penetrated rheometer. In the flow curves, which are the relationship between the shear stress and shear rate of waste rock materials, we found that the soil samples exhibited a shear thinning beahivor regardless of solid content. The Bingham, Herschel-Bulkley, Power-law, and Papanastasiou models are used to determine the rheological properties. Assuming that the soil samples behaved as the viscoplastic behavior, the yield stress and viscosity are determined for different water contents. As a result, there are clear relationships between the solid content and rheological values (i.e., Bingham yield stress and plastic viscosity). From these relationships, the maximum and minimum of Bingham yield stresses are ranged from 100 to 2000 Pa. The debris flow mobilization is analysed using a 1D BING and 2D Debris flow models. In addition, the effect of wall slip and test apparatus are discussed.

  12. Validation of a New Elastoplastic Constitutive Model Dedicated to the Cyclic Behaviour of Brittle Rock Materials

    Science.gov (United States)

    Cerfontaine, B.; Charlier, R.; Collin, F.; Taiebat, M.

    2017-10-01

    Old mines or caverns may be used as reservoirs for fuel/gas storage or in the context of large-scale energy storage. In the first case, oil or gas is stored on annual basis. In the second case pressure due to water or compressed air varies on a daily basis or even faster. In both cases a cyclic loading on the cavern's/mine's walls must be considered for the design. The complexity of rockwork geometries or coupling with water flow requires finite element modelling and then a suitable constitutive law for the rock behaviour modelling. This paper presents and validates the formulation of a new constitutive law able to represent the inherently cyclic behaviour of rocks at low confinement. The main features of the behaviour evidenced by experiments in the literature depict a progressive degradation and strain of the material with the number of cycles. A constitutive law based on a boundary surface concept is developed. It represents the brittle failure of the material as well as its progressive degradation. Kinematic hardening of the yield surface allows the modelling of cycles. Isotropic softening on the cohesion variable leads to the progressive degradation of the rock strength. A limit surface is introduced and has a lower opening than the bounding surface. This surface describes the peak strength of the material and allows the modelling of a brittle behaviour. In addition a fatigue limit is introduced such that no cohesion degradation occurs if the stress state lies inside this surface. The model is validated against three different rock materials and types of experiments. Parameters of the constitutive laws are calibrated against uniaxial tests on Lorano marble, triaxial test on a sandstone and damage-controlled test on Lac du Bonnet granite. The model is shown to reproduce correctly experimental results, especially the evolution of strain with number of cycles.

  13. Global thermo-mechanical effects from a KBS-3 type repository

    International Nuclear Information System (INIS)

    Hakami, E.; Olofsson, Stig-Olof

    1998-01-01

    The objective of this study has been to identify the global thermo-mechanical effects in the bedrock hosting a nuclear waste repository. Numerical thermo-mechanical modeling using distinct element models was performed. The number of fracture zones, the heat intensity of the waste, the material properties of the rock mass and the boundary conditions of the models were varied. Different models for multi-level repositories were also analyzed and compared to the main single-level case. Further, the global influence from the excavation of repository tunnels and deposition holes was examined by introducing weaker rock mass material properties in the repository region of one model. The maximum compression stress obtained for the main model is 44 MPa and occurs at the repository level after about 100 years of deposition. Due to thermal expansion, the rock mass displaces upward, and the maximum heave at the ground surface after 1000 years is calculated to be 16 cm. In the area close to the ground surface the horizontal stresses reduce, causing the rock to yield in tension down to a depth of about 80 meters. The fracture zones show opening displacements at shallow depths and closing and shearing at the repository level. The maximum displacements are 0.3-2.5 cm for closing, 0.0-0.8 cm for opening and 0.2-2.2 cm for shearing. The resultant stresses and displacements depend in large part on the assumptions made concerning the heat intensity of the waste. In the main model, an initial heat intensity of 10 W/m 2 is assumed, which gives larger effects than the case with 6 W/m 2 . Another important input parameter for the analysis is the Young's modulus of the rock mass. In the main model, a value of 30 GPa is assumed. Higher values of Young's modulus give larger thermo-mechanical effects. All multi-level repository layouts give rise to higher temperatures than the single-level layout, causing the compressive stresses to increase more at the repository level. The multi

  14. Use and Misuse of Material Transfer Agreements: Lessons in Proportionality from Research, Repositories, and Litigation

    OpenAIRE

    Bubela, Tania; Guebert, Jenilee; Mishra, Amrita

    2015-01-01

    Material transfer agreements exist to facilitate the exchange of materials and associated data between researchers as well as to protect the interests of the researchers and their institutions. But this dual mandate can be a source of frustration for researchers, creating administrative burdens and slowing down collaborations. We argue here that in most cases in pre-competitive research, a simple agreement would suffice; the more complex agreements and mechanisms for their negotiation should ...

  15. Methodological developments and materials in salt-rock preparation for irradiation experiments

    International Nuclear Information System (INIS)

    Garcia Celma, A.; Van Wees, H.; Miralles, L.

    1991-01-01

    For the first time synthetic salt-rock samples have been produced. Production and preparation of those samples and of other types of rock-salt for experiments and observation require many special handlings. We applied technical knowledge already developed by the HPT Laboratory of the Geology Department of the Rijksuniversiteit Utrecht (high pressure techniques, salt-rock preparation), and by the workshops of the ECN, Petten, and FDO, Amsterdam (mechanical precision). Procedures have been applied and/or modified to solve specific problems. Many of them were never reported before. Moreover, new techniques have been developed. Rock-salt samples have been machined, sawn, ground, glued, etc., with a maximum of precision, a minimum of damage and in dry conditions (without water). Etching, peeling and thin section production has been carried out on irradiated and unirradiated samples. Valves, end pieces, jackets, etc. have been tested and/or produced. These handlings were directed to produce samples for the HAW experiment. Their development required not only knowledge, but also a lot of trial, failures and time. To avoid repetition of this effort, the procedures, materials, instruments and their characteristics are described in detail in this report

  16. Development of a technical concept for a generic final repository for heat-generating wastes and spent fuel elements in crystalline rock formations in Germany. Final report; Entwicklung eines technischen Konzeptes fuer ein generisches Endlager fuer waermeentwickelnde Abfaelle und ausgediente Brennelemente im Kristallingestein in Deutschland. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Bertrams, Niklas; Herold, Philipp; Herold, Maxi; Krone, Juergen; Lommerzheim, Andree; Prignitz, Sabine; Kuate, Eric Simo

    2017-09-15

    The research project concerning the development of a generic concept for final repositories in crystalline rock formations has identified three different concepts for long-term safe enclosure efficacy: (i) the KBS-3 concept as pursued in Sweden and Finland based on corrosion resistant copper containers and bentonite buffers in vertical bore holes; (ii) The concept of ''multiple enclosure efficient rock zones'', based on several spatially separated rock zones that allow the demonstration of efficient enclosure; (iii) the concept of a ''superposed enclosure efficient rock zone'', where a sedimentary coverage of the crystalline host rock (for instance clay or salt) shows enclosure efficacy. For each of these concepts a separate final repository concept was developed covering the construction of shafts, ramps and transport routes, the preparation of boreholes, and the backfilling and closure technology, the planning of mine buildings, ventilation, time and cost estimation.

  17. Engineered barrier development for a nuclear waste repository in basalt

    International Nuclear Information System (INIS)

    Smith, M.J.

    1980-05-01

    The BWIP Engineered Barrier Program has been developed to provide an integrated approach to the development of site-specific Engineered Barrier assemblages for a repository located in basalt. The goal of this program is to specify engineered and natural barriers which will ensure that nuclear and non-radioactive hazardous materials emplaced in a repository in basalt do not exceed acceptable rates of release to the biosphere. A wide range of analytical and experimental activities related to the basalt repository environment, waste package environment, waste/barrier/rock interactions, and barrier performance assessment provide the basis for selection of systems capable of meeting licensing requirements. Work has concentrated on specifying and testing natural and man-made materials which can be used to plug boreholes in basalt and which can be used as multiple barriers to surround nuclear waste forms and containers. The Engineered Barriers Program is divided into two major activities: multiple barrier studies and borehole plugging. 8 figures, 4 tables

  18. Review of durability of cementitious engineered barriers in repository environments

    International Nuclear Information System (INIS)

    Parrott, L.J.; Lawrence, C.D.

    1992-01-01

    This report is concerned with the durability of cementitious engineered barriers in a repository for low and intermediate level nuclear waste. Following the introduction the second section of the review identifies the environmental conditions associated with a deep, hard rock repository for ILW and LLW that are relevant to the durability of cementitious barriers. Section three examines the microstructure and macrostructure of cementitious materials and considers the physical and chemical processes of radionuclide immobilization. Potential repository applications and compositions of cementitious materials are reviewed in Section four. The main analysis of durability is dealt with in Section five. The different types of cementitious barrier are considered separately and their most probable modes of degradation are analysed. Concluding remarks that highlight critical technical matters are given in Section six. (author)

  19. Thermal stresses in a repository for ultimate storage of high-level radioactive wastes

    International Nuclear Information System (INIS)

    Ehlert, C.

