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

International Nuclear Information System (INIS)

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

2008-01-01

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

'Practicality' as a key constraint to HLW repository design

International Nuclear Information System (INIS)

Kitayama, Kazumi; Sakabe, Yasushi; Ishiguro, Katsuhiko

2007-01-01

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

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

International Nuclear Information System (INIS)

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

2008-01-01

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

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

International Nuclear Information System (INIS)

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

2002-01-01

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

Thermo-mechanical analysis for multi-level HLW repository concept

International Nuclear Information System (INIS)

Kwon, Sang Ki; Choi, Jong Won

2004-01-01

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

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

International Nuclear Information System (INIS)

Liu Linqing; Yu Yunxiang

1994-01-01

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

The development of safeguards for geological repositories

International Nuclear Information System (INIS)

Van der Meer, K.

2009-01-01

Traditionally, research and development on geological repositories for High Level Waste (HLW) focuses on the short- and long-term safety aspects of the repository. If the repository will also be used for the disposal of spent fuel, safeguards aspects have to be taken into account. Safety and safeguards requirements may be contradictory; the safety of a geological repository is based on the non-intrusion of the geological containment, while safeguards require regular inspections of position and amount of the spent fuel. Examples to reconcile these contradictory requirements are the use of information required for the safety assessment of the geological repository for safeguards purposes and the adaptation of the safeguards approach to use non-intrusive inspection techniques. The principles of an inspection approach for a geological repository are now generally accepted within the IAEA. The practical applicability of the envisaged inspection techniques is still subject to investigation. It is specifically important for the Belgian situation that an inspection technique can be used in clay, the geological medium in which Belgium intends to dispose its HLW and spent fuel. The work reported in this chapter is the result of an international cooperation in the framework of the IAEA, in which SCK-CEN participates

Development of gap filling technique in HLW repository

International Nuclear Information System (INIS)

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

2016-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2006-01-15

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

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

International Nuclear Information System (INIS)

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

2006-01-01

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

SOURCE TERMS FOR HLW GLASS CANISTERS

International Nuclear Information System (INIS)

J.S. Tang

2000-01-01

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

Proposal for geological site selection for L/ILW and HLW repositories. Statement of requirements, procedure and results. Technical report 08-03

International Nuclear Information System (INIS)

2008-10-01

Important steps in the process of managing radioactive wastes have already been implemented in Switzerland. These include the handing and packaging of the waste, waste characterisation and documentation of waste inventories and interim storage along with associated transport. In terms of preparing for deep geological disposal, the necessary scientific and technical work is well advanced and the feasibility of constructing geological repositories that provide the required long-term safety has been successfully demonstrated for all waste types arising in Switzerland. Sufficient knowledge is available to allow the next steps in the selection of repository sites to be defined. The legal framework is also in place and organisational measures have been provided that will allow the tasks to be performed in the coming years to be implemented efficiently. The selection of geological siting regions and sites for repositories in Switzerland will be conducted in three stages. Stage 1 ends with the definition of geological siting regions within which the repository projects will be elaborated in more detail in stages 2 and 3. This report documents and justifies the siting proposals prepared by Nagra for the repositories for low- and intermediate-level waste (L/ILW) and high-level waste (HLW). Formulation of these proposals is conducted in five steps: 1) The waste inventory, which includes reserves for future developments, is allocated to the L/ILW and HLW repositories; 2) Based on this waste allocation, the second step involves defining the barrier and safety concepts for the two repositories. With a view to evaluating the geological siting possibilities, quantitative and qualitative guidelines and requirements on the geology are derived on the basis of these concepts. These relate to the time period to be considered, the space requirements for the repository, the properties of the host rock (depth, thickness, lateral extent, hydraulic conductivity), long-term stability

HLW Disposal System Development

Energy Technology Data Exchange (ETDEWEB)

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

2007-06-15

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

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

International Nuclear Information System (INIS)

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

2014-01-01

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

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

International Nuclear Information System (INIS)

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

2003-01-01

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

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

Geological boundary conditions for a safety demonstration and verification concept for a HLW repository in claystone in Germany. AnSichT

Energy Technology Data Exchange (ETDEWEB)

Stark, Lena; Bebiolka, Anke; Gerardi, Johannes [Federal Institute for Geosciences and Natural Resources (BGR), Hannover (Germany). Dept. of Underground Space for Storage and Economic Use; and others

2015-07-01

Within the framework of the R and D project ''AnSichT'', DBE TECHNOLOGY, BGR and GRS are developing a method to demonstrate the safety of a HLW repository in claystone in Germany. The methodological approach basing on a holistic concept, links the legal and geologic boundary conditions, the disposal and closure concept, the demonstration of barrier integrity, and the long-term analysis of the repository evolution as well. The geologic boundary conditions are specified by the description of the geological situation and generic models, the selection of representative parameters and geoscientific long-term predictions. They form a fundament for the system analysis.

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

International Nuclear Information System (INIS)

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

1985-04-01

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

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

International Nuclear Information System (INIS)

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

2004-01-01

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

HLW Long-term Management Technology Development

International Nuclear Information System (INIS)

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

2010-02-01

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

Application of QA to R ampersand D support of HLW programs

International Nuclear Information System (INIS)

Ryder, D.E.

1988-01-01

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

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.

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

Energy Technology Data Exchange (ETDEWEB)

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

2013-07-01

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

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

International Nuclear Information System (INIS)

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

2013-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2013-06-15

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

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

International Nuclear Information System (INIS)

Biurrun, E.; Hahne, K.

1989-01-01

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

Alternative measure for performance of HLW geologic repository

International Nuclear Information System (INIS)

Joonhang, Ahn; Chambre, P.L.

2001-01-01

A repository performance model that can show effects of canister multiplicity and repository configuration has been developed. Masses of a radionuclide in the repository and in the far field are proposed as performance measures. Canister multiplicity has significant effects on the release of long-lived radionuclides from the repository. As more canisters are included in the same water stream, the radionuclide concentration in the stream increases, but becomes independent of the number of canisters for sufficiently many canisters. Effects of reduction of radionuclide mass in the repository on the repository performance are clearly observed if the canister multiplicity is taken into account and the mass-based measures are applied. (author)

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

International Nuclear Information System (INIS)

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

2002-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-05-15

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

Nuclide Release Behavior from a Repository for a Pyro-process HLW and SF due to Variation of the MWCF Properties

International Nuclear Information System (INIS)

Lee, Youn Myoung; Hwang, Yong Soo

2009-01-01

An assessment program for an optional evaluation of a repository both for disposal of such high-level wastes (HLWs) from various steps of pyro-processes of PWR spent nuclear fuel (SF) and for direct disposal of PWR and CANDU SFs has been developed by utilizing general purpose GoldSim developing tool, by which nuclide transports in the near- and far-field of a repository as well as a transport through a biosphere under various natural and manmade disruptive events affecting a nuclide release could be modeled and evaluated. KAERI has been in charge of modeling and developing assessment tools by which the above mentioned repository system could be assessed in accordance with various features, events, and processes (FEPs) that could happen in and around the repository system. To cope with such various natural and manmade disruptive FEPs as well as normal release scenarios, all the possible cases in view of the Korean circumstances should be modeled and have been evaluated even though we have not yet have any repository. A possible case, among many others, with the variation of such physical properties as the fracture width and the rock matrix diffusion depth, associated with the natural fractures in the geological rock media, along which nuclide could be transported preferentially with the flow of groundwater is considered in the current study. Due to whatever the reason, such as e,g., the earthquake or human intrusion, it is assumed that the physical properties of the major water conducting fault (MWCF) is changed resulting in the size of fracture width and the matrix diffusion depth. For such case another illustration is made for probabilistic evaluation of a hypothetical Korean HLW repository, as similarly done in the previous studies

Siting Process for HLW Repository in Japan

International Nuclear Information System (INIS)

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

2002-01-01

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

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

International Nuclear Information System (INIS)

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

1994-02-01

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

Deep repositories for waste central to uranium debate

International Nuclear Information System (INIS)

Kannegieter, T.

1991-01-01

While no deep repositories for high level wastes (HLW) have yet been constructed it is shown that technology to safely entomb the wastes for tens of thousands of years already exists. The borosilicate glass (vitrification) developed in France has been accepted by all countries who are reprocessing. Meanwhile, the Australian Synroc has not yet been put into service. Synroc developers at the Australian Nuclear Science and Technology Organization believe it will be the second generation waste form. The advantages and disadvantages of both technologies are briefly discussed as well as some of the regulatory, political, legal and technical conflicts surrounding the issue of HLW repositories. 1 tab., ills

EBS modelling for the development of repository concepts tailored to siting environments

International Nuclear Information System (INIS)

Ishiguro, K.; Ueda, H.; Wakasugi, K.; Sakabe, Y.; Kitayama, K.; Umeki, H.; Takase, H.

2007-01-01

The Japanese siting approach for a HLW repository calls for volunteer host municipalities and thereby places particular emphasis on design flexibility. In particular, the repository concept needs to be tailored to the specific site characteristics. Starting from the H12 repository concept, NUMO has been examining a range of possible repository design options, including the EBS. In this paper. the requirements and strategy for the development of models for performance assessment and process understanding are discussed, taking into account the step-wise, iterative process of developing repository concepts. The areas requiring further development of models and databases in the long-term R and D programme have been identified as a 'wish list' that relates to the evaluation of a range of potential repository concepts, focusing on the near-field for the early stages of the development process. Among the issues on the list, NUMO has started the development of a flexible computer code for modelling three-dimensional mass transport to evaluate the impacts of various design options and components of the EBS. This tool has been applied to the analysis of the barrier effects of the tunnel plugs located in fractured rock. (authors)

Modelling spent fuel and HLW behaviour in repository conditions

International Nuclear Information System (INIS)

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

2003-01-01

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

Modelling spent fuel and HLW behaviour in repository conditions

Energy Technology Data Exchange (ETDEWEB)

Esparza, A M; Esteban, J A

2003-07-01

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

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

International Nuclear Information System (INIS)

Feldman, D.L.

1991-01-01

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

Biosphere modelling for a HLW repository - scenario and parameter variations

International Nuclear Information System (INIS)

Grogan, H.

1985-03-01

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

Licensing information needs for a high-level waste repository

International Nuclear Information System (INIS)

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

1985-01-01

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

Development of methodology to construct a generic conceptual model of river-valley evolution for performance assessment of HLW geological disposal

International Nuclear Information System (INIS)

Kawamura, Makoto; Tanikawa, Shin-ichi; Yasue, Ken-ichi; Niizato, Tadafumi

2011-01-01

In order to assess the long-term safety of a geological disposal system for high-level radioactive waste (HLW), it is important to consider the impact of uplift and erosion, which cannot be precluded on a timescale in the order of several hundred thousand years for many locations in Japan. Geomorphic evolution, caused by uplift and erosion and coupled to climatic and sea-level changes, will impact the geological disposal system due to resulting spatial and temporal changes in the disposal environment. Degradation of HLW barrier performance will be particularly significant when the remnant repository structures near, and are eventually exposed at, the ground surface. In previous studies, fluvial erosion was densified as the key concern in most settings in Japan. Interpretation of the impact of the phenomena at relevant locations in Japan has led to development of a generic conceptual model which contains the features typical at middle reach of rivers. Here, therefore, we present a methodology for development of a generic conceptual model based on best current understanding of fluvial erosion in Japan, which identifies the simplifications and uncertainties involved and assesses their consequences in the context of repository performance. (author)

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

International Nuclear Information System (INIS)

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

2006-01-01

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

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

International Nuclear Information System (INIS)

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

2002-01-01

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

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

International Nuclear Information System (INIS)

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

1992-12-01

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

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

Science.gov (United States)

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

2017-09-01

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

The status of radioactive waste repository development in the United States - December 2011

International Nuclear Information System (INIS)

Hill, David R.

2012-01-01

The current state of affairs concerning development in the United States of a permanent repository for disposal of spent nuclear fuel (SNF) and high-level radioactive waste (HLW) is, in a word, uncertain. The President of the United States has asserted that he believes licensing and development of the Yucca Mountain repository should be abandoned, while other important parties believe licensing and development should continue. And not surprisingly, there is a disagreement as to what the law requires and whether the licensing process for the Yucca Mountain repository can be terminated at this point, even if the President would like for that to happen. The future of Yucca Mountain, and the future of radioactive waste disposal in the United States generally, currently are pending before the US Court of Appeals for the District of Columbia Circuit, and eventually the Supreme Court of the United States may decide some of the important legal issues concerning Yucca Mountain's future. The November 2012 US elections also likely will have a significant impact on future radioactive waste repository development

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

Long-term stability of the near-field about high-level radioactive waste repository in thermo-hydro-mechanical coupling action condition

International Nuclear Information System (INIS)

Liu Yuemiao; Wang Ju; Ke Dan; Cai Meifeng

2008-01-01

It is a long-term process for the high-level radioactive waste repository, from opening, construction to end of its service. The long-term stability of the near-field is the key issue for the design of HLW repository because the opening and heat generated from the HLW. Through a nationwide investigation, Beishan area, a Gobi desert in Gansu province, is considered as a suitable candidate and GMZ bentonite deposit which located in Xinghe County, Inner Mongolia has been proposed for the supplier of buffer/backfill material for HLW geological repository in China. According to the R and D guide of high-level radioactive waste disposal in China, the 3D model of HLW repository with high-level radioactive waste, canister and buffer/backfill material is established using FLAC3D. To take into account in situ stress, geothermal gradient, groundwater, thermal relief of HLW and swelling pressure of buffer/backfill material, the evolution of temperature, stress and displacement of HLW repository under thermo-mechanical coupling, hydro-mechanical coupling and thermo-hydro-mechanical coupling conditions was analyzed respectively. The long-term stability of HLW repository in Beishan area was studied. (authors)

MRS/IS facility co-located with a repository: preconceptual design and life-cycle cost estimates

International Nuclear Information System (INIS)

Smith, R.I.; Nesbitt, J.F.