    1981-01-01

    An important factor to be considered in evaluating the suitability of a salt mine as a waste repository is the deformation behaviour of rock salt, as this is the predominant type of rock in this formation. Equations are presented and explained describing the elastic, plastic, and viscoplastic deformation mechanisms contributing to overall rock salt deformation, and use of these equations is made through a specially developed arithmetic method. As there are stratifications and discontinuties in the formation to be considered in the computation, additional criteria are to be taken into account in the integrity considerations, especially the figures of material equations for all other types of rock occurring in the formation. (DG) [de

  20. Corrosion of candidate materials for canister: applications in rock salt formations

    International Nuclear Information System (INIS)

    Azkarate, I.; Madina, V.; Barrio, A. del; Macarro, J.M.

    1994-01-01

    Previous corrosion studies carried out on various metallic materials in typical salt rock environments show that carbon steel and titanium alloys are the most promising candidates for canister applications in this geological formation. Although carbon steels have a low corrosion resistance, they are considered acceptable as corrosion-allowance materials for a thick walled container due to their practical immunity to the localized corrosion phenomena such as stress corrosion cracking, pitting or crevice corrosion. Aiming to improve the performances of these materials, studies on the effect of small additions of Ni and V on the general corrosion are in process. The improvement in the resistance to general corrosion should not be accompanied by a sensitivity to stress corrosion cracking. On the contrary, alfa titanium alloys are considered the most resistant materials to general corrosion in salt brines. However, pitting, are potential deficiencies of this corrosion-resistant materials for a thin walled container. (Author)

  1. Leaching properties of natural aggregates. Rock materials and tills; Lakegenskaper foer naturballast. Bergmaterial och moraener

    Energy Technology Data Exchange (ETDEWEB)

    Ekvall, Annika; Bahr, Bo von; Andersson, Tove; Lax, Kaj; Aakesson, Urban [Swedish National Testing and Research Inst., Boraas (Sweden)

    2006-02-15

    The aim of this project is to produce leaching data for natural aggregates needed for assessment of the environmental impact of alternative materials aimed for use in for example road constructions. Both rock materials and tills are tested. The results shows that very little is leached from natural aggregate. A comparison with landfill criteria for inert waste and the Swedish regulations for drinking water shows that a few samples exceeds the criteria for fluoride ions. All other values are lower then these criteria, and a vast majority of the measurements are below the quantification limit.

  2. Research of mountain rocks of Georgia for using in glass materials industry

    International Nuclear Information System (INIS)

    Gabunia, L.; Gabunia, N.; Jakhva, N.; Napetvaridze, Ts.; Alibegashvili, M.

    2009-01-01

    The article presents the results of the research of mountain rocks of Georgia in various peaces of basalts, andesite, andesite-basalt porphyry and trachite deposits in order to be used in the industry of fiber, glass-crystal materials and fasade decorative tiles. The prospects of basalt and andesite-basalt porphyry use in the industry of fiber materials in the form of monocomponents has been stated. The technology of obtaining of wear resistant, chemically stable and colored glass-crystal on the basis of andesit and trachyte, has been worked out and approved in industry. (author)

  3. Use and misuse of material transfer agreements: lessons in proportionality from research, repositories, and litigation.

    Directory of Open Access Journals (Sweden)

    Tania Bubela

    2015-02-01

    Full Text Available Material transfer agreements exist to facilitate the exchange of materials and associated data between researchers as well as to protect the interests of the researchers and their institutions. But this dual mandate can be a source of frustration for researchers, creating administrative burdens and slowing down collaborations. We argue here that in most cases in pre-competitive research, a simple agreement would suffice; the more complex agreements and mechanisms for their negotiation should be reserved for cases where the risks posed to the institution and the potential commercial value of the research reagents is high.

  4. Use and misuse of material transfer agreements: lessons in proportionality from research, repositories, and litigation.

    Science.gov (United States)

    Bubela, Tania; Guebert, Jenilee; Mishra, Amrita

    2015-02-01

    Material transfer agreements exist to facilitate the exchange of materials and associated data between researchers as well as to protect the interests of the researchers and their institutions. But this dual mandate can be a source of frustration for researchers, creating administrative burdens and slowing down collaborations. We argue here that in most cases in pre-competitive research, a simple agreement would suffice; the more complex agreements and mechanisms for their negotiation should be reserved for cases where the risks posed to the institution and the potential commercial value of the research reagents is high.

  5. 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

  6. Geologic environments for nuclear waste repositories

    Directory of Open Access Journals (Sweden)

    Paleologos Evan K.

    2017-01-01

    Full Text Available High-level radioactive waste (HLW results from spent reactor fuel and reprocessed nuclear material. Since 1957 the scientific consensus is that deep geologic disposal constitutes the safest means for isolating HLW for long timescales. Nuclear power is becoming significant for the Arab Gulf countries as a way to diversify energy sources and drive economic developments. Hence, it is of interest to the UAE to examine the geologic environments currently considered internationally to guide site selection. Sweden and Finland are proceeding with deep underground repositories mined in bedrock at depths of 500m, and 400m, respectively. Equally, Canada’s proposals are deep burial in the plutonic rock masses of the Canadian Shield. Denmark and Switzerland are considering disposal of their relative small quantities of HLW into crystalline basement rocks through boreholes at depths of 5,000m. In USA, the potential repository at Yucca Mountain, Nevada lies at a depth of 300m in unsaturated layers of welded volcanic tuffs. Disposal of low and intermediate-level radioactive wastes, as well as the German HLW repository favour structurally-sound layered salt stata and domes. Our article provides a comprehensive review of the current concepts regarding HLW disposal together with some preliminary analysis of potentially appropriate geologic environments in the UAE.

  7. Fundamental properties of monolithic bentonite buffer material formed by cold isostatic pressing for high-level radioactive waste repository

    International Nuclear Information System (INIS)

    Kawakami, S.; Yamanaka, Y.; Kato, K.; Asano, H.; Ueda, H.

    1999-01-01

    The methods of fabrication, handling, and emplacement of engineered barriers used in a deep geological repository for high level radioactive waste should be planned as simply as possible from the engineering and economic viewpoints. Therefore, a new concept of a monolithic buffer material around a waste package have been proposed instead of the conventional concept with the use of small blocks, which would decrease the cost for buffer material. The monolithic buffer material is composed of two parts of highly compacted bentonite, a cup type body and a cover. As the forming method of the monolithic buffer material, compaction by the cold isostatic pressing process (CIP) has been employed. In this study, monolithic bentonite bodies with the diameter of about 333 mm and the height of about 455 mm (corresponding to the approx. 1/5 scale for the Japanese reference concept) were made by the CIP of bentonite powder. The dry densities: ρd of the bodies as a whole were measured and the small samples were cut from several locations to investigate the density distribution. The swelling pressure and hydraulic conductivity as function of the monolithic body density for CIP-formed specimens were also measured. High density (ρd: 1.4--2.0 Mg/m 3 ) and homogeneous monolithic bodies were formed by the CIP. The measured results of the swelling pressure (3--15 MPa) and hydraulic conductivity (0.5--1.4 x 10 -13 m/s) of the specimens were almost the same as those for the uniaxial compacted bentonite in the literature. It is shown that the vacuum hoist system is an applicable handling method for emplacement of the monolithic bentonite

  8. Radiation effects on materials in the near-field of nuclear waste repository. 1998 annual progress report

    International Nuclear Information System (INIS)

    Ewing, R.C.; Wang, L.M.

    1998-01-01

    'Site restoration activities at DOE facilities and the permanent disposal of nuclear waste generated at DOE facilities involve working with and within various types and levels of radiation fields. Once the nuclear waste is incorporated into a final form, radioactive decay will decrease the radiation field over geologic time scales, but the alpha-decay dose for these solids will still reach values as high as 10 18 alpha-decay events/gm in periods as short as 1,000 years. This dose is well within the range for which important chemical (e.g., increased leach rate) and physical (e.g., volume expansion) changes may occur in crystalline ceramics. Release and sorption of long-lived actinides (e.g., 237 Np) can provide a radiation exposure to backfill materials, and changes in important properties (e.g., cation exchange capacity) may occur. The objective of this research program is to evaluate the long term radiation effects in the materials in the near-field of a nuclear waste repository with accelerated experiments in the laboratory using energetic particles (electrons, ions and neutrons). Experiments on the microstructural evolution during irradiation of two important groups of materials, sheet silicates (e.g., clays) and zeolites (analcime), have been conducted; and studies of radiation-induced changes in chemical properties (e.g. cation exchange capacity) are underway. As of the mid-2nd year of the 3-year project, experiments on the microstructural evolution during irradiation of two important group of materials, sheet silicates (mica) and zeolites (analcime), have been conducted; and studies of radiation-induced changes in chemical properties (e.g., cation exchange capacity) are underway.'

  9. Studies of the behaviour of backfill taking into account the interaction between rock and backfill, and other sealing components at a salina repository

    International Nuclear Information System (INIS)

    Diekmann, N.; Stuehrenberg, D.

    1991-09-01

    According to the present planning level of the designed Gorleben repository, the salt produced by opening up cavities for ultimate disposal will be used as salt fines for backfilling residual cavities after radioactive waste emplacement. The essential function properties of the backfill - compaction and permeability - were studied for salt fines, and the results achieved were discussed. (BBR) [de

  10. ROCKING. A computer program for seismic response analysis of radioactive materials transport AND/OR storage casks

    International Nuclear Information System (INIS)

    Ikushima, Takeshi

    1995-11-01

    The computer program ROCKING has been developed for seismic response analysis, which includes rocking and sliding behavior, of radioactive materials transport and/or storage casks. Main features of ROCKING are as follows; (1) Cask is treated as a rigid body. (2) Rocking and sliding behavior are considered. (3) Impact forces are represented by the spring dashpot model located at impact points. (4) Friction force is calculated at interface between a cask and a floor. (5) Forces of wire ropes against tip-over work only as tensile loads. In the paper, the calculation model, the calculation equations, validity calculations and user's manual are shown. (author)

  11. Corrosion test plan to guide canister material selection and design for a tuff repository

    International Nuclear Information System (INIS)

    McCright, R.D.; van Konynenburg, R.A.; Ballou, L.B.

    1983-11-01

    Corrosion rates and the mode of corrosion attack form a most important basis for selection of canister materials and design of a nuclear waste package. Type 304L stainless steel was selected as the reference material for canister fabrication because of its generally excellent corrosion resistance in water, steam and air. However, 304L may be susceptible to localized and stress-assisted forms of corrosion under certain conditions. Alternative alloys are also investigated; these alloys were chosen because of their improved resistance to these forms of corrosion. The fabrication and welding processes, as well as the glass pouring operation for defense and commercial high-level wastes, may influence the susceptibility of the canister to localized and stress forms of corrosion. 12 references, 2 figures, 4 tables

  12. Electrical Resistance Tomography to Monitor Mitigation of Metal-Toxic Acid-Leachates Ruby Gulch Waste Rock Repository Gilt Edge Mine Superfund Site, South Dakota USA

    Science.gov (United States)

    Versteeg, R.; Heath, G.; Richardson, A.; Paul, D.; Wangerud, K.

    2003-12-01

    At a cyanide heap-leach open-pit mine, 15-million cubic yards of acid-generating sulfides were dumped at the head of a steep-walled mountain valley, with 30 inches/year precipitation generating 60- gallons/minute ARD leachate. Remediation has reshaped the dump to a 70-acre, 3.5:1-sloped geometry, installed drainage benches and runoff diversions, and capped the repository and lined diversions with a polyethylene geomembrane and cover system. Monitoring was needed to evaluate (a) long-term geomembrane integrity, (b) diversion liner integrity and long-term effectiveness, (c) ARD geochemistry, kinetics and pore-gas dynamics within the repository mass, and (d) groundwater interactions. Observation wells were paired with a 600-electrode resistivity survey system. Using near-surface and down-hole electrodes and automated data collection and post-processing, periodic two- and three-dimensional resistivity images are developed to reflect current and changed-conditions in moisture, temperature, geochemical components, and flow-direction analysis. Examination of total resistivity values and time variances between images allows direct observation of liner and cap integrity with precise identification and location of leaks; likewise, if runoff migrates from degraded diversion ditches into the repository zone, there is an accompanying and noticeable change in resistivity values. Used in combination with monitoring wells containing borehole resistivity electrodes (calibrated with direct sampling of dump water/moisture, temperature and pore-gas composition), the resistivity arrays allow at-depth imaging of geochemical conditions within the repository mass. The information provides early indications of progress or deficiencies in de-watering and ARD- mitigation that is the remedy intent. If emerging technologies present opportunities for secondary treatment, deep resistivity images may assist in developing application methods and evaluating the effectiveness of any reagents

  13. Data for the sorption of actinides on candidate materials for use in repositories

    International Nuclear Information System (INIS)

    Morgan, R.D.; Pryke, D.C.; Rees, J.H.

    1988-02-01

    The sorptive behaviour of the actinides uranium, neptunium, plutonium and americium has been investigated under air-saturated conditions on a number of candidate near-field materials by batch sorption experiments. Distribution ratios were measured with respect to initial actinide concentration, the solid:liquid ratio and contact time. Desorption experiments were carried out to help elucidate the mechanism of sorption. The fit of the data to the Freundlich isotherm was assessed. This work contains the data obtained in the investigation. (author)

  14. AFLOWLIB.ORG: a Distributed Materials Properties Repository from High-throughput Ab initio Calculations

    Science.gov (United States)

    2011-11-15

    9001, Beer -Sheva, 84190, Israel † Present address: Pacific Northwest National Laboratory, Richland, WA 99354. ‡ Present address: Department of Physics...919 660 8963 Abstract Empirical databases of crystal structures and thermodynamic properties are fundamental tools for materials research. Recent...contains over 150,000 thermodynamic entries for alloys, covering the entire composition range of more than 650 binary systems, 13,000 electronic

  15. BENTONITE-QUARTZ SAND AS THE BACKFILL MATERIALS ON THE RADIOACTIVE WASTE REPOSITORY

    Directory of Open Access Journals (Sweden)

    Raharjo Raharjo

    2010-06-01

    Full Text Available An investigation of the contribution of quartz sand in the bentonite mixture as the backfill materials on the shallow land burial of radioactive waste has been done. The experiment objective is to determine the effect of quartz sand in a bentonite mixture with bentonite particle sizes of -20+40, -40+60, and -60+80 mesh on the retardation factor and the uranium dispersion in the simulation of uranium migration in the backfill materials. The experiment was carried out by the fixed bed method in the column filled by the bentonite mixture with a bentonite-to-quartz sand weight percent ratio of 0/100, 25/75, 50/50, 75/25, and 100/0 on the water saturated condition flown by uranyl nitrate solution at concentration (Co of 500 ppm. The concentration of uranium in the effluents in interval 15 minutes represented as Ct was analyzed by spectrophotometer, then using Co and Ct, retardation factor (R and dispersivity ( were determined. The experiment data showed that the bentonite of -60+80 mesh and the quartz sand of -20+40 mesh on bentonite-to-quartz sand with weight percent ratio of 50/50 gave the highest retardation factor and dispersivity of 18.37 and 0.0363 cm, respectively.   Keywords: bentonite, quartz sand, backfill materials, radioactive waste

  16. Method of measuring material properties of rock in the wall of a borehole

    Science.gov (United States)

    Overmier, David K.

    1985-01-01

    To measure the modulus of elasticity of the rock in the wall of a borehole, a plug is cut in the borehole wall. The plug, its base attached to the surrounding rock, acts as a short column in response to applied forces. A loading piston is applied to the top of the plug and compression of the plug is measured as load is increased. Measurement of piston load and plug longitudinal deformation are made to determine the elastic modulus of the plug material. Poisson's ratio can be determined by simultaneous measurements of longitudinal and lateral deformation of the plug in response to loading. To determine shear modulus, the top of the plug is twisted while measurements are taken of torsional deformation.

  17. Geohydrological simulation of a deep coastal repository

    International Nuclear Information System (INIS)

    Follin, S.

    1995-12-01

    This conceptual-numerical study treats the dewatering and resaturation phases associated with the construction, use and closure of a coastal nuclear waste repository located at depth in sparsely fractured Baltic Shield rocks. The main objective is to simulate the extent and duration of saline intrusion for a reasonable set of geohydrological assumptions. Long-term changes in the chemical environment associated with saline intrusion may affect the properties of the buffer zone material (bentonite). The first part of the study deals with history matching of a simple model geometry and the second part treats the dewatering and resaturation phases of the simulated repository. The history matching supports the standpoint that the occurrence of saline ground water reflects an ongoing but incomplete Holocene flushing of the Baltic Shield. The drawdown after fifty years of dewatering is highly dependent on the permeability of the excavated damaged zone. If the permeability close the repository is unaltered the entire region between the top side of the model and the repository is more or less partially saturated at the end of the simulation period. The simulations of a fifty year long recovery period suggest that the distribution between fresh and saline ground waters may be quite close to the conditions prior to the dewatering phase already after fifty years of closure despite an incomplete pressure recovery, which is an interesting result considering the objective of the study. 12 refs

  18. Chemical and physical characteristics of phosphate rock materials of varying reactivity

    International Nuclear Information System (INIS)

    Syers, J.K.; Currie, L.D.