1982-11-01

A program is described to examine the various alternatives for monitored retrievable storage (MRS) and interim storage (IS) of spent nuclear fuel, solidified high-level waste (HLW), and transuranic (TRU) waste until appropriate geologic repository/repositories are available. The objectives of this study are: (1) to develop a preconceptual design for an MRS/IS facility that would become the principal surface facility for a deep geologic repository when the repository is opened, (2) to examine various issues such as transportation of wastes, licensing of the facility, and environmental concerns associated with operation of such a facility, and (3) to estimate the life cycle costs of the facility when operated in response to a set of scenarios which define the quantities and types of waste requiring storage in specific time periods, which generally span the years from 1990 until 2016. The life cycle costs estimated in this study include: the capital expenditures for structures, casks and/or drywells, storage areas and pads, and transfer equipment; the cost of staff labor, supplies, and services; and the incremental cost of transporting the waste materials from the site of origin to the MRS/IS facility. Three scenarios are examined to develop estimates of life cycle costs of the MRS/IS facility. In the first scenario, HLW canisters are stored, starting in 1990, until the co-located repository is opened in the year 1998. Additional reprocessing plants and repositories are placed in service at various intervals. In the second scenario, spent fuel is stored, starting in 1990, because the reprocessing plants are delayed in starting operations by 10 years, but no HLW is stored because the repositories open on schedule. In the third scenario, HLW is stored, starting in 1990, because the repositories are delayed 10 years, but the reprocessing plants open on schedule

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

International Nuclear Information System (INIS)

Ove Arup and Partners, London

1990-01-01

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

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

International Nuclear Information System (INIS)

Lee, Youn-Myoung; Hwang, Yong-Soo

2007-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-11-01

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

Discrete and continuum approaches for the analysis of coupled thermal-mechanical processes in the near field of a HLW repository

International Nuclear Information System (INIS)

Shimizu, Hiroyuki; Fujita, Tomoo; Nakama, Shigeo; Koyama, Tomofumi; Chijimatsu, Masakazu

2011-01-01

This paper reports on the results of the numerical simulations for the analysis of coupled thermal-mechanical processes in the near field of a HLW repository using Finite Element Method (FEM) and Distinct Element Method (DEM). The FEM approach provides quantitative information of the change of stress during excavation and heating process. On the other hand, the DEM approach shows the crack propagation process at the borehole surface, and this result agrees qualitatively well with the experimental observation. By comparing these results obtained from both approaches, quantitative and qualitative insights into various aspects of the processes occurred in the near field can be obtained. (author)

Siting regions for deep geological repositories. Why just here?

International Nuclear Information System (INIS)

Rieser, A.

2009-09-01

This report helps to the popularization of the Nagra works accomplished for the management and disposal of the radioactive wastes in Switzerland. The programme for management and disposal of the radioactive wastes are extensively determined by regulations. Protection of mankind and environment is the primary objective. The basic storage process is considered as having been solved. The question addressed in the report is where the facility has to be built; the site selection procedure includes five steps: 1) according to their type the wastes have to be allocated to two different repositories: for low- and intermediate-level wastes (L/ILW), and for high-level and alpha-toxic wastes (HLW); 2) the safety concept for both repositories and the requirements on the geology have to be determined; 3) large suitable geological-tectonic zones must be found where repositories could be built; 4) in these geological zones a suitable host rock has to be identified; 5) the most important spatial geological conditions of the host rock (minimum depth with respect to surface erosion, maximum depth in terms of engineering requirements, lateral extent) have to be identified. Based on these criteria, three suitable siting regions for a HLW repository were found in the North of Switzerland. The preferred host rock is Opalinus clay because of its very low permeability; it is therefore an excellent barrier against nuclide transport. In the three proposed siting regions, Opalinus clay is present in sufficient volumes at a suitable depth. For a L/ILW repository six different possible siting regions were identified, five in Northern Switzerland and one in Central Switzerland. In the three siting regions found for a possible HLW repository, it would also be possible to built a combined repository for both HLW and L/ILW wastes

The general situation of clay site for high-level waste geological disposal repository

International Nuclear Information System (INIS)

Wang Changxuan; Liu Xiaodong; Liu Pinghui

2008-01-01

Host medium is vitally important for safety of high-level radiaoactive waste (HLW) geological disposal. Clay, as host media of geological repository of HLW, has received greater attention for its inherent advantages. This paper summarizes IAEA and OECD/NEA's some safety guides on site selection and briefly introduces the process of the site selection, their studies and the characteristics of the clay formations in Switz-erland, France and Belgian. Based on these analyses, some suggestions are made to China's HLW repository clay site selection. (authors)

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

International Nuclear Information System (INIS)

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

1999-01-01

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

Status of Equipment Development for a High-Level Waste Repository in Germany

International Nuclear Information System (INIS)

Biurrun, E.; Haverkamp, B.; Filbert, W.; Bollingerfehr, W.; Graf, R.

2009-01-01

In Germany, a potential site for a deep geological repository mainly for heat generating high-level waste was selected back on February 22, 1977: a salt dome located near the village of Gorleben, at the shores of the Elbe River in Northern Germany. Concurrent with site exploration from the surface, and later exploration mine construction, a large-scale R and D effort was conducted to have the science and technology needed to license and later operate a repository available when needed. An important part of these efforts was the development and 1:1 scale demonstration of all technologies required to run a repository, which had not been state-of-the-art. Underground exploration of the Gorleben site has come to a moratorium due to political decisions taken by the former Federal Government formed by a coalition between the Social Democrats and the Green Party back in 2000. However, research and development in regard to emplacement techniques and equipment has continued. This paper reflects the consistent, long-term technological effort that has led to having available all advanced technologies required to run a repository for spent fuel and HLW. The paper details previous achievements and points at present optimization work. (authors)

HLW Glass Studies: Development of Crystal-Tolerant HLW Glasses

Energy Technology Data Exchange (ETDEWEB)

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

2012-04-02

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-09-14

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

Oxidation-reduction reactions. Overview and implications for repository studies

International Nuclear Information System (INIS)

Apted, Michael J.; Arthur, Randolph C.; Sasamoto, Hiroshi; Yui, Mikazu; Iwatsuki, Teruki

2001-02-01

The purpose of this report is to provide a survey and review on oxidation-reduction ('redox') reactions, with particular emphasis on implications for disposal of high-level waste (HLW) in deep geological formations. As an overview, the focus is on basic principles, problems, and proposed research related specifically to the assessment of redox for a HLW repository in Japan. For a more comprehensive treatment of redox and the myriad associated issues, the reader is directed to the cited textbooks used as primary references in this report. Low redox conditions in deep geological formations is a key assumption in the 'Second Progress Report on Research and Development for the Geological Disposal of HLW in Japan' (hereafter called H12'). The release behavior of multi-valent radioelements (e.g., Tc, Se, U, Pu, Np), as well as daughter radioelements of these radioelements, from a deep geological repository are sensitively related to redox conditions. Furthermore, the performance of certain barrier materials, such as overpack and buffer, may be impacted by redox conditions. Given this importance, this report summarizes some key topics for future technical studies supporting site characterization and repository performance as follows: To fully test the conceptual models for system Eh, it will be necessary to measure and evaluate trace element and isotopic information of both coexisting groundwater and reactive minerals of candidate rocks. Because of importance of volatile species (e.g., O 2 , H 2 etc.) in redox reactions, and given the high total pressure of a repository located 500 to 1000 meter deep, laboratory investigations of redox will necessarily require use of pressurized test devices that can fully simulate repository conditions. The stability (redox capacity) of the repository system with respect to potential changes in redox boundary condition induced by oxidizing waters intrusion should be established experimentally. An overall conceptual model that unifies

A GoldSim Model and a Sensitivity Study for Safety Assessment of a Repository for Disposal of Spent Nuclear Fuel

International Nuclear Information System (INIS)

Lee, Youn Myoung; Hwang, Yong Soo

2008-11-01

An assessment program for the evaluation of a high-level waste (HLW) repository has been developed by utilizing GoldSim, by which nuclide transports in the near- and far-field of a repository as well as a transport through a biosphere under various natural and manmade disruptive events affecting a nuclide release could be modeled and evaluated. To demonstrate its usability, three illustrative cases including the influence of a groundwater flow pattern through canisters associated with a flowing groundwater through fractures, and the possible disruptive events caused by an accidental human intrusion or an earthquake have been investigated and illustrated for a hypothetical Korean HLW repository

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

Directory of Open Access Journals (Sweden)

Yuemiao Liu

2014-04-01

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

Source term measurements on vitrified HLW

International Nuclear Information System (INIS)

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

1988-01-01

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

Technical and economic optimization study for HLW waste management

International Nuclear Information System (INIS)

Deffes, A.

1989-01-01

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

VerSi - A Methodology for a Comparison of Potential Repository Sites

International Nuclear Information System (INIS)

Hund, Wilhelm

2010-09-01

In the year 2000 the moratorium on the exploration of the Gorleben salt dome as a potential repository for all kinds of radioactive waste became effective as a result of the consensus agreement between the Federal Government and the utilities about phasing out nuclear energy in Germany. All exploration activities were interrupted for at maximum ten years to clarify conceptual and safety relevant questions. A new set of safety requirements for the final disposal of heat-generating radioactive waste in deep geological formations was established in July 2009 by the Federal Ministry for the Environment, Nature Conservation and Nuclear Safety (BMU). As the BMU intended to carry out a comparison of potential repository sites it was necessary to initiate the development of a methodology for the identification of the site with the highest level of safety. A comparison of different repository sites requires a tool ensuring most confident and objective criteria for the comparison, whereas up to present long-term safety analyses were focused on confirming the suitability of sites by meeting the protection objectives by the measures of dose and risk. Within the 2006 established project VerSi a methodology for comparing different sites in different host rocks will be developed on the basis of long-term safety analyses taking into account geoscientific databases, inventory of radioactive waste, waste containers, corresponding disposal concepts and the feasibility of appropriate backfilling and closure concepts. The development of the method is aiming at providing measures other than dose and risk for the evaluation of the level of safety. For testing the tools a HLW-repository hosted in a salt dome (Gorleben) will be compared with a generic HLW-repository in consolidated clay as a host rock. As until now in Germany no clay stone site has been investigated for hosting a HLW repository, the required data are transferred from international research projects and repository concepts

Industrial complementarities between interim storage and reversible geological repository - 59237

International Nuclear Information System (INIS)

Hoorelbeke, Jean-Michel

2012-01-01

The French Act voted in 2006 made the choice of deep geological disposal as the reference option for the long term management of high level (HLW) and intermediate level long-lived waste. The CIGEO repository project aims at avoiding or limiting burden to future generations, which could not be achieved by the extension in time of interim storage. The reversibility as provided by the Act will maintain a liberty of choice for waste management on a duration which is comparable to new storage facility. Interim storage is required to accommodate waste as long as the repository is not available. The commissioning of the repository in 2025 will not suppress needs for interim storage. The paper describes the complementarities between existing and future interim storage facilities and the repository project: repository operational issues and planning, HLW thermal decay, support for the reversibility, etc. It shows opportunities to prepare a global optimization of waste management including the utilization at best of storage capacities and the planning of waste emplacement in the repository in such a way to facilitate operational conditions and to limit cost. Preliminary simulations of storage-disposal scenarios are presented. Thanks to an optimal use of the waste management system, provision can be made for a progressive increase of waste emplacement flow during the first operation phase of the repository. It is then possible to stabilize the industrial activity level of the repository site. An optimal utilization of interim storage can also limit the diversity of waste packages emplaced simultaneously, which facilitates the operation of the repository. 60 years minimum interim storage duration is generally required with respect to HLW thermal output. Extending this interim storage period may reduce the underground footprint of the repository. Regarding reversibility, the capability to manage waste packages potentially retrieved from the repository should be analyzed. The

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.)

The surface mock-up KENTEX: on the thermal-hydro-mechanical behaviors in the buffer of a Korean HLW repository

International Nuclear Information System (INIS)

Lee, Jae Owan; Cho, Won Jin; Choi, Jong Won

2008-01-01

The concept for a disposal of high-level wastes (HLW) in Korea is based upon a multi barrier system composed of engineered barriers and its surrounding plutonic rock (Kang et. al., 2002). A repository is constructed in a bedrock of several hundred meters in depth below the ground surface. The engineered barrier system (EBS), which is similar to the configuration considered by many other countries, consists of the HLW-encapsulating disposal container, the buffer between the container and the wall of a borehole, and the backfill in the inside space of the emplacement room, to isolate the HLW from the surrounding rock masses. The engineering performance of a HLW repository is dependent, to a large extent, upon the thermal-hydro-mechanical (THM) behaviors in the buffer which are complicated by the processes such as the decay heat generated from the HLW, the ground water flowing in from the surrounding host rock, and the swelling pressure exerted by compacted bentonite. For this reason, the Korea Atomic Energy Research Institute (KAERI), to investigate the THM behaviors in the buffer of the Korean reference disposal system (KRS), planned large-scale tests to be conducted in two stages: a surface mock-up and then a full-scale 'in situ' test. This paper deals with the surface mock-up called as 'KENTEX' and presents the THM behaviors in the buffer which have been investigated from the KENTEX test. The KENTEX is a third scale of the KRS. It consists of five major components: a heating system, a confining cylinder, a hydration tank, bentonite blocks, and sensors and instruments. The heating system measures 0.41 m in diameter and 0.68 m in length, which includes three heating elements in its inside, capable of supplying a thermal power of 1 kW each. The confining cylinder, which plays a role of the wall of a borehole excavated in the host rock, is a steel body with a length of 1.36 m and an inner diameter of 0.75 m, the inside wall of which is lined with layers of geotextile

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

International Nuclear Information System (INIS)

Lindstroem Jensen, K.E.