    1986-01-01

    Several chemical and physical properties of 10 phosphate rock (PR) materials of varying reactivity were evaluated. The highest concentrations of As and Cd were noted. Because Cd and U can accumulate in biological systems, it may be necessary to direct more attention towards the likely implications of Cd and U concentrations when evaluating a PR for direct application. Three sequential extractions with 2% citric acid may be more useful for comparing the chemical solubility of PR materials, particularly for those containing appreciable CaC0 3 . The poor relationship obtained between surface area and the solubility of the PR materials suggests that surface area plays a secondary role to chemical reactivity in controlling the solubility of a PR in a chemical extractant. A Promesh plot provided an effective method for describing the particle-size characteristics of those PR materials which occurred as sands. Fundamental characteristics, such as mean particle size and uniformity, can readily be determined from a Promesh plot. (author)

  19. Conceptual design of repository facilities

    International Nuclear Information System (INIS)

    Beale, H.; Engelmann, H.J.; Souquet, G.; Mayence, M.; Hamstra, J.

    1980-01-01

    As part of the European Economic Communities programme of research into underground disposal of radioactive wastes repository design studies have been carried out for application in salt deposits, argillaceous formations and crystalline rocks. In this paper the design aspects of repositories are reviewed and conceptual designs are presented in relation to the geological formations under consideration. Emphasis has been placed on the disposal of vitrified high level radioactive wastes although consideration has been given to other categories of radioactive waste

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  1. Effects of porosity on seismic velocities, elastic moduli and Poisson's ratios of solid materials and rocks

    Directory of Open Access Journals (Sweden)

    Chengbo Yu

    2016-02-01

    Full Text Available The generalized mixture rule (GMR is used to provide a unified framework for describing Young's (E, shear (G and bulk (K moduli, Lame parameter (λ, and P- and S-wave velocities (Vp and Vs as a function of porosity in various isotropic materials such as metals, ceramics and rocks. The characteristic J values of the GMR for E, G, K and λ of each material are systematically different and display consistent correlations with the Poisson's ratio of the nonporous material (ν0. For the materials dominated by corner-shaped pores, the fixed point at which the effective Poisson's ratio (ν remains constant is at ν0 = 0.2, and J(G > J(E > J(K > J(λ and J(G  0.2 and ν0  J(Vp and J(Vs  0.2 and ν0  0.2 and ν0 = 0.2, respectively. For natural rocks containing thin-disk-shaped pores parallel to mineral cleavages, grain boundaries and foliation, however, the ν fixed point decreases nonlinearly with decreasing pore aspect ratio (α: width/length. With increasing depth or pressure, cracks with smaller α values are progressively closed, making the ν fixed point rise and finally reach to the point at ν0 = 0.2.

  2. Sellafield repository design concept

    International Nuclear Information System (INIS)

    1998-01-01

    Between 1989 and 1997, UK Nirex Ltd carried out a programme of investigations to evaluate the potential of a site adjacent to the BNFL Sellafield works to host a deep repository for the United Kingdom's intermediate-level and certain low-level radioactive waste. The programme of investigations was wound down following the decision in March 1997 to uphold the rejection of the Company's planning application for the Rock Characterisation Facility (RCF), an underground laboratory which would have allowed further investigations to confirm whether or not the site would be suitable. Since that time, the Company's efforts in relation to the Sellafield site have been directed towards documenting and publishing the work carried out. The design concept for a repository at Sellafield was developed in parallel with the site investigations through an iterative process as knowledge of the site and understanding of the repository system performance increased. This report documents the Sellafield repository design concept as it had been developed, from initial design considerations in 1991 up to the point when the RCF planning application was rejected. It shows, from the context of a project at that particular site, how much information and experience has been gained that will be applicable to the development of a deep waste repository at other potential sites

  3. Geochemical homogeneity of tuffs at the potential repository level, Yucca Mountain, Nevada

    International Nuclear Information System (INIS)

    Peterman, Zell E.; Cloke, Paul

    2001-01-01

    In a potential high-level radioactive waste repository at Yucca Mountain, Nevada, radioactive waste and canisters, drip shields protecting the waste from seepage and from rock falls, the backfill and invert material of crushed rock, the host rock, and water and gases contained within pores and fractures in the host rock together would form a complex system commonly referred to as the near-field geochemical environment. Materials introduced into the rock mass with the waste that are designed to prolong containment collectively are referred to as the Engineered Barrier System, and the host rock and its contained water and gases compose the natural system. The interaction of these component parts under highly perturbed conditions including temperatures well above natural ambient temperatures will need to be understood to assess the performance of the potential repository for long-term containment of nuclear waste. The geochemistry and mineralogy of the rock mass hosting the emplacement drifts must be known in order to assess the role of the natural system in the near-field environment. Emplacement drifts in a potential repository at Yucca Mountain would be constructed in the phenocryst-poor member of the Topopah Spring Tuff which is composed of both lithophysal and nonlithophysal zones. The chemical composition of the phenocryst-poor member has been characterized by numerous chemical analyses of outcrop samples and of core samples obtained by surface-based drilling. Those analyses have shown that the phenocryst-poor member of the Topopah Spring Tuff is remarkably uniform in composition both vertically and laterally. To verify this geochemical uniformity and to provide rock analyses of samples obtained directly from the potential repository block, major and trace elements were analyzed in core samples obtained from drill holes in the cross drift, which was driven to provide direct access to the rock mass where emplacement drifts would be constructed

  4. Industrial Application of Valuable Materials Generated from PLK Rock-A Bauxite Mining Waste

    Science.gov (United States)

    Swain, Ranjita; Routray, Sunita; Mohapatra, Abhisek; Ranjan Patra, Biswa

    2018-03-01

    PLK rock classified in to two products after a selective grinding to a particular size fraction. PLK rocks ground to below 45-micron size which is followed by a classifier i.e. hydrocyclone. The ground product classified in to different sizes of apex and vortex finder. The pressure gauge was attached for the measurement of the pressure. The production of fines is also increasing with increase in the vortex finder diameter. In order to increase in the feed capacity of the hydrocyclone, the vortex finder 11.1 mm diameter and the spigot diameter 8.0 mm has been considered as the best optimum condition for recovery of fines from PLK rock sample. The overflow sample contains 5.39% iron oxide (Fe2O3) with 0.97% of TiO2 and underflow sample contains 1.87% Fe2O3 with 2.39% of TiO2. The cut point or separation size of overflow sample is 25 μm. The efficiency of separation, or the so-called imperfection I, is at 6 μm size. In this study, the iron oxide content in underflow sample is less than 2% which is suitable for making of refractory application. The overflow sample is very fine which can also be a raw material for ceramic industry as well as a cosmetic product.

  5. A novel design for storage of inner stress by colloidal processing on rock-like materials

    Science.gov (United States)

    Chen, Weichang; Wang, Sijing; Lekan Olatayo, Afolagboye; Fu, Huanran

    2018-06-01

    Inner stress exists in rocks, affecting rock engineering, yet has received very little attention and quantitative investigation because of uncertainty about its characteristics. Previous studies have suggested that the inner stresses of rock materials are closely related to their physical state variation. In this work, a novel mold was designed to simulate the storage process of inner stress in specimens composed of quartz sands and epoxy. Then, thermal tests were carried out to change the physical state of the specimens, and expansion of the specimens was monitored. The results indicated that inner stress could be partly locked by the mold and it could also be released by heating. It can be inferred from the analysis that one necessary condition of storage and release of inner stress is physical state variation. Additionally, by using an XRD method, the variations in the interplanar spacing of the quartz sands were detected, and the results reflect that inner stress could be locked-in aggregates (quartz sands) by a cement constraint (solid epoxy). The inner stress stored in quartz sands was calculated using height and interplanar spacing variations.

  6. Study on radionuclide migration through a buffer material of the repository for high level nuclear waste

    International Nuclear Information System (INIS)

    Tsukamoto, Masaki; Ohe, Toshiaki

    1989-01-01

    The present report discusses radionuclide migration through a buffer material from the view point of experimental and data analysis. Na-bentonite loosely compacted with dry density of 0.8 - 1 g/cm 3 was contacted with cesium chloride solution of about 100 ppb containing Cs-134 as a tracer at 40degC and at 70degC. After the experiments, the bentonite cake was sliced and cesium distribution in the cake was measured by gamma-spectrometry. Apparent diffusivities of 2∼5 x 10 -7 cm 2 /sec was determined through the analysis method where pore diffusion and adsorption were involved. Numerical solution well described the observed data. The pore diffusion would be clearfied to be a dominant mechanism of the radionuclide migration in the bentonite, through discussing the pore diffusion mechanism and activation energy of the diffusion. This report also discusses the capability of the chemical transport model CHEMTRN for long-term predictions of the radionuclide migration. (author)

  7. Aespoe Hard Rock Laboratory. Annual report 1997

    International Nuclear Information System (INIS)

    1998-05-01

    The Aespoe Hard Rock Laboratory has been constructed as part of the preparations for the deep geological repository for spent nuclear fuel in Sweden. The surface and borehole investigations and the research work performed in parallel with construction have provided a thorough test of methods for investigation and evaluation of bedrock conditions for construction of a deep repository. The Tracer Retention Understanding Experiments are made to gain a better understanding of radionuclide retention in the rock and create confidence in the radionuclide transport models that are intended to be used in the licensing of a deep repository for spent fuel. The experimental results of the first tracer test with sorbing radioactive tracers have been obtained. These tests have been subject to blind predictions by the Aespoe Task Force on groundwater flow and transports of solutes. The manufacturing of the CHEMLAB probe was completed during 1996, and the first experiments were started early in 1997. During 1997 three experiments on diffusion in bentonite using 57 Co, 114 Cs, 85 Sr, 99 Tc, and 131 I were conducted. The Prototype Repository Test is focused on testing and demonstrating repository system function. A full scale prototype including six deposition holes with canisters with electric heaters surrounded by highly compacted bentonite will be built and instrumented. The characterization of the rock mass in the area of the prototype repository is in progress. The objectives of the Demonstration of Repository Technology are to develop, test, and demonstrate methodology and equipment for encapsulation and deposition of spent nuclear fuel. The demonstration of handling and deposition will be made in a new drift. The Backfill and Plug Test includes tests of backfill materials and emplacement methods and a test of a full scale plug. The backfill and rock will be instrumented with about 230 transducers for measuring the thermo-hydro-mechanical processes. The Retrieval Test is

  8. Aespoe Hard Rock Laboratory. Annual report 1997

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-05-01

    The Aespoe Hard Rock Laboratory has been constructed as part of the preparations for the deep geological repository for spent nuclear fuel in Sweden. The surface and borehole investigations and the research work performed in parallel with construction have provided a thorough test of methods for investigation and evaluation of bedrock conditions for construction of a deep repository. The Tracer Retention Understanding Experiments are made to gain a better understanding of radionuclide retention in the rock and create confidence in the radionuclide transport models that are intended to be used in the licensing of a deep repository for spent fuel. The experimental results of the first tracer test with sorbing radioactive tracers have been obtained. These tests have been subject to blind predictions by the Aespoe Task Force on groundwater flow and transports of solutes. The manufacturing of the CHEMLAB probe was completed during 1996, and the first experiments were started early in 1997. During 1997 three experiments on diffusion in bentonite using {sup 57}Co, {sup 114}Cs,{sup 85}Sr, {sup 99}Tc, and {sup 131}I were conducted. The Prototype Repository Test is focused on testing and demonstrating repository system function. A full scale prototype including six deposition holes with canisters with electric heaters surrounded by highly compacted bentonite will be built and instrumented. The characterization of the rock mass in the area of the prototype repository is in progress. The objectives of the Demonstration of Repository Technology are to develop, test, and demonstrate methodology and equipment for encapsulation and deposition of spent nuclear fuel. The demonstration of handling and deposition will be made in a new drift. The Backfill and Plug Test includes tests of backfill materials and emplacement methods and a test of a full scale plug. The backfill and rock will be instrumented with about 230 transducers for measuring the thermo-hydro-mechanical processes. The

  9. Diffusivity database (DDB) system for major rocks and buffer materials (Released on 2007/specification)

    International Nuclear Information System (INIS)

    Tochigi, Yoshikatsu; Shibata, Masahiro; Sato, Haruo; Kitamura, Akira

    2007-03-01

    The Diffusivity Database (DDB) System developed on early 2006 was upgraded to apply the data of effective diffusion coefficient of the nuclides in the rock matrix for the 'H12: Project to Establish the Scientific and Technical Basis for HLW Disposal in Japan', and the data in the buffer materials from literature survey was newly added. Some functions of data search and selection were reformed to improve the level of convenience. This DDB system (work on MS-Access TM ) is released to the public through Web server managed by JAEA. (author)

  10. Project Guarantee 1985. Repository for low- and intermediate-level radioactive waste: construction and operation

    International Nuclear Information System (INIS)

    Anon.

    1985-01-01

    A constructional engineering project study aimed at clarification of the feasibility of a repository for low- and intermediate-level radioactive waste (type B repository) has been carried out; the study is based on a model data-set derived from the geological, rock mechanical and topographical characterictics of one of Nagra's planned exploration areas. Final storage is effected in subterranean rock caverns accessed by horizontal tunnel. The reception area also is sited below the surface. Storage is conceived in such a way that, after closure of the repository, maintenance and supervision can be dispensed with and a guarantee of high long-term safety can nevertheless be provided. The envisaged repository consists of an entry tunnel for road vehicles and a reception area with a series of caverns for receiving waste, for additional technical facilities and for the production of the concrete back-fill material. The connecting tunnel is serviced by a tunnel railway and the actual repository area consists of several storage caverns. The repository is intended to accomodate a total of 200'000 m3 of solidified low- and intermediate-level waste. Valanginian marl is assumed as the host rock, although it would also be basically possible to house the proposed installations in other host rocks. The excavated material will total around 1'000'000 m3. The construction time for the whole installation is estimated as about 7 years and a working team of around 30 people will be required for the estimated 60-year operational duration. The project described in the present report justifies the conclusion that construction of a repository for low-and intermediate-level radioactive waste is feasible with present-day technology. This conclusion takes into consideration quantitative and operational constraints as well as geological and hydrogeological data relevant to constructional engineering. The latter are derived from a model data-set based on a specific locality

  11. Site selection for deep geologic repositories - Consequences for society, economy and environment

    International Nuclear Information System (INIS)

    2010-03-01

    In a few years, Switzerland will make the decision regarding site selection for geological underground repositories for the storage of radioactive wastes. Besides the safety issue, many citizens are interested in how such a repository will affect environment, economy and society in the selected site's region. This brochure summarizes the results of many studies on the socio-economic impacts of nuclear waste repositories. Radioactive wastes must be stored in such a way that mankind and environment are safely protected for a long period of time. How this goal may be achieved, is already known: geologic deep repositories warrant long-term safety. For the oncoming years in Switzerland the question is where the repository will be built. The search for an appropriate site for a repository in the proposed regions will launch discussions. Within the participative framework the regions may bring their requests. The demonstration of the safety of potential repository sites has the highest priority in the selection process. In the third procedural step additional rock investigations will be made. The socio-economic studies and the experience with existing plants show that radioactive waste management plants can be built and operated in good agreement with environmental requirements. The radioactive wastes in a deep underground repository are stored many hundred meters below the Earth's surface. There, they are isolated from our vital space. Technical barriers and the surrounding dense rock confinement prevent the release of radioactive materials into the environment. A deep repository has positive consequences for the regional economy. It increases trade and value creation and creates work places. The socio-economic impacts practically extend over one century, but strongly vary with time; they are the largest during the building period. High life quality and a positive population development in the selected site region are compatible with a deep repository. A fair and

  12. Cyclic and Fatigue Behaviour of Rock Materials: Review, Interpretation and Research Perspectives

    Science.gov (United States)

    Cerfontaine, B.; Collin, F.

    2018-02-01

    The purpose of this paper is to provide a comprehensive state of the art of fatigue and cyclic loading of natural rock materials. Papers published in the literature are classified and listed in order to ease bibliographical review, to gather data (sometimes contradictory) on classical experimental results and to analyse the main interpretation concepts. Their advantages and limitations are discussed, and perspectives for further work are highlighted. The first section summarises and defines the different experimental set-ups (type of loading, type of experiment) already applied to cyclic/fatigue investigation of rock materials. The papers are then listed based on these different definitions. Typical results are highlighted in next section. Fatigue/cyclic loading mainly results in accumulation of plastic deformation and/or damage cycle after cycle. A sample cyclically loaded at constant amplitude finally leads to failure even if the peak load is lower than its monotonic strength. This subcritical crack is due to a diffuse microfracturing and decohesion of the rock structure. The third section reviews and comments the concepts used to interpret the results. The fatigue limit and S- N curves are the most common concepts used to describe fatigue experiments. Results published from all papers are gathered into a single figure to highlight the tendency. Predicting the monotonic peak strength of a sample is found to be critical in order to compute accurate S- N curves. Finally, open questions are listed to provide a state of the art of grey areas in the understanding of fatigue mechanisms and challenges for the future.

  13. Hydrothermal conditions around a radioactive waste repository

    International Nuclear Information System (INIS)

    Thunvik, R.; Braester, C.