1987-01-01

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

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.)

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

International Nuclear Information System (INIS)

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

2010-04-01

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

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

International Nuclear Information System (INIS)

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

1992-01-01

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

Coupled thermo-mechanical analysis of granite for high-level radioactive waste repository

International Nuclear Information System (INIS)

Liu Wengang; Wang Ju; Zhou Hongwei; Jiang Pengfei; Yang Chunhe

2008-01-01

High-level radioactive wastes (HLW) repository is a special deep underground engineering, and in the stages of site selection, designing, constructing ,the stability evaluation, lots of important rock mechanics problems need to be resolved. During the decay of nuclear waste, enormous thermal energy was released and temperature variation caused dynamic distribution of stress and deformation field of surrounding rock of repository. BeiShan region of Gansu province was selected to be the repository field in the future, it is of practical significance to do research on granite in this region. Based on the concept model of HLW repository, this thesis calculates temperature field, stress field and deformation field of HLW repository surrounding rock under the condition of TM coupled with applying the finite difference FLAC 3D . From this study, thermo-mechanical characteristic of granite is obtained primarily under given canister heat source and given decay law function. And these results show that the reasonable space between disposal hole is 8 m-12 m, and the peak temperature of the canister surface is 130 ℃, the centerline temperature between pits is about 40 ℃ which is maintained for about hundreds of years under given heating output at -500 m depth. (authors)

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

International Nuclear Information System (INIS)

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

2011-01-01

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

Optimization method for dimensioning a geological HLW waste repository

International Nuclear Information System (INIS)

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

1990-01-01

This method was developed by the CEA to optimize the dimensions of a geological repository by taking account of technical and economic parameters. It involves optimizing radioactive waste storage conditions on the basis of economic criteria with allowance for specified thermal constraints. The results are intended to identify trends and guide the choice from among available options: simple and highly flexible models were therefore used in this study, and only nearfield thermal constraints were taken into consideration. Because of the present uncertainty on the physicochemical properties of the repository environment and on the unit cost figures, this study focused on developing a suitable method rather than on obtaining definitive results. The optimum values found for the two media investigated (granite and salt) show that it is advisable to minimize the interim storage time, implying the containers must be separated by buffer material, whereas vertical spacing may not be required after a 30-year interim storage period. Moreover, the boreholes should be as deep as possible, on a close pitch in widely spaced handling drifts. These results depend to a considerable extent on the assumption of high interim storage costs

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

International Nuclear Information System (INIS)

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

2000-01-01

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

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

International Nuclear Information System (INIS)

Liu Xiaodong; Luo Taian; Zhu Guoping; Chen Qingchun

2007-12-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2007-12-15

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

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

International Nuclear Information System (INIS)

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

2008-01-01

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

Systems study of the feasibility of high-level nuclear waste fractionation for thermal stress control in a geologic repository: appendices

International Nuclear Information System (INIS)

McKee, R.W.; Elder, H.K.; McCallum, R.F.; Silviera, D.J.; Swanson, J.L.; Wiles, L.E.

1983-06-01

This study assesses the benefits and costs of fractionating the cesium and strontium (Cs/Sr) components in commercial high-level waste (HLW) to a separate waste stream for the purpose of reducing geologic-repository thermal stresses in the region of the HLW. The major conclusion is that the Cs/Sr fractionation concept offers the prospect of a substantial total system cost advantage for HLW disposal if reduced HLW package temperatures in a basalt repository are desired. However there is no cost advantage if currently designated maximum design temperatures are acceptable. Aging the HLW for 50 to 100 years can accomplish similar results at equivalent or lower costs. Volume II contains appendices for: (1) thermal analysis supplement; (2) fractionation process experimental results supplement; (3) cost analysis supplement; and (4) radiological risk analysis supplement

Nuclear Regulatory Commission staff development of the license application review plan for a high-level radioactive waste repository

International Nuclear Information System (INIS)

Johnson, R.L.; Holonich, J.J.; Lee, M.P.; Delligatti, M.S.

1993-01-01

The Nuclear Regulatory Commission staff has recently started a new initiative to develop the License Application Review Plan (LARP) which the staff will use in its reviews of the U.S. Department of Energy's (DOE's) license application (LA) for a geologic repository for the disposal of high-level radioactive waste (HLW). This paper describes the staff's approach for developing the LARP, the development schedule and current status, the organization and content of the LARP, and the staff's LA review strategy. Therefore, it gives a preview of the draft LARP which will be made available in late 1993. It also describes how the LARP will be used as guidance to the staff in conducting reviews of regulatory and technical issues important to the licensing of a geologic repository. Finally, the benefits to the NRC staff, DOE, and other parties are discussed

Reduction of HLW repository environmental impact by Na-Cooled FBR

International Nuclear Information System (INIS)

Ahn, Joonhong; Ikegami, Tetsuo

2005-01-01

The present study proposes the total toxicity index of radionuclides that have accumulated in the region exterior to the repository as the environmental impact measure. The measure is quantitatively evaluated by a radionuclide transport models that incorporate the effects of canister-array configuration and the initial mass loading in the waste canister. With the measure, it is demonstrated that the environmental impact of the repository can be effectively reduced by reduction of the initial mass loading resulting from FBR deployment and change in the canister-array configuration in the repository. The rate of increase in the environmental impact with the increase in the repository size can be reduced by reducing the initial mass loading of Np and its precursors. (author)

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)

Siting regions for deep geological repositories. Why just here?; Standortgebiete fuer geologische Tiefenlager. Warum gerade hier?

Energy Technology Data Exchange (ETDEWEB)

Rieser, A

2009-09-15

This report helps to the popularization of the Nagra works accomplished for the management and disposal of the radioactive wastes in Switzerland. The programme for management and disposal of the radioactive wastes are extensively determined by regulations. Protection of mankind and environment is the primary objective. The basic storage process is considered as having been solved. The question addressed in the report is where the facility has to be built; the site selection procedure includes five steps: 1) according to their type the wastes have to be allocated to two different repositories: for low- and intermediate-level wastes (L/ILW), and for high-level and alpha-toxic wastes (HLW); 2) the safety concept for both repositories and the requirements on the geology have to be determined; 3) large suitable geological-tectonic zones must be found where repositories could be built; 4) in these geological zones a suitable host rock has to be identified; 5) the most important spatial geological conditions of the host rock (minimum depth with respect to surface erosion, maximum depth in terms of engineering requirements, lateral extent) have to be identified. Based on these criteria, three suitable siting regions for a HLW repository were found in the North of Switzerland. The preferred host rock is Opalinus clay because of its very low permeability; it is therefore an excellent barrier against nuclide transport. In the three proposed siting regions, Opalinus clay is present in sufficient volumes at a suitable depth. For a L/ILW repository six different possible siting regions were identified, five in Northern Switzerland and one in Central Switzerland. In the three siting regions found for a possible HLW repository, it would also be possible to built a combined repository for both HLW and L/ILW wastes.

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

International Nuclear Information System (INIS)

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

1991-01-01

Under the sponsorship of the US Nuclear Regulatory Commission (NRC), Sandia National Laboratories (SNL) is developing a performance assessment methodology for the analysis of long-term disposal and isolation of high-level nuclear wastes (HLW) in alternative geologic media. As part of this exercise, SNL created a conceptualization of ground-water flow and radionuclide transport in the far field of a hypothetical HLW repository site located in unsaturated, fractured tuff formations. This study provides a foundation for the development of conceptual mathematical, and numerical models to be used in this performance assessment methodology. This conceptualization is site specific in terms of geometry, the regional ground-water flow system, stratigraphy, and structure in that these are based on information from Yucca Mountain located on the Nevada Test Site. However, in terms of processes in unsaturated, fractured, porous media, the model is generic. This report also provides a review and evaluation of previously proposed conceptual models of unsaturated and saturated flow and solute transport. This report provides a qualitative description of a hypothetical HLW repository site in fractured tuff. However, evaluation of the current knowledge of flow and transport at Yucca Mountain does not yield a single conceptual model. Instead, multiple conceptual models are possible given the existing information

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

Energy Technology Data Exchange (ETDEWEB)

Parsons, A.M.; Olague, N.E.; Gallegos, D.P. [Sandia National Labs., Albuquerque, NM (USA)

1991-01-01

Under the sponsorship of the US Nuclear Regulatory Commission (NRC), Sandia National Laboratories (SNL) is developing a performance assessment methodology for the analysis of long-term disposal and isolation of high-level nuclear wastes (HLW) in alternative geologic media. As part of this exercise, SNL created a conceptualization of ground-water flow and radionuclide transport in the far field of a hypothetical HLW repository site located in unsaturated, fractured tuff formations. This study provides a foundation for the development of conceptual mathematical, and numerical models to be used in this performance assessment methodology. This conceptualization is site specific in terms of geometry, the regional ground-water flow system, stratigraphy, and structure in that these are based on information from Yucca Mountain located on the Nevada Test Site. However, in terms of processes in unsaturated, fractured, porous media, the model is generic. This report also provides a review and evaluation of previously proposed conceptual models of unsaturated and saturated flow and solute transport. This report provides a qualitative description of a hypothetical HLW repository site in fractured tuff. However, evaluation of the current knowledge of flow and transport at Yucca Mountain does not yield a single conceptual model. Instead, multiple conceptual models are possible given the existing information.

Development Of Glass Matrices For HLW Radioactive Wastes

International Nuclear Information System (INIS)

Jantzen, C.

2010-01-01

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

A Nuclide Release Model for a Deep Well Scenario near a Repository

International Nuclear Information System (INIS)

Lee, Youn Myoung; Jung, Jong Tae; Choi, Jong Won

2010-01-01

Recently several template programs ready for the safety assessment of a high-level radioactive waste repository (HLW) and a low- and intermediate level radioactive waste repository (LILW) systems, that are conceptually modeled, have been developed by utilizing GoldSim and AMBER at KAERI. During the last few years, such template programs have been utilized for the evaluation of nuclide transports in the nearand far-field of a repository as well as transport through the biosphere under various normal and disruptive release scenarios with assumed data. The GoldSim program, another template program by which influence due to a well located very near to the repository has been modeled and evaluated for an assumed case in order to simulate the worst exposure scenario. This seems very useful to evaluate an accidental event a long time after closure of the repository has been developed and illustrated

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

International Nuclear Information System (INIS)

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

2012-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2012-12-13

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

Safety assessment of HLW geological disposal system

International Nuclear Information System (INIS)

Naito, Morimasa

2006-01-01

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

DEVELOPMENT OF GLASS MATRICES FOR HLW RADIOACTIVE WASTES

Energy Technology Data Exchange (ETDEWEB)

Jantzen, C.

2010-03-18

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

Effects of repository conditions on environmental impact reduction by recycling

International Nuclear Information System (INIS)

Ahn, Joonhong

2010-01-01

The environmental impacts (EI) of high-level wastes (HLW) disposed of in a water-saturated repository (WSR) and in the Yucca Mountain Repository (YMR) for various fuel cycle cases have been evaluated and compared to observe the difference in the recycling effects for differing repository conditions. With the impacts of direct spent fuel disposal in each repository as the reference level, separation of actinides by Urex+ and borosilicate vitrification clearly reduces the environmental impacts of YMR, while separation by Purex and borosilicate vitrification would not necessarily reduce the environmental impact of WSR. (authors)

Salt Repository Project Waste Package Program Plan: Draft

International Nuclear Information System (INIS)

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

1986-01-01

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

Compliance with NRC subsystem requirements in the repository licensing process

International Nuclear Information System (INIS)

Minwalla, H.

1994-01-01

Section 121 of the Nuclear Waste Policy Act of 1982 requires the Nuclear Regulatory Commission (Commission) to issue technical requirements and criteria, for the use of a system of multiple barriers in the design of the repository, that are not inconsistent with any comparable standard promulgated by the Environmental Protection Agency (EPA). The Administrator of the EPA is required to promulgate generally applicable standards for protection of the general environment from offsite releases from radioactive material in repositories. The Commission's regulations pertaining to geologic repositories are provided in 10 CFR part 60. The Commission has provided in 10 CFR 60.112 the overall post-closure system performance objective which is used to demonstrate compliance with the EPA high-level waste (HLW) disposal standard. In addition, the Commission has provided, in 10 CFR 60.113, subsystem performance requirements for substantially complete containment, fractional release rate, and groundwater travel time; however, none of these subsystem performance requirements have a causal technical nexus with the EPA HLW disposal standard. This paper examines the issue of compliance with the conflicting dual regulatory role of subsystem performance requirements in the repository licensing process and recommends several approaches that would appropriately define the role of subsystem performance requirements in the repository licensing process

Radionuclide release from high level waste forms under repository conditions in clay or granite

International Nuclear Information System (INIS)

Godon, N.; Lanza, F.