    1981-12-01

    Numerical solutions for the hydrothermal conditions around a hard rock repository for nuclear fuel waste are presented. The objective of the present investigation is to illustrate in principle the effect of heat released from a hypothetical radioactive waste repository with regard to anisotropy in the rock permeability. Permeability and porosity are assumed to be constant or to decrease exponentially with depth. The hypothetical repository is situated below a horizontal ground surface or below the crest of a hill, and it is assumed that the water table follows the topography. Major interest in the analysis is directed towards the influence of anisotropy in the permeability on the flow patterns and travel times for water particles, being traced from the repository to the ground surface. The presented results show that anisotropy in the permeability may have a significant influence on the flow conditions around the repository and subsequently also on the travel times from the repository. (Authors)

  14. Experimental Field Tests and Finite Element Analyses for Rock Cracking Using the Expansion of Vermiculite Materials

    Directory of Open Access Journals (Sweden)

    Chi-hyung Ahn

    2016-01-01

    Full Text Available In the previous research, laboratory tests were performed in order to measure the expansion of vermiculite upon heating and to convert it into expansion pressure. Based on these test results, this study mainly focuses on experimental field tests conducted to verify that expansion pressure obtained by heating vermiculite materials is enough to break massive and hard granite rock with an intention to excavate the tunnel. Hexahedral granite specimens with a circular hole perforated in the center were constructed for the experimental tests. The circular holes were filled with vermiculite plus thermal conduction and then heated using the cartridge heater. As a result, all of hexahedral granite specimens had cracks in the surface after 700-second thermal heating and were finally spilt into two pieces completely. The specimen of larger size only requires more heating time and expansion pressure. The material properties of granite rocks, which were obtained from the experimental tests, were utilized to produce finite element models used for numerical analyses. The analysis results show good agreement with the experimental results in terms of initial cracking, propagation direction, and expansion pressure.

  15. Safeguards for geological repositories

    International Nuclear Information System (INIS)

    Fattah, A.

    2000-01-01

    Direct disposal of spent nuclear fuel in geological repositories is a recognised option for closing nuclear fuel cycles. Geological repositories are at present in stages of development in a number of countries and are expected to be built and operated early next century. A State usually has an obligation to safely store any nuclear material, which is considered unsuitable to re-enter the nuclear fuel cycle, isolated from the biosphere. In conjunction with this, physical protection has to be accounted for to prevent inadvertent access to such material. In addition to these two criteria - which are fully under the State's jurisdiction - a third criterion reflecting international non-proliferation commitments needs to be addressed. Under comprehensive safeguards agreements a State concedes verification of nuclear material for safeguards purposes to the IAEA. The Agency can thus provide assurance to the international community that such nuclear material has been used for peaceful purposes only as declared by the State. It must be emphasised that all three criteria mentioned constitute a 'unit'. None can be sacrificed for the sake of the other, but compromises may have to be sought in order to make their combination as effective as possible. Based on comprehensive safeguards agreements signed and ratified by the State, safeguards can be terminated only when the material has been consumed or diluted in such a way that it can no longer be utilised for any nuclear activities or has become practicably irrecoverable. As such safeguards for nuclear material in geological repositories have to be continued even after the repository has been back-filled and sealed. The effective application of safeguards must assure continuity-of-knowledge that the nuclear material in the repository has not been diverted for an unknown purpose. The nuclear material disposed in a geological repository may eventually have a higher and long term proliferation risk because the inventory is

  16. Flow behavior and mobility of contaminated waste rock materials in the abandoned Imgi mine in Korea

    Science.gov (United States)

    Jeong, S. W.; Wu, Y.-H.; Cho, Y. C.; Ji, S. W.

    2018-01-01

    Incomplete mine reclamation can cause ecological and environmental impacts. This paper focuses on the geotechnical and rheological characteristics of waste rock materials, which are mainly composed of sand-size particles, potentially resulting in mass movement (e.g., slide or flow) and extensive acid mine drainage. To examine the potential for contaminant mobilization resulting from physicochemical processes in abandoned mines, a series of scenario-based debris flow simulations was conducted using Debris-2D to identify different hazard scenarios and volumes. The flow behavior of waste rock materials was examined using a ball-measuring rheometric apparatus, which can be adapted for large particle samples, such as debris flow. Bingham yield stresses determined in controlled shear rate mode were used as an input parameter in the debris flow modeling. The yield stresses ranged from 100 to 1000 Pa for shear rates ranging from 10- 5 to 102 s- 1. The results demonstrated that the lowest yield stress could result in high mobility of debris flow (e.g., runout distance > 700 m from the source area for 60 s); consequently, the material contaminants may easily reach the confluence of the Suyoung River through a mountain stream. When a fast slide or debris flow occurs at or near an abandoned mine area, it may result in extremely dynamic and destructive geomorphological changes. Even for the highest yield stress of debris flow simulation (i.e., τy = 2000 Pa), the released debris could flow into the mountain stream; therefore, people living near abandoned mines may become exposed to water pollution throughout the day. To maintain safety at and near abandoned mines, the physicochemical properties of waste materials should be monitored, and proper mitigation measures post-mining should be considered in terms of both their physical damage and chemical pollution potential.

  17. Repository simulation tests

    International Nuclear Information System (INIS)

    Wicks, G.G.; Bibler, N.E.; Jantzen, C.M.; Plodinec, M.J.

    1984-01-01

    The repository simulation experiments described in this paper are designed to assess the performance of SRP waste glass under the most realistic repository conditions that can be obtained in the laboratory. These tests simulate the repository environment as closely as possible and introduce systematically the variability of the geology, groundwater chemistry, and waste package components during the leaching of the waste glass. The tests evaluate waste form performance under site-specific conditions, which differ for each of the geologic repositories under consideration. Data from these experiments will aid in the development of a realistic source term that can describe the release of radionuclides from SRP waste glass as a component of proposed waste packages. Hence, this information can be useful to optimize waste package design for SRP waste glass and to provide data for predicting long-term performance and subsequent conformance to regulations. The repository simulation tests also help to bridge the gap in interpreting results derived from tests performed under the control of the laboratory to the uncertainity and variability of field tests. In these experiments, site-specific repository components and conditions are emphasized and only the site specific materials contact the waste forms. An important feature of these tests is that both actual and simulated waste glasses are tested identically. 7 figures, 2 tables

  18. Numerical modeling of magma-repository interactions

    NARCIS (Netherlands)

    Bokhove, Onno

    2001-01-01

    This report explains the numerical programs behind a comprehensive modeling effort of magma-repository interactions. Magma-repository interactions occur when a magma dike with high-volatile content magma ascends through surrounding rock and encounters a tunnel or drift filled with either a magmatic

  19. Heterogeneous redox conditions, arsenic mobility, and groundwater flow in a fractured-rock aquifer near a waste repository site in New Hampshire, USA

    Science.gov (United States)

    Anthropogenic sources of carbon from landfill or waste leachate can promote reductive dissolution of in situ arsenic (As) and enhance the mobility of As in groundwater. Groundwater from residential-supply wells in a fractured crystalline-rock aquifer adjacent to a Superfund site ...

  20. Sample management implementation plan: Salt Repository Project

    International Nuclear Information System (INIS)

    1987-01-01

    The purpose of the Sample Management Implementation Plan is to define management controls and building requirements for handling materials collected during the site characterization of the Deaf Smith County, Texas, site. This work will be conducted for the US Department of Energy Salt Repository Project Office (SRPO). The plan provides for controls mandated by the US Nuclear Regulatory Commission and the US Environmental Protection Agency. Salt Repository Project (SRP) Sample Management will interface with program participants who request, collect, and test samples. SRP Sample Management will be responsible for the following: (1) preparing samples; (2) ensuring documentation control; (3) providing for uniform forms, labels, data formats, and transportation and storage requirements; and (4) identifying sample specifications to ensure sample quality. The SRP Sample Management Facility will be operated under a set of procedures that will impact numerous program participants. Requesters of samples will be responsible for definition of requirements in advance of collection. Sample requests for field activities will be approved by the SRPO, aided by an advisory group, the SRP Sample Allocation Committee. This document details the staffing, building, storage, and transportation requirements for establishing an SRP Sample Management Facility. Materials to be managed in the facility include rock core and rock discontinuities, soils, fluids, biota, air particulates, cultural artifacts, and crop and food stuffs. 39 refs., 3 figs., 11 tabs

  1. Effects of water inflow and early water uptake on buffer and backfill materials in a KBS-3V repository

    International Nuclear Information System (INIS)

    Boergesson, L.; Sanden, T.; Dueck, A.; Nilsson, U.; Goudarzi, R.; Andersson, L.; Jensen, V.

    2012-01-01

    Document available in extended abstract form only. Bentonite is an excellent sealing material when it has reached full water saturation and swelling pressure. However, bentonite is not good for sealing inflowing water from fractures with potential to build high water pressure. It cannot stop inflow of water at the depth of a repository. The water inflow into the pellets filled slots in the deposition holes and the tunnels in a KBS-3V repository is expected to continue until these slots are water filled and the water flow stopped by an end plug. Then the water pressure gradient is transferred from the fracture/bentonite interface to the plug and the bentonite will have time to homogenize and seal. This scenario leads to a number of processes that can either be harmful to the bentonite or affect the water saturation and homogenization evolution. Last year a project (EVA) started in order to investigate the processes involved by this early water inflow. The project aims at developing a model for the processes piping, erosion, water filling of pellets filled slots, early water absorption and resulting water pressure increase against the plug. The project studies the effects of water inflow in deposition holes and deposition tunnels and the emergence of piping and erosion during installation and wetting of the buffer and backfill until all slots and the pellet fillings have been water filled and piping and erosion have ceased. The project includes laboratory tests of nine different processes and modeling. The laboratory program includes tests of the following processes: 1. Erosion; 2. Piping; 3. Water flow in pellet filled slots; 4. Sealing ability of bentonite; 5. Water absorption of the bentonite blocks; 6. Formation of water or gel pockets in a pellet filled slot; 7. Formation and outflow of bentonite gel; 8. Self-sealing of cracks by eroding water; 9. Buffer swelling before placement of backfill. The laboratory tests are ongoing and preliminary results and

  2. Source term estimation and the isotopic ratio of radioactive material released from the WIPP repository in New Mexico, USA

    International Nuclear Information System (INIS)

    Thakur, P.

    2016-01-01

    lifted its lid to allow release of waste into the underground air. - Highlights: • February 14, 2014 radiation release from the Nation's only TRU waste repository is discussed. • Environmental monitoring results in the vicinity of the WIPP site are presented. • Source term estimation of the radioactive material released is assessed. • The isotopic ratio of the radioactive material release from the WIPP is discussed.

  3. Sealing properties of cement-based grout materials. Final report on the Rock sealing project

    International Nuclear Information System (INIS)

    Onofrei, M.; Gray, Malcolm; Shenton, B.; Walker, Brad; Pusch, R.; Boergesson, L.; Karnland, O.

    1992-10-01

    This report presents the results of laboratory studies of material properties. A number of different high performance grouts were investigated. The laboratory studies focused on mixtures of sulphate resistant portland cement, silica fume, superplasticizer and water. The ability of the thin films to self seal was confirmed. The surface reactions were studied in specimens of hardened grouts. The leach rates were found to vary with grout and water composition and with temperature. The short-term hydraulic and strength or properties of the hardened grout were determined. These properties were determined for the grouts both in-bulk and as thin-films. The hydraulic conductivities of the bulk, hardened material were found to be less than 10 -14 m/s. The hydraulic conductivities of thin films were found to be less than 10 -11 m/s. Broken, the hydraulic conductivity of the thin films could be increased to 10 -7 m/s. Examination of the leached grout specimens revealed a trend for the pore sizes to decrease with time. The propensity for fractured grouts to self seal was also observed in tests in which the hydraulic conductivity of recompacted mechanically disrupted, granulated grouts was determined. These tests showed that the hydraulic conductivity decreased rapidly with time. The decreases were associated with decreases in mean pore size. In view of the very low hydraulic conductivity it is likely that surface leaching at the grout/groundwater interface will be that major process by which bulk high-performance grouts may degrade. With the completion of the laboratory, in situ and modelling studies it appears that high-performance cement based grouts can be considered as viable materials for some repository sealing applications. Some of the uncertainties that remain are identified in this report. (54 refs.)

  4. Radionuclides in hydrothermal systems as indicators of repository conditions

    International Nuclear Information System (INIS)

    Wollenberg, H.A.; Flexser, S.; Smith, A.R.

    1990-11-01

    Hydrothermal systems in tuffaceous and older sedimentary rocks contain evidence of the interaction of radionuclides in fluids with rock matrix minerals and with materials lining fractures, in settings somewhat analogous to the candidate repository site at Yucca Mountain, NV. Earlier studies encompassed the occurrences of U and Th in a ''fossil'' hydrothermal system in tuffaceous rock of the San Juan Mountains volcanic field, CO. More recent and ongoing studies examine active hydrothermal systems in calderas at Long Valley, CA and Valles, NM. At the Nevada Test Site, occurrences of U and Th in fractured and unfractured rhyolitic tuff that was heated to simulate the introduction of radioactive waste are also under investigation. Observations to date suggest that U is mobile in hydrothermal systems, but that localized reducing environments provided by Fe-rich minerals and/or carbonaceous material concentrate U and thus attenuate its migration. 11 refs., 6 figs., 1 tab

  5. Modelling of Fracture Initiation, Propagation and Creep of a KBS-3V and KBS-3H Repository in Sparsely Fractured Rock with Application to the Design at Forsmark Candidate Site

    International Nuclear Information System (INIS)

    Backers, Tobias; Stephansson, Ove

    2008-01-01

    The stability issues of deposition holes of a repository layout according to the KBS-3 concept in the sparsely fractured Forsmark granites are analysed with the emphasis on fracture mechanics. At the start of the project the rock mass is viewed as a continuum. In a second step explicit fracture networks are introduced and included in the numerical rock fracture models. The software Fracod2D was used for the rock fracture mechanics analysis. Assuming deposition holes located in a continuous, homogeneous elastic rock mass and The presented stress state of the rock mass the following results were obtained: For single KBS-3H deposition holes oriented in the direction of the minimum horizontal stress, Sh, bore hole breakouts are introduced for all depth levels. For KBS-3H holes which are oriented in direction of SH, no significant fracturing can be expected. In case of vertical deposition holes according to KBS-3V an increased risk of fracturing at greater depth levels (> 500m) is evident. At shallow depth levels ( 5MPa gives a favourable situation about spalling for the KBS-3H and KBS-3V layouts. To prevent spalling, it is important to build up a swelling pressure soon after excavation, so that the enhanced stresses in the surrounding of the deposition ii holes are reduced. This has a positive impact on other excavation activities and also on time-dependent fracturing. After excavation and filling of the deposition holes with subsequent increase of swelling pressure, the temperature will increase in the vicinity of the excavation. For the range of swelling pressures predicted for the KBS-3 concept, i.e. 5.5MPa to 7.2MPa, no significant fracturing for the KBS-3H concept with the axis parallel SH at depths below about 600m was discovered. The results from other layouts bare the risk of partly significant fracturing. About 60ka from closing the repository an ice cover of approximately 3km is expected over Forsmark. This dead load increases the in-situ stresses and

  6. Barriers to migration of radionuclides from radioactive waste repositories

    International Nuclear Information System (INIS)

    Stefanova, I.

    1999-01-01

    Natural inorganic sorbents are known as effective barriers that reduce the migration of radionuclides from radioactive waste repositories and contaminated sites. They could be used as buffer, backfill and sealing materials in the repository and their presence in the host rock and the surrounding geological formations increases the retention properties of the strata. Natural occurring minerals from local origin are used in the study - zeolites (clinoptilolite and mordenite), celadonite and loess. Sorption of wide range of radionuclides is studies. Batch capacity is determined. Sorption of radionuclides from simulated natural solution is studied. Distribution coefficients are calculated from sorption isotherms. Desorption in presence of different natural solutions is studied. Sorption properties are compared. It is shown that clinoptilolite acts as effective barrier against migration of radionuclides from repositories. The presence of celadonite in the clinoptilolite rock increases the retention of polyvalent ions. The retention of radionuclides on loess samples fulfils the requirements for host media for repository for low and intermediate level waste. A method for construction of additional barrier to the existing in the country disposal vault for spent sealed sources is proposed

  7. The Getafe rock: Fall, composition and cosmic ray records of an unusual ultrarefractory scoriaceous material

    International Nuclear Information System (INIS)

    Martinez-Frias, J.; Weigel, A.; Marti, K.; Boyd, T.; Wilson, G. H.; Jull, T.