1990-01-01

The behaviour of both fully active and simulated vitrified high level waste (HLW) has been studied under conditions that are likely to occur in future repositories in clay and granite. The simulated HLW was doped with Cs, Sr, Tc and the actinides and the leaching of these elements from the glass has been measured together with their concentrations in the water of the near-field and their distribution between the various components of the repository. The diffusion coefficients of several elements in Boom clay has also been measured. The results show that the concentrations of Tc and the actinides in the near-field of a repository will be very low and that the actinides will only diffuse very slowly away from the vicinity of the glass. 24 refs., 1 figs., 7 tabs

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

International Nuclear Information System (INIS)

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

1986-01-01

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

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

International Nuclear Information System (INIS)

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

2001-03-01

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

High-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 high-level waste (HLW) 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 on the host rock and engineered barriers arising principally from the heat generated by the waste. - Rock-mechanical effects: i.e. effects arising from the mechanical disturbance to the rock caused by the excavation of the emplacement rooms 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, with a focus on chemical effects of the waste and engineered materials on the host rock. The assessment of the repository-induced effects shows that detrimental chemical and mechanical impacts are largely confined to the rock immediately adjacent to the excavations, thermal impacts are controllable by limiting the heat load and gas effects are limited by ensuring acceptably low gas production rates and by the natural tendency of the gas to escape along the excavations and the excavation damaged zone (EDZ) rather than through the undisturbed rock. Specific measures that are part of the current reference design are discussed in relation to their significance with respect to repository-induced effects. The SF/HLW emplacement rooms (emplacement drifts) are designed, constructed, operated and finally backfilled in such a way that formation of excavation damaged zones is limited. Specifically this is achieved

High-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 high-level waste (HLW) 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 on the host rock and engineered barriers arising principally from the heat generated by the waste. - Rock-mechanical effects: i.e. effects arising from the mechanical disturbance to the rock caused by the excavation of the emplacement rooms 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, with a focus on chemical effects of the waste and engineered materials on the host rock. The assessment of the repository-induced effects shows that detrimental chemical and mechanical impacts are largely confined to the rock immediately adjacent to the excavations, thermal impacts are controllable by limiting the heat load and gas effects are limited by ensuring acceptably low gas production rates and by the natural tendency of the gas to escape along the excavations and the excavation damaged zone (EDZ) rather than through the undisturbed rock. Specific measures that are part of the current reference design are discussed in relation to their significance with respect to repository-induced effects. The SF/HLW emplacement rooms (emplacement drifts) are designed, constructed, operated and finally backfilled in such a way that formation of excavation damaged zones is limited. Specifically this is achieved

Situation concerning the HLW repository in Germany

International Nuclear Information System (INIS)

Lempert, J.P.

1992-01-01

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

Industrial scale-plant for HLW partitioning in Russia

International Nuclear Information System (INIS)

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

1996-01-01

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

Construction of Basic Evaluation Criteria for Candidate HLW Repository Sites (1)

International Nuclear Information System (INIS)

Koh, Yong Kwon; Bae, Dae Seok; Kim, Kyung Su; Kim, Geon Young; Park, Kyung Woo; Ji, Sung Hoon; Ryu, Ji Hun

2009-08-01

We constructed the preliminary site assessment system for selection of proper site as high level radioactive disposal repository with consideration of Korean geological characteristics and underground environments. And, site assessment factor and standard as a point of geologic aspect were suggested for decision of candidate radioactive disposal site in Korea. The results can be used to develop the geological information system for assessment of candidate disposal site

Analysis of the processes defining radionuclide migration from deep geological repositories in porous medium

International Nuclear Information System (INIS)

Brazauskaite, A.; Poskas, P.

2004-01-01

Due to the danger of exposure arising from long-lived radionuclides to humans and environment, spent nuclear fuel (SNF) and high level waste (HLW) are not allowed to be disposed of in near surface repositories. There exists an international consensus that such high level and long-lived radioactive wastes are best disposed of in geological repositories using a system of engineered and natural barriers. At present, the geological repository of SNF and HLW has not been realized yet in any country but there is a lot of experience in the assessment of radionuclide migration from deep repositories, investigations of different processes related to the safety of a disposal system. The aim of this study was to analyze the processes related to the radionuclide migration from deep geological repositories in porous medium such as SNF matrix dissolution, release mechanism of radionuclides from SNF matrix, radionuclide solubility, sorption, diffusive, advective transport of radionuclides from the canister and through the engineered and natural barriers. It has been indicated that SNF matrix dissolution, radionuclide solubility and sorption are sensitive to ambient conditions prevailing in the repository. The approaches that could be used for modeling the radionuclide migration from deep repositories in porous medium are also presented. (author)

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)

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

International Nuclear Information System (INIS)

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

2004-01-01

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

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

International Nuclear Information System (INIS)

Put, M.; Henrion, P.

1992-01-01

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

Establishing the concept of buffer for a high-level radioactive waste repository: An approach

Energy Technology Data Exchange (ETDEWEB)

Lee, Jae Owan; Lee, Min Soo; Choi, Heui Joo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-12-15

The buffer is a key component of the engineered barrier system in a high-level radioactive waste (HLW) repository. The present study reviewed the requirements and functional criteria of the buffer reported in literature, and also based on the results, proposed an approach to establish a buffer concept which is applicable to an HLW repository in Korea. The hydraulic conductivity, radionuclide-retarding capacity (equilibrium distribution coefficient and diffusion coefficient), swelling pressure, thermal conductivity, mechanical properties, organic carbon content, and initialization rate were considered as major technical parameters for the functional criteria of the buffer. Domestic bentonite (Ca-bentonite) and, as an alternative, MX-80 (Na-bentonite) were proposed for the buffer of an HLW repository in Korea. The technical specifications for those proposed bentonites were set to parameter values that conservatively satisfy Korea's functional criteria for the Ca-bentonite and Swedish criteria for the Na-bentonite. The thickness of the buffer was determined by evaluating the means of shear behavior, radionuclide release, and heat conduction, which resulted in the proper buffer thickness of 0.25 to 0.5 m. However, the final thickness of the buffer should be determined by considering coupled thermal-hydraulic-mechanical evaluation and economics and engineering aspects as well.

HLW Tank Space Management, Final Report

International Nuclear Information System (INIS)

Sessions, J.

1999-01-01

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

Development of the web-based site investigation flow diagram in repository development program

International Nuclear Information System (INIS)

Yamamoto, Shuichi; Yoshimura, Kimitaka; Ohuchi, Jin; Tsuboya, Takao; Ando, Kenichi

2005-01-01

In siting a repository for high level radioactive wastes (HLW), it is essential for consensus building intelligibly and visually present why and how the area is selected as a suitable site. However 'information asymmetry' exists especially between society and an implementation body because various types of investigation, analysis and assessment are implemented in site characterization on the basis of a wide variety of advanced science and technology. Communication between experts (e.g. surveyors and modelers) is also important for efficient and reliable site investigation/ characterization. The Web-based Site Investigation Flow Diagram (SIFD) has been developed as a tool for information sharing among stake holders and society-jointed decision making. To test applicability of the SIFD, virtual site characterization ('dry run') is performed using the existing site investigation data. It is concluded that the web-based SIFD enhance traceability and transparency of the site investigation/ characterization, and therefore it would be a powerful communication tool among experts for efficient and reliable site investigation/characterization and among stake holders for consensus building

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

International Nuclear Information System (INIS)

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

2012-01-01

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

Coupling diffusion and high-pH precipitation/dissolution in the near field of a HLW repository in clay by means of reactive solute transport models

Science.gov (United States)

Samper, J.; Font, I.; Yang, C.; Montenegro, L.

2004-12-01

The reference concept for a HLW repository in clay in Spain includes a 75 cm thick bentonite buffer which surrounds canisters. A concrete sustainment 20 cm thick is foreseen between the bentonite buffer and the clay formation. The long term geochemical evolution of the near field is affected by a high-pH hyperalkaline plume induced by concrete. Numerical models of multicomponent reactive transport have been developped in order to quantify the evolution of the system over 1 Ma. Water flow is negligible once the bentonite buffer is saturated after about 20 years. Therefore, solute transport occurs mainly by diffusion. Models account for aqueous complexation, acid-base and redox reactions, cation exchange, and mineral dissolution precipitation in the bentonite, the concrete and the clay formation. Numerical results obtained witth CORE2D indicate that the high-pH plume causes significant changes in porewater chemistry both in the bentonite buffer and the clay formation. Porosity changes caused by mineral dissolution/precipitation are extremely important. Therefore, coupled modes of diffusion and reactive transport accounting for changes in porosity caused by mineral precipitation are required in order to obtain realistic predictions.

Waste Isolation Pilot Plant in situ experimental program for HLW

International Nuclear Information System (INIS)

Molecke, M.A.

1977-01-01

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

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

International Nuclear Information System (INIS)

Yoshida, Hidekazu; Yamamoto, Koushi; Amano, Yuki

2005-01-01

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

Elicitation and use of expert judgment in performance assessment for high-level radioactive waste repositories

International Nuclear Information System (INIS)

Bonano, E.J.; Hora, S.C.; Keeney, R.L.; von Winterfeldt, D.

1990-05-01

This report presents the concept of formalizing the elicitation and use of expert judgment in the performance assessment of high-level radioactive waste (HLW) repositories in deep geologic formations. The report begins with a discussion of characteristics (advantages and disadvantages) of formalizing expert judgment examples of previous uses of expert judgment in radioactive waste programs, criteria that can assist in deciding when to formalize expert judgment, and the relationship of formal use of expert judgment to data collection and modeling. The current state of the art with respect to the elicitation, use, and communication of formal expert judgment is presented. The report concludes with a discussion on potential applications of formal expert judgment in performance assessment of HLW repositories. 93 refs

Integrated safety case development for deep geological repositories

International Nuclear Information System (INIS)

Kawamura, Hideki; McKinley, Ian G.

2008-01-01

The paper will illustrate an 'integrated safety case', which involves combining both pre-closure and post-closure safety arguments from the point of view of a repository implementer, who must also ensure that projects are practical, acceptable and economic. The post-closure safety case is based on the performance of a number of barriers, which are established during construction, operation and closure. Such barriers must be confirmed using quality assured methods, supported, as required, by inspection and monitoring. The requirement for integrated assessment means that even the final process to end institutional control and transfer any liabilities from the implementer needs to be considered at present, even though this will undoubtedly be refined and tailored to the site characteristics over the many decades that will pass before this occurs. To illustrate the practical application of this approach, assessment of variants for remote-handled emplacement of the EBS for disposal of HLW in Japan will be discussed. (author)

Comparing technical concepts for disposal of Belgian vitrified HLW

International Nuclear Information System (INIS)

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

2004-01-01

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

Coupled thermo-hydro-mechanical analysis for the conceptual repository of high-level radioactive waste in China

International Nuclear Information System (INIS)

Lin, Y.M.; Wang, J.; Ke, D.; Cai, M.F.

2010-01-01

In order to safely dispose of the high-level radioactive waste (HLW), RD guide of HLW disposal was published in February 2006 in China. The spent fuel from nuclear power plants will be reprocessed first, followed by verification and final disposal. A conceptual repository 3D configuration comprises a single vertical borehole in a continuous and homogeneous hard rock, containing a canister surrounded by an over-pack and a bentonite layer, and the backfilled upper portion of the gallery using FLAC3D. To take into account in situ stress, geothermal gradient and groundwater of Beishan area, thermal relief of HLW and swelling pressure of buffer/backfill material made by GMZ01 bentonite, the TM, HM and THM evolution of the whole configuration is simulated over a period of 100 years. The results demonstrate that temperature is hardly affected by the couplings. In contrast, the influence of the couplings on the mechanical stresses is considerable. The repository has long-term stability in fully THM coupling action condition. (authors)

Modelling of gas generation in deep geological repositories after closure

International Nuclear Information System (INIS)

Poller, A.; Mayer, G.; Darcis M; Smith, P.

2016-12-01

In deep geological repositories for radioactive waste, significant quantities of gases will be generated in the long term as a result of various processes, notably the anaerobic corrosion of metals and the degradation of organic materials. Therefore, the impact of gas production on post-closure safety of the repositories needs to be assessed as part of a safety case. The present report provides a comprehensive description of the quantitative modelling of gas generation and associated water consumption during the post-closure phase of deep geological repositories in Opalinus Clay based on current scientific knowledge and on current preliminary repository designs. This includes a presentation of the modelling basis, namely the conceptual and mathematical models, the input data used, the computer tools developed, the relevant uncertainties and principal programme / design options, as well as the derivation, analysis and discussion of specific assessment cases. The modelling is carried out separately for the two main sources of gas, which are the emplaced waste including the disposal containers; and the construction materials. The contribution of construction materials to gas generation rates in emplacement tunnels for spent fuel (SF) and vitrified high-level waste (HLW) is significant during several thousand years after closure. In the long term, however, the corrosion of the disposal canisters, which are in the reference case assumed to be fabricated of carbon steel, accounts for the vast majority of the total gas produced in these tunnels. The contribution of construction materials in emplacement caverns for long lived intermediate-level waste (ILW) and low- and intermediate-level waste (L/ILW) to gas generation is generally small. In ILW emplacement caverns, gas generation is generally dominated by hydrogen generation from the corrosion of cast iron Mosaik-II waste containers for PWR internals and from the corrosion of aluminium in operational waste from the

Modelling of gas generation in deep geological repositories after closure

Energy Technology Data Exchange (ETDEWEB)

Poller, A. [National Cooperative for the Disposal of Radioactive Waste (NAGRA), Wettingen (Switzerland); Mayer, G.; Darcis M [AF-Consult Switzerland Ltd, Baden-Dättwil, (Switzerland); Smith, P. [Safety Assessment Management Ltd, Henley-On-Thames, Oxfordshire (United Kingdom)

2016-12-15

In deep geological repositories for radioactive waste, significant quantities of gases will be generated in the long term as a result of various processes, notably the anaerobic corrosion of metals and the degradation of organic materials. Therefore, the impact of gas production on post-closure safety of the repositories needs to be assessed as part of a safety case. The present report provides a comprehensive description of the quantitative modelling of gas generation and associated water consumption during the post-closure phase of deep geological repositories in Opalinus Clay based on current scientific knowledge and on current preliminary repository designs. This includes a presentation of the modelling basis, namely the conceptual and mathematical models, the input data used, the computer tools developed, the relevant uncertainties and principal programme / design options, as well as the derivation, analysis and discussion of specific assessment cases. The modelling is carried out separately for the two main sources of gas, which are the emplaced waste including the disposal containers; and the construction materials. The contribution of construction materials to gas generation rates in emplacement tunnels for spent fuel (SF) and vitrified high-level waste (HLW) is significant during several thousand years after closure. In the long term, however, the corrosion of the disposal canisters, which are in the reference case assumed to be fabricated of carbon steel, accounts for the vast majority of the total gas produced in these tunnels. The contribution of construction materials in emplacement caverns for long lived intermediate-level waste (ILW) and low- and intermediate-level waste (L/ILW) to gas generation is generally small. In ILW emplacement caverns, gas generation is generally dominated by hydrogen generation from the corrosion of cast iron Mosaik-II waste containers for PWR internals and from the corrosion of aluminium in operational waste from the

Sample performance assessment of a high-level radioactive waste repository: sensitivity analysis

International Nuclear Information System (INIS)

Tkaczyk, A.

2001-01-01

The Yucca Mountain Project (YMP) is the USA's first attempt at long-term storage of High-Level Radioactive Waste (HLW). In theory, the reasoning for such a repository seems sound. In practice, there are many scenarios and cases to be considered while putting such a project into effect. Since a goal of YMP is to minimize dangers associated with long-term storage of HLW, it is important to estimate the dose rate to which current and future generations will be subjected. The lifetime of the repository is simulated to indicate the radiation dose rate to the maximally exposed individual; it is assumed that if the maximally exposed individual would not be harmed by the annual dose, the remaining population will be at even smaller risk. The determination of what levels of exposure can be deemed harmless is a concern, and the results from the simulations as compared against various regulations are discussed. (author)

Sample performance assessment of a high-level radioactive waste repository: sensitivity analysis

Energy Technology Data Exchange (ETDEWEB)

Tkaczyk, A. [Iowa State Univ. of Science and Technology, Ames, IA (United States). Dept. of Mechanical Engineering

2001-07-01

The Yucca Mountain Project (YMP) is the USA's first attempt at long-term storage of High-Level Radioactive Waste (HLW). In theory, the reasoning for such a repository seems sound. In practice, there are many scenarios and cases to be considered while putting such a project into effect. Since a goal of YMP is to minimize dangers associated with long-term storage of HLW, it is important to estimate the dose rate to which current and future generations will be subjected. The lifetime of the repository is simulated to indicate the radiation dose rate to the maximally exposed individual; it is assumed that if the maximally exposed individual would not be harmed by the annual dose, the remaining population will be at even smaller risk. The determination of what levels of exposure can be deemed harmless is a concern, and the results from the simulations as compared against various regulations are discussed. (author)

Safety analysis of a sub-seabed repository of HLW

International Nuclear Information System (INIS)

Karpf, A.D.

1989-01-01

This national safety study of a repository in the Atlantic Ocean uses as much input from German research work as possible. It is based on a computer code IMPONADOR which can be run on a personal computer. It also introduces a new way of presenting information about radionuclide confinement to individual compartments to facilitate the understanding of release processes. The work is part of the international cooperation within the Seabed Working Group of the OECD/NEA

HLW Canister and Can-In-Canister Drop Calculation

International Nuclear Information System (INIS)

H. Marr

1999-01-01

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

Nuclear waste repository in basalt: preconceptual design guidelines

International Nuclear Information System (INIS)

1979-06-01

The development of the basalt waste isolation program parallels the growing need for permanent, environmentally safe, and secure means to store nuclear wastes. The repository will be located within the Columbia Plateau basalt formations where these ends can be met and radiological waste can be stored. These wastes will be stored such that the wastes may be retrieved from storage for a period after placement. After the retrieval period, the storage locations will be prepared for terminal storage. The terminal storage requirements will include decommissioning provisions. The facility boundaries will encompass no more than several square miles of land which will be above a subsurface area where the geologic makeup is primarily deep basaltic rock. The repository will receive, from an encapsulation site(s), nuclear waste in the form of canisters (not more than 18.5 feet x 16 inches in diameter) and containers (55-gallon drums). Canisters will contain spent fuel (after an interim 5-year storage period), solidified high-level wastes (HLW), or intermediate-level wastes (ILW). The containers (drums) will package the low-level transuranic wastes (LL-TRU). The storage capacity of the repository will be expanded in a time-phased program which will require that subsurface development (repository expansion) be conducted concurrently with waste storage operations. The repository will be designed to store the nuclear waste generated within the predictable future and to allow for reasonable expansion. The development and assurance of safe waste isolation is of paramount importance. All activities will be dedicated to the protection of public health and the environment. The repository will be licensed by the US Nuclear Regulatory Commission (NRC). Extensive efforts will be made to assure selection of a suitable site which will provide adequate isolation

Nuclear waste repository in basalt: preconceptual design guidelines

Energy Technology Data Exchange (ETDEWEB)

1979-06-01

The development of the basalt waste isolation program parallels the growing need for permanent, environmentally safe, and secure means to store nuclear wastes. The repository will be located within the Columbia Plateau basalt formations where these ends can be met and radiological waste can be stored. These wastes will be stored such that the wastes may be retrieved from storage for a period after placement. After the retrieval period, the storage locations will be prepared for terminal storage. The terminal storage requirements will include decommissioning provisions. The facility boundaries will encompass no more than several square miles of land which will be above a subsurface area where the geologic makeup is primarily deep basaltic rock. The repository will receive, from an encapsulation site(s), nuclear waste in the form of canisters (not more than 18.5 feet x 16 inches in diameter) and containers (55-gallon drums). Canisters will contain spent fuel (after an interim 5-year storage period), solidified high-level wastes (HLW), or intermediate-level wastes (ILW). The containers (drums) will package the low-level transuranic wastes (LL-TRU). The storage capacity of the repository will be expanded in a time-phased program which will require that subsurface development (repository expansion) be conducted concurrently with waste storage operations. The repository will be designed to store the nuclear waste generated within the predictable future and to allow for reasonable expansion. The development and assurance of safe waste isolation is of paramount importance. All activities will be dedicated to the protection of public health and the environment. The repository will be licensed by the US Nuclear Regulatory Commission (NRC). Extensive efforts will be made to assure selection of a suitable site which will provide adequate isolation.

Database and Interim Glass Property Models for Hanford HLW Glasses

International Nuclear Information System (INIS)

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

2001-01-01

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

The Nagra Research, Development and Demonstration (RD and D) plan for the disposal of radioactive waste in Switzerland

International Nuclear Information System (INIS)

2009-11-01

NAGRA's mission is to develop safe geological repositories for all radioactive wastes arising in Switzerland. Two types of repositories are foreseen, one for low and intermediate level waste (L/ILW) and one for spent fuel, vitrified high level waste and long-lived ILW (SF/HLW/ILW). Repository implementation involves a stepwise process that takes several decades, which is presented in the RD and D (research, development and demonstration) plan). The main objective of the RD and D Plan is to establish the purpose, scope, nature and timing of future RD and D activities. Chapter 1 presents the overall objectives of the report and a brief history of the steps leading to the present situation. NAGRA had shown in the project 'Entsorgungsnachweis' (disposal feasibility) that safe disposal of SF/HLW/ILW in Switzerland is technically feasible. Based on this result, the Federal Government initiated the Sectoral Plan for Geological Repositories. This Plan provides a framework within which specific objectives must be met for selecting suitable sites for disposal of both L/ILW and SF/HLW/ILW for which general licence applications are to be made. Stage 1 of the Sectoral Plan results in the selection of geologically suitable siting regions. Chapter 2 presents the overall planning premises for implementation of repositories for L/ ILW and SF/HLW/ILW including the assumed schedule, the waste types and quantities and the safety strategy for the repositories. The time plan includes the general licence procedure, construction and operation of Underground Research Laboratories at the sites, construction licence procedure and the operating licence procedure. It is expected that emplacement of L/ILW could begin in about 2035, whereas SF/HLW emplacement would begin about 2050. Waste emplacement is followed by a monitoring period (planning assumption of 50 years), at the end of which an application would be made for closure. The various waste types and quantities produced by nuclear power

Cooling and cracking of technical HLW glass products

International Nuclear Information System (INIS)

Kienzler, B.

1989-01-01

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

Evaluation of site-generated radioactive waste treatment and disposal methods for the Yucca Mountain repository

International Nuclear Information System (INIS)

Subramanian, C.V.; Jardine, L.J.

1989-01-01

This study identifies the sources of radioactive wastes that may be generated at the proposed high-level waste (HLW) repository at Yucca Mountain, NV, estimates the waste quantities and characteristics, compares technologies available for waste treatment and disposal, and develops recommended concepts for site-generated waste treatment and disposal. The scope of this study is limited to operations during the emplacement phase, in which 70,000 MTU of high-level waste will be received and emplaced at the proposed repository. The evaluations consider all radioactive wastes generated during normal operations in surface and underground facilities. Wastes generated as a result of accidents are not addressed; accidents that could result in large quantities of radioactive waste are expected to occur very infrequently and temporary, portable systems could be used for any necessary cleanup. The results of this study can be used to develop more definitive plans for managing the site-generated wastes and as a basis for the design of associated facilities at the proposed repository

R and D on HLW Partitioning in Russia

International Nuclear Information System (INIS)

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

2015-01-01

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

Impacts of new developments in partitioning and transmutation on the disposal of high-level nuclear waste in a mined geologic repository

International Nuclear Information System (INIS)

Ramspott, L.D.; Jor-Shan Choi; Halsey, W.; Pasternak, A.; Cotton, T.; Burns, J.; McCabe, A.; Colglazier, W.; Lee, W.W.L.

1992-03-01

During the 1970s, the United States and other countries thoroughly evaluated the options for the safe and final disposal of high-level radioactive wastes (HLW). The worldwide scientific community concluded that deep geologic disposal was clearly the most technically feasible alternative. They also ranked the partitioning and transmutation (P-T) of radionuclides among the least favored options. A 1982 report by the International Atomic Energy Agency summarized the key reasons for that ranking: ''Since the long-term hazards are already low, there is little incentive to reduce them further by P-T. Indeed the incremental costs of introducing P-T appear to be unduly high in relation to the prospective benefits.'' Recently, the delays encountered by the US geologic disposal program for HLW, along with advanced in the development of P-T concepts, have led some to propose P-T as a means of reducing the long-term risks from the radioactive wastes that require disposal and thus making it easier to site, license, and build a geologic repository. This study examines and evaluates the effects that introducing P-T would have on the US geologic disposal program

How many geologic repositories will be needed

International Nuclear Information System (INIS)

Evans, T.J.; Halstead, R.J.

1987-01-01

DOE's postponement of site-specific work on the second repository program had rekindled debate over the number of geologic repositories needed for disposal of high level radioactive waste. The multiple repository approach grew out of the March, 1979 IRG report, which recommended co-disposal of civilian and defense HLW in a system of regional repositories. The multiple repository approach was adopted by DOE, and incorporated in the Nuclear Waste Policy Act passed by Congress in December, 1982. Since the late 1970's, the slower than anticipated growth of the nuclear power industry has substantially reduced earlier estimates of the amount of civilian spent fuel which will require geologic disposal. Reactors currently in operation (78.5 GWe) and reactors in the construction pipeline (28 GWe) are expected to discharge about 103,200 MTU of spent fuel by the year 2036, assuming no increase in fuel burnup rate. By the year 2020, defense high level radioactive wastes equivalent to as much as 27,000 MTU could require geologic disposal. Small amounts of high level waste from other sources will also require geologic disposal. Total disposal requirements appear to be less than 140,000 MTU. The five sites nominated for the first repository, as well as hypothetical sites in granite, the host rock under primary consideration for the second repository, all appear capable of accommodating up to 140,000 MTU

Thermal analysis of the vertical disposal for HLW

International Nuclear Information System (INIS)

Zhao Honggang; Wang Ju; Liu Yuemiao; Su Rui

2013-01-01

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

Investigation of two-phase flow phenomena associated with corrosion in an SF/HLW repository in Opalinus Clay, Switzerland

International Nuclear Information System (INIS)

Senger, R.; Marschall, P.; Finsterle, S.