    1999-01-01

    In 1994 a moving car and its driver, on a highway in southern Madrid (Getafe) were struck by a falling rock. Eighty-one additional fragments (total weight: 55.926 kg) were later recovered, which all pointed towards a meteorite fall. A study of the composition of this object revealed an ultrarefractory material displaying a most unusual chemical make-up which differs from any known meteorite class, and for some elements and minerals approaches the composition of CAI (Ca-Al. rich inclusions in chondrites). A study of some cosmic-ray-produced stable and radioactive nuclides indicates: a) space and terrestrial exposure ages which do not exceed 1,000 and 520,000 years, respectively; b) the presence of a small ''227 Ne excess (1,100 deg C fraction), which suggest either a nucleogenic contribution from the ''19 F (α, n) ''22Ne reaction or a trapped Ne signature distinct from atmospheric Ne, and c) the existence of minor variations in the ''38Ar/Ar ratios also indicating a nucleogenic component or fractionation effects ''14C data are consistent with modern carbon originated in the period 1955-1958 and not earlier or more recently. The possibility that the Getafe rock could have a man-made origin (i.e. ceramic and refractory tiles, industrial slag) is also considered. (Author) 29 refs

  8. Selection of candidate container materials for the conceptual waste package design for a potential high level nuclear waste repository at Yucca Mountain

    Energy Technology Data Exchange (ETDEWEB)

    Van Konynenburg, R.A.; Halsey, W.G.; McCright, R.D.; Clarke, W.L. Jr. [Lawrence Livermore National Lab., CA (United States); Gdowski, G.E. [KMI, Inc., Albuquerque, NM (United States)

    1993-02-01

    Preliminary selection criteria have been developed, peer-reviewed, and applied to a field of 41 candidate materials to choose three alloys for further consideration during the advanced conceptual design phase of waste package development for a potential high level nuclear waste repository at Yucca Mountain, Nevada. These three alloys are titanium grade 12, Alloy C-4, and Alloy 825. These selections are specific to the particular conceptual design outlined in the Site Characterization Plan. Other design concepts that may be considered in the advanced conceptual design phase may favor other materials choices.

  9. Selection of candidate container materials for the conceptual waste package design for a potential high level nuclear waste repository at Yucca Mountain

    International Nuclear Information System (INIS)

    Van Konynenburg, R.A.; Halsey, W.G.; McCright, R.D.; Clarke, W.L. Jr.; Gdowski, G.E.

    1993-02-01

    Preliminary selection criteria have been developed, peer-reviewed, and applied to a field of 41 candidate materials to choose three alloys for further consideration during the advanced conceptual design phase of waste package development for a potential high level nuclear waste repository at Yucca Mountain, Nevada. These three alloys are titanium grade 12, Alloy C-4, and Alloy 825. These selections are specific to the particular conceptual design outlined in the Site Characterization Plan. Other design concepts that may be considered in the advanced conceptual design phase may favor other materials choices

  10. Low- and intermediate-level waste repository-induced effects

    Energy Technology Data Exchange (ETDEWEB)

    Leupin, O.X.; Marschall, P.; Johnson, L.; Cloet, V.; Schneider, J. [National Cooperative for the Disposal of Radioactive Waste (NAGRA), Wettingen (Switzerland); Smith, P. [Safety Assessment Management Ltd, Henley-On-Thames, Oxfordshire (United Kingdom); Savage, D. [Savage Earth Associates Ltd, Bournemouth, Dorset (United Kingdom); Senger, R. [Intera Inc., Ennetbaden (Switzerland)

    2016-10-15

    This status report aims at describing and assessing the interactions of the radioactive waste emplaced in a low- and intermediate level waste (L/ILW) repository with the engineered materials and the Opalinus Clay host rock. The Opalinus Clay has a thickness of about 100 m in the proposed siting regions. Among other things the results are used to steer the RD and D programme of NAGRA. The repository-induced effects considered in this report are of the following broad types: - Thermal effects: i.e. effects arising principally from the heat generated by the waste and the setting of cement. - Rock-mechanical effects: i.e. effects arising from the mechanical disturbance to the rock caused by the excavation of the emplacement caverns and other underground structures. - Hydraulic and gas-related effects: i.e. the effects of repository resaturation and of gas generation, e.g. due to the corrosion of metals within the repository, on the host rock and engineered barriers. - Chemical effects: i.e. chemical interactions between the waste, the engineered materials and the host rock. Deep geological repositories are designed to avoid or mitigate the impact of potentially detrimental repository-induced effects on long-term safety. For the repository under consideration in the present report, an assessment of those repository-induced effects that remain shows that detrimental chemical and mechanical impacts are largely confined to the rock adjacent to the excavations, thermal impacts are minimal and gas effects can be mitigated by appropriate design measures to reduce gas production and provide pathways for gas transport that limit gas pressure build-up (engineered gas transport system, or EGTS). Specific measures that are part of the current reference design are discussed in relation to their significance with respect to repository-induced effects. The disposal system described in this report provides a system of passive barriers with multiple safety functions. The disposal

  11. Low- and intermediate-level waste repository-induced effects

    International Nuclear Information System (INIS)

    Leupin, O.X.; Marschall, P.; Johnson, L.; Cloet, V.; Schneider, J.; Smith, P.; Savage, D.; Senger, R.

    2016-10-01

    This status report aims at describing and assessing the interactions of the radioactive waste emplaced in a low- and intermediate level waste (L/ILW) repository with the engineered materials and the Opalinus Clay host rock. The Opalinus Clay has a thickness of about 100 m in the proposed siting regions. Among other things the results are used to steer the RD and D programme of NAGRA. The repository-induced effects considered in this report are of the following broad types: - Thermal effects: i.e. effects arising principally from the heat generated by the waste and the setting of cement. - Rock-mechanical effects: i.e. effects arising from the mechanical disturbance to the rock caused by the excavation of the emplacement caverns and other underground structures. - Hydraulic and gas-related effects: i.e. the effects of repository resaturation and of gas generation, e.g. due to the corrosion of metals within the repository, on the host rock and engineered barriers. - Chemical effects: i.e. chemical interactions between the waste, the engineered materials and the host rock. Deep geological repositories are designed to avoid or mitigate the impact of potentially detrimental repository-induced effects on long-term safety. For the repository under consideration in the present report, an assessment of those repository-induced effects that remain shows that detrimental chemical and mechanical impacts are largely confined to the rock adjacent to the excavations, thermal impacts are minimal and gas effects can be mitigated by appropriate design measures to reduce gas production and provide pathways for gas transport that limit gas pressure build-up (engineered gas transport system, or EGTS). Specific measures that are part of the current reference design are discussed in relation to their significance with respect to repository-induced effects. The disposal system described in this report provides a system of passive barriers with multiple safety functions. The disposal

  12. 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

  13. Thermal Management and Analysis for a Potential Yucca Mountain Repository

    International Nuclear Information System (INIS)

    Dr. A. Van Luik

    2004-01-01

    In the current Yucca Mountain repository design concept, heat from the emplaced waste (mostly from spent nuclear fuel) would keep the temperature of the rock around the waste packages higher than the boiling point of water for hundreds to thousands of years after the repository is closed. The design concept allows below-boiling portions of the pillars between drifts to serve as pathways for the drainage of thermally mobilized water and percolating groundwater by limiting the distance that boiling temperatures extend into the surrounding rock. This design concept takes advantage of host rock dry out, which would create a dry environment within the emplacement drifts and reduce the amount of water that might otherwise be available to enter the drifts and contact the waste packages during this thermal pulse. Table 1 provides an overview of design constraints related to thermal management after repository closure. The Yucca Mountain repository design concept also provides flexibility to allow for operation over a range of lower thermal operating conditions. The thermal conditions within the emplacement drifts can be varied, along with the relative humidity, by modifying operational parameters such as the thermal output of the waste packages, the spacing of the waste packages in the emplacement drifts, and the duration and rate of active and passive ventilation. A lower range has been examined to quantify lower-temperature thermal conditions (temperatures and associated humidity conditions) in the emplacement drifts and to quantify impacts to the required emplacement area and excavated drift length. This information has been used to evaluate the potential long-term performance of a lower-temperature repository and to estimate the increase in costs associated with operating a lower-temperature repository. This presentation provides an overview of the thermal management evaluations that have been conducted to investigate a range of repository thermal conditions and

  14. Identification of a Suitable 3D Printing Material for Mimicking Brittle and Hard Rocks and Its Brittleness Enhancements

    Science.gov (United States)

    Zhou, T.; Zhu, J. B.

    2018-03-01

    Three-dimensional printing (3DP) is a computer-controlled additive manufacturing technique which is able to repeatedly and accurately fabricate objects with complicated geometry and internal structures. After 30 years of fast development, 3DP has become a mainstream manufacturing process in various fields. This study focuses on identifying the most suitable 3DP material from five targeted available 3DP materials, i.e. ceramics, gypsum, PMMA (poly(methyl methacrylate)), SR20 (acrylic copolymer) and resin (Accura® 60), to simulate brittle and hard rocks. Firstly, uniaxial compression tests were performed to determine the mechanical properties and failure patterns of the 3DP samples fabricated by those five materials. Experimental results indicate that among current 3DP techniques, the resin produced via stereolithography (SLA) is the most suitable 3DP material for mimicking brittle and hard rocks, although its brittleness needs to be improved. Subsequently, three methods including freezing, incorporation of internal macro-crack and addition of micro-defects were adopted to enhance the brittleness of the 3DP resin, followed by uniaxial compression tests on the treated samples. Experimental results reveal that 3DP resin samples with the suggested treatments exhibited brittle properties and behaved similarly to natural rocks. Finally, some prospective improvements which can be used to facilitate the application of 3DP techniques to rock mechanics were also discussed. The findings of this paper could contribute to promoting the application of 3DP technique in rock mechanics.