2008-01-01

Gas generation from corrosion of the waste canisters and gas accumulation in the backfilled emplacement tunnels is a key issue in the assessment of long-term radiological safety of the proposed repository for spent fuel and high-level waste (SF/HLW) sited in the Opalinus Clay formation of Northern Switzerland. Previous modeling studies indicated a significant pressure buildup in the backfilled emplacement tunnels for those sensitivity runs, where corrosion rates were high and the permeability of the Opalinus Clay was very low. As an extension to those studies, a refined process model of the canister corrosion phenomena has been developed, which accounts not only for the gas generation but also for the water consumption associated with the chemical reaction of corrosion of steel under anaerobic conditions. The simulations with the new process model indicate, that with increasing corrosion rates and decreasing host-rock permeability, pressure buildup increased, as expected. However, the simulations taking into account water consumption show that the pressure buildup is reduced compared to the simulation considering only gas generation. The pressure reduction is enhanced for lower permeability of the Opalinus Clay and for higher corrosion rates, which correspond to higher gas generations rates and higher water consumption rates. Moreover, the simulated two-phase flow patterns in the engineered barrier system (EBS) and surrounding Opalinus Clay show important differences at late time of the gas production phase as the generated gas continues to migrate outward into the surrounding host rock. For the case without water consumption, the water flow indicates overall downward flow due to a change in the overall density of the gas-fluid mixture from that based on the initially prescribed hydrostatic pressure gradient. For the case with water consumption, water flow converges toward the waste canister at a rate corresponding to the water consumption rate associated with the

Retrieval effects on ventilation and cooling requirements for a nuclear waste repository

International Nuclear Information System (INIS)

Hambley, D.F.

1985-01-01

The Nuclear Waste Policy Act of 1982 (Public Law 97-425) and the regulations promulgated in Title 10, Part 60 of the Code of Federal Regulations (10CFR60) by the US Nuclear Regulatory Commission (NRC) for an underground repository for spent fuel and high level nuclear waste (HLW) require that it is possible to retrieve waste, for whatever reason, from such a facility for a period of 50 years from initial storage or until the completion of the performance confirmation period, whichever comes first. This paper considers the effects that the retrievability option mandates on ventilation and cooling systems required for normal repository operations. An example is given for a hypothetical repository in salt. 18 refs., 1 tab

NRC iterative performance assessment Phase 2. Development of capabilities for review of a performance assessment for a high-level waste repository

International Nuclear Information System (INIS)

Wescott, R.G.; Lee, M.P.; Eisenberg, N.A.; Baca, R.G.

1995-10-01

In order to better review a potential license application to construct and operate a geologic repository for spent nuclear fuel and high-level radioactive waste (HLW), the Nuclear Regulatory Commission staff (and its contractor) has expanded and improved its capability to conduct performance assessments. This report documents the demonstration of the second phase of this capability. The demonstration made use of the scenario selection procedure developed by Sandia National Laboratories to provide a set of scenarios, with corresponding probabilities, for use in the consequence analysis of a potential HLW disposal site in unsaturated tuff. Models of release of radionuclides from the waste form and transport in ground water, air and by direct pathways provided preliminary estimates of releases to the accessible environment for a 10,000 year period. The input values of parameters necessary for the consequence models were sampled numerous times using Latin Hypercube Sampling from assumed probability distributions. The results from the consequence models were then used to generate Complementary Cumulative Distribution Functions (CCDFs) for either release to the accessible environment or effective dose equivalents to a target population. CCDFs were calculated for probabilistically significant combinations (scenarios) of four disruptive events; drilling, pluvial climate, seismicity and magmatism. Sensitivity and uncertainty analyses of the calculated releases and effective dose equivalents were also used to determine the importance of the parameters. Because of the preliminary nature of the analysis and the lack of an adequate data base, the results and conclusions presented in this report should be carefully interpreted. They should not be misconstrued to represent the actual performance of the proposed Yucca Mountain repository nor serve as an endorsement of the methods used

A comparison of risks due to HLW and SURF repositories in bedded salt

International Nuclear Information System (INIS)

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

1983-01-01

A methodology was developed to analyze risks from geologic disposal of nuclear wastes. It was applied to estimate the risk due to spent fuel and high-level waste hypothetical repositories in bedded salt. A number of disruptive scenarios were analyzed for each waste type. A comparison between the spent fuel and high-level waste results is presented. The methodology enables one to identify important radionuclides, parameters, and scenarios in terms of their relative contribution to the overall risk or compliance with the proposed EPA Standard

Generic Repository Concepts and Thermal Analysis for Advanced Fuel Cycles - 12477

Energy Technology Data Exchange (ETDEWEB)

Hardin, Ernest [Sandia National Laboratories, P.O. Box 5800 MS 0736, Albuquerque, NM 87185 (United States); Blink, James [Lawrence Livermore National Laboratory, P.O. Box 808, Livermore, CA 94551-0808 (United States); Carter, Joe [Savannah River National Laboratory, Savannah River Site, Aiken, SC 29808 (United States); Fratoni, Massimiliano; Greenberg, Harris; Sutton, Mark [Lawrence Livermore National Laboratory (United States); Howard, Robert [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831 (United States)

2012-07-01

A geologic disposal concept for spent nuclear fuel (SNF) or high-level waste (HLW) consists of three components: waste inventory, geologic setting, and concept of operations. A set of reference geologic disposal concepts has been developed by the U.S. Department of Energy (DOE), Used Fuel Disposition campaign. Reference concepts are identified for crystalline rock, clay/shale, bedded salt, and deep borehole (crystalline basement) geologic settings. These were analyzed for waste inventory cases representing a range of waste types that could be produced by advanced nuclear fuel cycles. Concepts of operation consisting of emplacement mode, repository layout, and engineered barrier descriptions, were selected based on international progress. All of these disposal concepts are enclosed emplacement modes, whereby waste packages are in direct contact with encapsulating engineered or natural materials. Enclosed modes have less capacity to dissipate heat than open modes such as that proposed for a repository at Yucca Mountain. Thermal analysis has identified important relationships between waste package size and capacity, and the duration of surface decay storage needed to meet temperature limits for different disposal concepts. For the crystalline rock and clay/shale repository concepts, a waste package surface temperature limit of 100 deg. C was assumed to prevent changes in clay-based buffer material or clay-rich host rock. Surface decay storage of 50 to 100 years is needed for disposal of high-burnup LWR SNF in 4-PWR packages, or disposal of HLW glass from reprocessing LWR uranium oxide (UOX) fuel. High-level waste (HLW) from reprocessing of metal fuel used in a fast reactor could be disposed after decay storage of 50 years or less. For disposal in salt the rock thermal conductivity is significantly greater, and higher temperatures (200 deg. C) can be tolerated at the waste package surface. Decay storage of 10 years or less is needed for high-burnup LWR SNF in 4-PWR

Review of scenario selection approaches for performance assessment of high-level waste repositories and related issues

International Nuclear Information System (INIS)

Banano, E.J.; Baca, R.G.

1995-08-01

The selection of scenarios representing plausible realizations of the future conditions-with associated probabilities of occurrence-that can affect the long-term performance of a high-level radioactive waste (HLW) repository is the commonly used method for treating the uncertainty in the prediction of the future states of the system. This method, conventionally referred to as the ''scenario approach,'' while common is not the only method to deal with this uncertainty; other method ''ch as the environmental simulation approach (ESA), have also been proposed. Two of the difficulties with the scenario approach are the lack of uniqueness in the definition of the term ''scenario'' and the lack of uniqueness in the approach to formulate scenarios, which relies considerably on subjective judgments. Consequently, it is difficult to assure that a complete and unique set of scenarios can be defined for use in a performance assessment. Because scenarios are key to the determination of the long-term performance of the repository system, this lack of uniqueness can present a considerable challenge when attempting to reconcile the set of scenarios, and their level of detail, obtained using different approaches, particularly among proponents and regulators of a HLW repository

Review of scenario selection approaches for performance assessment of high-level waste repositories and related issues.

Energy Technology Data Exchange (ETDEWEB)

Banano, E.J. [Beta Corporation International, Albuquerque, NM (United States); Baca, R.G. [Southwest Research Inst., San Antonio, TX (United States). Center for Nuclear Waste Regulatory Analyses

1995-08-01

The selection of scenarios representing plausible realizations of the future conditions-with associated probabilities of occurrence-that can affect the long-term performance of a high-level radioactive waste (HLW) repository is the commonly used method for treating the uncertainty in the prediction of the future states of the system. This method, conventionally referred to as the ``scenario approach,`` while common is not the only method to deal with this uncertainty; other method ``ch as the environmental simulation approach (ESA), have also been proposed. Two of the difficulties with the scenario approach are the lack of uniqueness in the definition of the term ``scenario`` and the lack of uniqueness in the approach to formulate scenarios, which relies considerably on subjective judgments. Consequently, it is difficult to assure that a complete and unique set of scenarios can be defined for use in a performance assessment. Because scenarios are key to the determination of the long-term performance of the repository system, this lack of uniqueness can present a considerable challenge when attempting to reconcile the set of scenarios, and their level of detail, obtained using different approaches, particularly among proponents and regulators of a HLW repository.

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

International Nuclear Information System (INIS)

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

2000-01-01

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

Influence of Glass Property Restrictions on Hanford HLW Glass Volume

International Nuclear Information System (INIS)

Kim, Dong-Sang; Vienna, John D.

2001-01-01

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

Reduction on high level radioactive waste volume and geological repository footprint with high burn-up and high thermal efficiency of HTGR

Energy Technology Data Exchange (ETDEWEB)

Fukaya, Yuji, E-mail: fukaya.yuji@jaea.go.jp; Nishihara, Tetsuo

2016-10-15

Highlights: • We evaluate the number of canisters and its footprint for HTGR. • We proposed new waste loading method for direct disposal of HTGR. • HTGR can significantly reduce HLW volume compared with LWR. - Abstract: Reduction on volume of High Level radioactive Waste (HLW) and footprint in a geological repository due to high burn-up and high thermal efficiency of High Temperature Gas-cooled Reactor (HTGR) has been investigated. A helium-cooled and graphite-moderated commercial HTGR was designed as a Gas Turbine High Temperature Reactor (GTHTR300), and that has particular features such as significantly high burn-up of approximately 120 GWd/t, high thermal efficiency around 50%, and pin-in-block type fuel. The pin-in-block type fuel was employed to reduce processed graphite volume in reprocessing. By applying the feature, effective waste loading method for direct disposal is proposed in this study. By taking into account these feature, the number of HLW canister generations and its repository footprint are evaluated by burn-up fuel composition, thermal calculation and criticality calculation in repository. As a result, it is found that the number of canisters and its repository footprint per electricity generation can be reduced by 60% compared with Light Water Reactor (LWR) representative case for direct disposal because of the higher burn-up, higher thermal efficiency, less TRU generation, and effective waste loading proposed in this study for HTGR. But, the reduced ratios change to 20% and 50% if the long term durability of LWR canister is guaranteed. For disposal with reprocessing, the number of canisters and its repository footprint per electricity generation can be reduced by 30% compared with LWR because of the 30% higher thermal efficiency of HTGR.

The role of natural analogs in the repository licensing process

International Nuclear Information System (INIS)

Murphy, W.M.

1995-01-01

The concept of a permanent geologic repository for high-level nuclear waste (NLW) is implicitly based on analogy to natural systems that have been stable for millions or billions of years. The time of radioactive and chemical toxicity of HLW exceeds the duration of human civilization, and it is impossible to demonstrate the accuracy of predictions of the behavior of engineered or social systems over such long periods

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

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

Technetium Chemistry in HLW

International Nuclear Information System (INIS)

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

2005-01-01

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

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

International Nuclear Information System (INIS)

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

2008-01-01

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

HIGH ALUMINUM HLW GLASSES FOR HANFORD'S WTP

International Nuclear Information System (INIS)

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

2009-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-06-15

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

Field instrumentation and testing needs for a high level waste repository

International Nuclear Information System (INIS)

Marti, J.; Maini, T.

1981-03-01

A review has been conducted of the testing and measurement needs posed by a deep geologic High Level Waste (HLW) repository in crystalline or argillaceous rocks. Siting, design, construction, operation and decommissioning of the repository have been covered, together with the planning of a Test and Demonstration Facility. Instruments and methods available have been critically assessed in their ability to fulfil the aforementioned testing and monitoring programmes. Special attention has been paid to the relation of measurements to the data needs and to the tests likely to generate such data. This assessment has concentrated on measurements of absolute rock stresses, monitoring of changes in rock stress, evaluation of the rock mass deformability, measurement of relative displacements and determination of the hydrogeologic parameters of the rock mass. Other measurements have been studied with a lesser degree of attention. The overall conclusion is that, from the instrumentation and testing points of view, present plans for a test and demonstration facility in the early nineties and a repository soon after 2000 are indeed feasible. Specific conclusions on the state-of-the-art and development needs are presented in the report. (author)

Integrating repositories with fuel cycles: The airport authority model

Energy Technology Data Exchange (ETDEWEB)

Forsberg, C. [Massachusetts Inst. of Technology, 77 Massachusetts Avenue, Cambridge, MA 02139-4307 (United States)

2012-07-01

The organization of the fuel cycle is a legacy of World War II and the cold war. Fuel cycle facilities were developed and deployed without consideration of the waste management implications. This led to the fuel cycle model of a geological repository site with a single owner, a single function (disposal), and no other facilities on site. Recent studies indicate large economic, safety, repository performance, nonproliferation, and institutional incentives to collocate and integrate all back-end facilities. Site functions could include geological disposal of spent nuclear fuel (SNF) with the option for future retrievability, disposal of other wastes, reprocessing with fuel fabrication, radioisotope production, other facilities that generate significant radioactive wastes, SNF inspection (navy and commercial), and related services such as SNF safeguards equipment testing and training. This implies a site with multiple facilities with different owners sharing some facilities and using common facilities - the repository and SNF receiving. This requires a different repository site institutional structure. We propose development of repository site authorities modeled after airport authorities. Airport authorities manage airports with government-owned runways, collocated or shared public and private airline terminals, commercial and federal military facilities, aircraft maintenance bases, and related operations - all enabled and benefiting the high-value runway asset and access to it via taxi ways. With a repository site authority the high value asset is the repository. The SNF and HLW receiving and storage facilities (equivalent to the airport terminal) serve the repository, any future reprocessing plants, and others with needs for access to SNF and other wastes. Non-public special-built roadways and on-site rail lines (equivalent to taxi ways) connect facilities. Airport authorities are typically chartered by state governments and managed by commissions with members

Integrating repositories with fuel cycles: The airport authority model

International Nuclear Information System (INIS)

Forsberg, C.

2012-01-01

The organization of the fuel cycle is a legacy of World War II and the cold war. Fuel cycle facilities were developed and deployed without consideration of the waste management implications. This led to the fuel cycle model of a geological repository site with a single owner, a single function (disposal), and no other facilities on site. Recent studies indicate large economic, safety, repository performance, nonproliferation, and institutional incentives to collocate and integrate all back-end facilities. Site functions could include geological disposal of spent nuclear fuel (SNF) with the option for future retrievability, disposal of other wastes, reprocessing with fuel fabrication, radioisotope production, other facilities that generate significant radioactive wastes, SNF inspection (navy and commercial), and related services such as SNF safeguards equipment testing and training. This implies a site with multiple facilities with different owners sharing some facilities and using common facilities - the repository and SNF receiving. This requires a different repository site institutional structure. We propose development of repository site authorities modeled after airport authorities. Airport authorities manage airports with government-owned runways, collocated or shared public and private airline terminals, commercial and federal military facilities, aircraft maintenance bases, and related operations - all enabled and benefiting the high-value runway asset and access to it via taxi ways. With a repository site authority the high value asset is the repository. The SNF and HLW receiving and storage facilities (equivalent to the airport terminal) serve the repository, any future reprocessing plants, and others with needs for access to SNF and other wastes. Non-public special-built roadways and on-site rail lines (equivalent to taxi ways) connect facilities. Airport authorities are typically chartered by state governments and managed by commissions with members

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.

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

Development of the Korean Reference Vertical Disposal System Concept for Spent Fuels

International Nuclear Information System (INIS)

Lee, J.Y.; Cho, D.K.; Kim, S.G.; Choi, H.J.; Choi, J.W.; Hahn, P.S.

2006-01-01

The development of a deep geologic disposal system for the spent fuel from nuclear power plants has been carried out since this program was launched at 1997 in Korea. In ' this paper, a pre-conceptual design of the Korean Reference HLW Vertical disposal System (KRS-V1) is presented. Though no site for the underground repository has yet been specified in Korea, a generic site with granitic rock is considered for reference HLW repository design. Depth of the repository is assumed to be 500 meters. The repository consists of the disposal area, technical rooms with four shafts to connect them to the ground level in the controlled area and technical rooms with an access tunnel and three shafts to connect them to the ground level in the uncontrolled area. Disposal area consists of disposal tunnels, panel tunnels and a central tunnel. The repository will be excavated, operated and backfilled in several phases including an Underground Research Laboratory (URL) phase. The result of this preliminary conceptual design will be used for an evaluation of the feasibility, analyses of the long term safety, information for public communication and a cost estimation etc. (authors)

Hydrothermal modeling for the efficient design of thermal loading in a nuclear waste repository

International Nuclear Information System (INIS)

Cho, Won-Jin; Kim, Jin-Seop; Choi, Heui-Joo

2014-01-01

Highlights: • Three-dimensional hydrothermal modeling for HLW repository is performed. • The model reduces the peak temperature in the repository by about 10 °C. • Decreasing the tunnel distance is more efficient to improve the disposal density. • The EDZ surrounding the deposition hole increases the peak temperature. • The peak temperature for the double-layer repository remains below the limit. - Abstract: The thermal analysis of a geological repository for nuclear waste using the three-dimensional hydrothermal model is performed. The hydrothermal model reduces the maximum peak temperature in the repository by about 10 °C compared to the heat conduction model with constant thermal conductivities. Decreasing the tunnel distance is more efficient than decreasing the deposition hole spacing to improve the disposal density for a given thermal load. The annular excavation damaged zone surrounding the deposition hole has a considerable effect on the peak temperature. The possibility of double-layer repository is analyzed from the viewpoint of the thermal constraints of the repository. The maximum peak temperature for the double-layer repository is slightly higher than that for the single-layer repository, but remains below the temperature limit

Efficiency analyses of the CANDU spent fuel repository using modified disposal canisters for a deep geological disposal system design

International Nuclear Information System (INIS)

Lee, J.Y.; Cho, D.K.; Lee, M.S.; Kook, D.H.; Choi, H.J.; Choi, J.W.; Wang, L.M.

2012-01-01

Highlights: ► A reference disposal concept for spent nuclear fuels in Korea has been reviewed. ► To enhance the disposal efficiency, alternative disposal concepts were developed. ► Thermal analyses for alternative disposal concepts were performed. ► From the result of the analyses, the disposal efficiency of the concepts was reviewed. ► The most effective concept was suggested. - Abstract: Deep geological disposal concept is considered to be the most preferable for isolating high-level radioactive waste (HLW), including nuclear spent fuels, from the biosphere in a safe manner. The purpose of deep geological disposal of HLW is to isolate radioactive waste and to inhibit its release of for a long time, so that its toxicity does not affect the human beings and the biosphere. One of the most important requirements of HLW repository design for a deep geological disposal system is to keep the buffer temperature below 100 °C in order to maintain the integrity of the engineered barrier system. In this study, a reference disposal concept for spent nuclear fuels in Korea has been reviewed, and based on this concept, efficient alternative concepts that consider modified CANDU spent fuels disposal canister, were developed. To meet the thermal requirement of the disposal system, the spacing of the disposal tunnels and that of the disposal pits for each alternative concept, were drawn following heat transfer analyses. From the result of the thermal analyses, the disposal efficiency of the alternative concepts was reviewed and the most effective concept suggested. The results of these analyses can be used for a deep geological repository design and detailed analyses, based on exact site characteristics data, will reduce the uncertainty of the results.

Long-term storage or disposal of HLW-dilemma

International Nuclear Information System (INIS)

Ninkovic, M. M.; Raicevic, J.

1995-01-01

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

Overly Honest Data Repository Development

Directory of Open Access Journals (Sweden)

Colleen Fallaw

2016-10-01

Full Text Available After a year of development, the library at the University of Illinois at Urbana-Champaign has launched a repository, called the Illinois Data Bank (https://databank.illinois.edu/, to provide Illinois researchers with a free, self-serve publishing platform that centralizes, preserves, and provides persistent and reliable access to Illinois research data. This article presents a holistic view of development by discussing our overarching technical, policy, and interface strategies. By openly presenting our design decisions, the rationales behind those decisions, and associated challenges this paper aims to contribute to the library community's work to develop repository services that meet growing data preservation and sharing needs.

Groundwater impacts of foreseeable human activities on a HLW repository

International Nuclear Information System (INIS)

Coleman, N.M.

1993-01-01

The Nuclear Regulatory Commission (NRC) staff has begun a program of Systematic Regulatory Analysis (SRA) to help ensure that all important technical issues related to the disposal of civilian, high-level nuclear wastes will be identified prior to the receipt of a license application. Large-scale groundwater withdrawals near a repository could have significant impacts on the groundwater flow system. Future large-scale withdrawals of groundwater could occur to support irrigation to growing population centers, such as Las Vegas. Various scenarios of groundwater withdrawals, along with other scenarios of future human activity, will need to be tested before evaluation of the Yucca Mountain site is complete

Staged Repository Development Programmes

International Nuclear Information System (INIS)

Isaacs, T

2003-01-01

Programs to manage and ultimately dispose of high-level radioactive wastes are unique from scientific and technological as well as socio-political aspects. From a scientific and technological perspective, high-level radioactive wastes remain potentially hazardous for geological time periods-many millennia-and scientific and technological programs must be put in place that result in a system that provides high confidence that the wastes will be isolated from the accessible environment for these many thousands of years. Of course, ''proof'' in the classical sense is not possible at the outset, since the performance of the system can only be known with assurance, if ever, after the waste has been emplaced for those geological time periods. Adding to this challenge, many uncertainties exist in both the natural and engineered systems that are intended to isolate the wastes, and some of the uncertainties will remain regardless of the time and expense in attempting to characterize the system and assess its performance. What was perhaps underappreciated in the early days of waste management and repository program development were the unique and intense reactions that the institutional, political, and public bodies would have to repository program development, particularly in programs attempting to identify and then select sites for characterization, design, licensing, and ultimate development. Reactions in most nations were strong, focused, unrelenting, and often successful in hindering, derailing, and even stopping national repository programs. The reasons for such reactions and the measures to successfully respond to them are still evolving and continue to be the focus of many national program and political leaders. Adaptive Staging suggests an approach to repository program development that reflects the unique challenges associated with the disposal of high-level radioactive waste. The step-wise, incremental, learn-as-you-go approach is intended to maximize the

Natural geochemical analogues of the near field of high-level nuclear waste repositories

International Nuclear Information System (INIS)

Apps, J.A.

1995-01-01

United States practice has been to design high-level nuclear waste (HLW) geological repositories with waste densities sufficiently high that repository temperatures surrounding the waste will exceed 100 degrees C and could reach 250 degrees C. Basalt and devitrified vitroclastic tuff are among the host rocks considered for waste emplacement. Near-field repository thermal behavior and chemical alteration in such rocks is expected to be similar to that observed in many geothermal systems. Therefore, the predictive modeling required for performance assessment studies of the near field could be validated and calibrated using geothermal systems as natural analogues. Examples are given which demonstrate the need for refinement of the thermodynamic databases used in geochemical modeling of near-field natural analogues and the extent to which present models can predict conditions in geothermal fields

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.'

Study of nuclear waste storage capacity at Yucca mountain repository

International Nuclear Information System (INIS)

Zhou Wei; Apted, M.; Kessler, J.H.

2008-01-01

The Yucca Mountain repository is applying license for storing 70000 MTHM nuclear waste including commercial spent nuclear fuel (CSNF) and defense high-level radioactive waste (HLW). The 70000 MTHM is a legal not the technical limit. To study the technical limit, the Electric Power Research Institute (EPRI) carried out a systematic study to explore the potential impact if the repository will accept more waste. This paper describes the model and results for evaluating the spent-fuel disposal capacity for a repository at Yucca Mountain from the thermal and hydrological point of view. Two proposed alternative repository designs are analyzed, both of which would fit into the currently well-characterized site and, therefore, not necessitating any additional site characterization at Yucca Mountain. The two- and three-dimensional models for coupled thermo-hydrological analysis extends from the surface to the water table, covering all the major and subgroup rock layers of the planned repository, as well as formations above and below the repository horizon. A dual-porosity and dual-permeability approach is used to model coupled heat and mass transfer through fracture formations. The waste package heating and ventilation are all assumed to follow those of the current design. The results show that the repository is able to accommodate three times the amount of spent fuel compared to the current design, without extra spatial expansion or exceeding current thermal and hydrological constraints. (authors)

R and D programme for HLW disposal in Japan

International Nuclear Information System (INIS)

Tsuboya, Takao

1997-01-01

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

Hydrogen modelling for vitrified wastes repository

International Nuclear Information System (INIS)

Voinis, S.; Breton, J.

1992-01-01

Safety assessments for High Level Wastes (HLW) have led ANDRA (Agence Nationale pour la gestion des Dechets Radioactifs) to study the occurrence of a gas production rate in a repository. This paper deals with the description of an analytical model used for the gas production rate assessment and brings us the first results. The geometry used is restrained to a single borehole associated with a drift in a crystalline formation. Different concepts were studied in this assessment. First results have been obtained. For example, in the case of a permeable plug, the saturation time of the borehole is about 300 years. 5 refs., 5 figs

Melter Throughput Enhancements for High-Iron HLW

Energy Technology Data Exchange (ETDEWEB)

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

2012-12-26

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

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)

Safety case for license application for a final repository: The French example

International Nuclear Information System (INIS)

Boissier, Fabrice; Voinis, Sylvie

2014-01-01

The reversible repository in a deep geological formation is the French reference solution for the long-term management of high-level and intermediate-level long-lived radioactive waste (HLW and ILW). Twenty years of R and D work and conceptual and basic studies since the first French Act of 1991 led, in particular, to a feasibility demonstration in 2005. According to the French Act on Radioactive Waste of 28 of June 2006, Andra shall design a reversible repository in order to apply for license in 2015. In response to this demand, Andra developed the industrial project known as 'Cigeo', a reversible geological disposal facility for HLW and ILW located in Meuse/Haute-Marne. Two years before applying for authorisation, Andra's project is now focusing on three main targets: developing Cigeo's industrial design, preparing the authorisation process through increased exchanges with stakeholders and the preparation of a safety case to support authorisation application. The latter draws on the previous safety cases of 2005 and 2009, which give a sound basis to assess Cigeo's safety, both for the operational and post-closure periods. In this new stage of the project, the challenging issues for the preparation of the safety case are the following: - to identify the various regulatory frameworks (nuclear and non-nuclear) and guides applicable to the facility; - to ensure that the industrial design complies in particular with the safety requirements as presented in the safety case and its supporting safety assessment; - to identify crucial inputs (R and D, tests,...) needed to support the authorisation application, in particular, to bring convincing arguments to assess the technical feasibility of the design and when appropriate its ability to meet the safety requirements; - to ensure that all the requirements from previous regulatory and peer reviews (national and international?) are taken into account. (authors)

NPP Krsko: LILW Repository or Long Term Storage

International Nuclear Information System (INIS)

Lokner, V.; Subasic, D.; Levanat, I.

2008-01-01

Construction of the facilities for LILW and SF management, as planned in Decommissioning and LILW and SF management program for NPP Krsko, would be a rather expensive and challenging project for such a small nuclear program. In order to accommodate waste arising from a single nuclear power plant, one LILW repository should be constructed before the end of the NPP operation, then one SF dry storage, and finally one geological repository. This requires relatively urgent identification within Slovenian/Croatian territory of three locations that meet the criteria for establishment of such facilities and are acceptable to the local communities. There are very few such potential locations. The siting process for the first of the three facilities is well under way in Slovenia, because the country wants to have its LILW repository in operation by the year 2013. In order to facilitate public acceptance, Slovenian government has introduced financial incentives to local communities for the repository construction and operation. These 'compensations for limited land use' may significantly increase the overall costs of disposal if the repository is in operation for a long period. In the recent years, however, a possibility of long term storage (LTS) is gaining an increased attention in the waste management community, and has already been introduced e.g. in the Netherlands. It is a particularly viable option for limited waste quantities. Disposal remains the final solution, but present technologies have made possible a relatively inexpensive storage up to about hundred years, which can accommodate LILW, HLW and SF from nuclear programs as well as research reactor waste and NORM. Such storage would be a safe and simple temporary solution, encompassing all immediate and near future waste management needs. In addition, it would increase flexibility and reduce financing requirements for the final waste disposal: providing additional time for reduction of radiation emission and heat

Current Status of Deep Geological Repository Development

International Nuclear Information System (INIS)

Budnitz, R J

2005-01-01

This talk provided an overview of the current status of deep-geological-repository development worldwide. Its principal observation is that a broad consensus exists internationally that deep-geological disposal is the only long-term solution for disposition of highly radioactive nuclear waste. Also, it is now clear that the institutional and political aspects are as important as the technical aspects in achieving overall progress. Different nations have taken different approaches to overall management of their highly radioactive wastes. Some have begun active programs to develop a deep repository for permanent disposal: the most active such programs are in the United States, Sweden, and Finland. Other countries (including France and Russia) are still deciding on whether to proceed quickly to develop such a repository, while still others (including the UK, China, Japan) have affirmatively decided to delay repository development for a long time, typically for a generation of two. In recent years, a major conclusion has been reached around the world that there is very high confidence that deep repositories can be built, operated, and closed safely and can meet whatever safety requirements are imposed by the regulatory agencies. This confidence, which has emerged in the last few years, is based on extensive work around the world in understanding how repositories behave, including both the engineering aspects and the natural-setting aspects, and how they interact together. The construction of repositories is now understood to be technically feasible, and no major barriers have been identified that would stand in the way of a successful project. Another major conclusion around the world is that the overall cost of a deep repository is not as high as some had predicted or feared. While the actual cost will not be known in detail until the costs are incurred, the general consensus is that the total life-cycle cost will not exceed a few percent of the value of the

Production, consumption and transport of gases in deep geological repositories according to the Swiss disposal concept

International Nuclear Information System (INIS)

Diomidis, N; Cloet, V.; Leupin, O.X.; Marschall, P.; Poller, A.; Stein, M.

2016-12-01

In a deep geological repository for radioactive waste, in absence of oxygen and in presence of water, corrosion of various metals and alloys will lead to the formation of hydrogen. If present, organic materials may slowly degrade and generate carbon dioxide, methane and other gaseous species. Depending on local conditions, gaseous species can be consumed by chemical reactions and by microbial activity. If the resulting rate of gas generation exceeds the rate of migration of dissolved gas molecules in the pores of the engineered barriers or the host rock, the solubility limit of the gas will eventually be exceeded and the formation of a discrete gas phase will occur. Gases could continue to accumulate until the pressure becomes sufficient to be released in gaseous form. This report deals with the evolution of gas-related processes that can influence the long-term behaviour and safety of low- and intermediate-level waste (L/ILW) and high-level waste (HLW) repositories in Opalinus Clay. The main aim is to present a synthesis of processes and phenomena related to repository-produced gases and to assess their influence on repository performance. A current overview of gas sources, reactions and interactions, generation, consumption, and transport is provided. Furthermore, current scientific understanding is used to define safety function indicators and criteria, which are employed to evaluate the potential influence of repository-generated gas on safety-relevant properties of engineered and natural barriers. The assessment of gas generation, consumption and transport is addressed separately for the HLW and the L/ILW deep geological repositories. The employed methodology, which is common for both repository types, consists of the description and quantification of the potential gas sources, which include the waste, barrier components such as disposal canisters and other gas-generating repository components, and of the processes and reactions leading to the generation or

Production, consumption and transport of gases in deep geological repositories according to the Swiss disposal concept

Energy Technology Data Exchange (ETDEWEB)

Diomidis, N; Cloet, V.; Leupin, O.X.; Marschall, P.; Poller, A.; Stein, M.

2016-12-15

In a deep geological repository for radioactive waste, in absence of oxygen and in presence of water, corrosion of various metals and alloys will lead to the formation of hydrogen. If present, organic materials may slowly degrade and generate carbon dioxide, methane and other gaseous species. Depending on local conditions, gaseous species can be consumed by chemical reactions and by microbial activity. If the resulting rate of gas generation exceeds the rate of migration of dissolved gas molecules in the pores of the engineered barriers or the host rock, the solubility limit of the gas will eventually be exceeded and the formation of a discrete gas phase will occur. Gases could continue to accumulate until the pressure becomes sufficient to be released in gaseous form. This report deals with the evolution of gas-related processes that can influence the long-term behaviour and safety of low- and intermediate-level waste (L/ILW) and high-level waste (HLW) repositories in Opalinus Clay. The main aim is to present a synthesis of processes and phenomena related to repository-produced gases and to assess their influence on repository performance. A current overview of gas sources, reactions and interactions, generation, consumption, and transport is provided. Furthermore, current scientific understanding is used to define safety function indicators and criteria, which are employed to evaluate the potential influence of repository-generated gas on safety-relevant properties of engineered and natural barriers. The assessment of gas generation, consumption and transport is addressed separately for the HLW and the L/ILW deep geological repositories. The employed methodology, which is common for both repository types, consists of the description and quantification of the potential gas sources, which include the waste, barrier components such as disposal canisters and other gas-generating repository components, and of the processes and reactions leading to the generation or

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

International Nuclear Information System (INIS)

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

2015-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-05-15

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

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

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

International Nuclear Information System (INIS)

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

2014-01-01

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

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.

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

Energy Technology Data Exchange (ETDEWEB)

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

2002-12-01

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

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

International Nuclear Information System (INIS)

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

2002-12-01

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

Radioactive waste disposal programme and siting regions for geological deep repositories. Executive summary. November 2008

International Nuclear Information System (INIS)

2008-11-01

the first stage, the producers of the radioactive waste have to propose some geological siting regions to the Federal Office of Energy (SFOE). On behalf of the waste producers NAGRA evaluated geological regions everywhere in Switzerland. Based on this process three regions are proposed to the SFOE for the repository for high-level radioactive wastes (HLW). For the repository for low- and intermediate-level radioactive wastes (L/ILW) six sites were retained. The regions proposed for a HLW repository could also match the requirements for a L/ILW repository

Characteristics of potential repository wastes

International Nuclear Information System (INIS)

Notz, K.J.

1989-01-01

The Office of Civilian Radioactive Waste Management (OCRWM) is responsible for the spent fuels and other wastes that will be disposed of in a geologic repository. The two major sources of these materials are commercial light-water reactor (LWR) spent fuel and immobilized high-level waste (HLW). Other wastes that may require long-term isolation include non-LWR spent fuels and miscellaneous sources such as activated metals. Detailed characterizations are required for all of these potential repository wastes. These characterizations include physical, chemical, and radiological properties. The latter must take into account decay as a function of time. This information has been extracted from primary data sources, evaluated, and assembled in a Characteristics Data Base which provides data in four formats: hard copy standard reports, menu-driven personal computer (PC) data bases, program-level PC data bases, and mainframe computer files. The Characteristics Data Base provides a standard set of self-consistent data to the various areas of responsibility including systems integration and waste stream analysis, storage, transportation, and geologic disposal. The data will be used for design studies, evaluation of alternatives, and system optimization by OCRWM and supporting contractors. 7 refs., 5 figs., 7 tabs

Repository development status in Spain

Energy Technology Data Exchange (ETDEWEB)

Rodriguez Beceiro, Alvaro; Zuloaga, Pablo [ENRESA (Spain)

2009-06-15

The paper describes the status of repository development for the different waste categories. Low and Intermediate Waste Disposal facility of El Cabril was commissioned in 1992 and is in normal operation. The major modifications and activities during the last years are the adaptation to waste streams not initially foreseen such as some decommissioning waste or waste from steel industry, and the improvement of its performance assessment, supported by a R and D and refined models. As part of this facility, a new disposal facility specifically intended for very low activity waste has been constructed and commissioned in July 2008. Its design is based on the European Directive for hazardous waste disposal. National policy for Nuclear Spent Fuel and High-Level waste is focused on the development of a centralized storage facility of the vault type, whose site location would be selected through a volunteering process. Meanwhile, with the aim of solving specific problems, three individual storage facilities are in different status at reactor sites. Research on final solution, including some repository aspects as well as separation and transmutation are being carried out in accordance to ENRESA's R and D program. ENRESA has developed conceptual designs for non site specific repositories, both in granite and clay, and has carried out their corresponding performance assessment exercises. (authors)

Massachusetts' participation in the US Department of Energy's Crystalline Repository Project: Final report

International Nuclear Information System (INIS)

Dukakis, M.S.; Johnston, P.W.; Walker, B. Jr.

1986-12-01

The purpose of this report is to review activities of the Crystalline Rock Project (CRP) in Massachusetts, the context in which it arose, and the implications and possibilities of future federal policy decisions on the siting of a HLW repository in a crystalline rock body. Although Massachusetts has been eliminated from consideration, there are still issues of concern to be resolved regarding the potential impacts upon Massachusetts if a site were ever to be chosen in New England

Waste package transfer, emplacement and retrievability in the French deep geological repository

Energy Technology Data Exchange (ETDEWEB)

Roulet, Alain; Delort, Daniel; Herve, Jean Francois; Bosgiraud, Jean Michel; Guenin, Jean Jacques [Technical Department ANDRA (France)

2009-06-15

Safe, reliable and reversible handling of waste is a significant issue related to the design and safety assessment of deep geological repository in France. The first step taken was to study various waste handling solutions. ANDRA also decided to fabricate and demonstrate industrial scale handling equipment for HLW (since 2003) and for ILW-LL wastes (since 2008). We will review the main equipment developed for the transfer process in the repository, for both types of waste, and underline the benefits of developing industrial demonstrators within the framework of international cooperation agreements. Waste retrieval capability will be simultaneously examined. Two types of waste have to be handled underground in Andra's repository. The HLW disposal package for vitrified waste is a 2 ton carbon steel cylindrical canister with a diameter of 600 mm. The weight of ILW-LL concrete disposal packages range from a minimum of 6 tonnes to over 20 tonnes, and their volume from approximately 5 to 10 m3. The underground transfer to the disposal drift requires moving the disposal package within a shielded transfer cask placed on a trailer. Transfer cask design has evolved since 2005, due to optimisation studies and as a result of industrial feedback from SKB. For HLW handling equipment two design options have been studied. In the first solution (Andra's Dossier 2005), the waste package are emplaced, one at a time, in the disposal drift by a pushing robot. Successive steps in design and proto-typing have lead to improve the design of the equipment and to gain confidence. Recently a fully integrated process has been successfully demonstrated, at full scale, (in a 100 m long mock up drift) as part of the EC funded ESDRED Project. This demonstrator is now on display in Andra's Technology Centre at Saudron, near the Bure Underground Laboratory. The second disposal option which has been investigated is based on a concept of utilising an external apparatus to push a row of

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-03-20

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

Stress analysis of HLW containers. Compas project

International Nuclear Information System (INIS)

1989-01-01

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

Center for Leadership Development (CLD) Repository

Data.gov (United States)

Office of Personnel Management — The Center for Leadership Development Repository stores various data including policies, procedures, governance, guidance, security, and financial documents of the...

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-11-15

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

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

International Nuclear Information System (INIS)

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

2014-11-01

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

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

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

International Nuclear Information System (INIS)

1987-10-01

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

Spent fuel and high level waste: Chemical durability and performance under simulated repository conditions. Results of a coordinated research project 1998-2004. Part 2: Results of a previously unpublished CRP: Performance of high level waste forms and packages under repository conditions. Results of a co-ordinated research project 1991-1998

International Nuclear Information System (INIS)

2007-07-01

The objective of the CRP (Coordinated Research Projekt) on the 'Performance of High Level Waste Forms and Packages under Repository Conditions' was to contribute to the development and implementation of proper and sound technologies for HLW and spent fuel management. Special emphasis was given to the identification of various waste form properties and the study of their long term durability in simulated repository conditions. Another objective was to promote the co-operation and exchange of information between Member States on experimental concerning behaviour of the waste form. The CRP was composed of research contracts and agreements with Argentina, Australia, Belgium, Canada, China, Czech Republic, Finland, France, Germany, India, Japan, Russia, and the United States of America. The publication includes 14 individual contributions of the participants to the CRP, which are indexed separately.

The influences of scientific information on the growing in opinion for high level waste repository. Focusing on education in civil engineering course

International Nuclear Information System (INIS)

Amemiya, Kiyoshi; Chijimatsu, Masakazu

2002-01-01

In this research, survey of awareness and attitude to high level radioactive waste (HLW) disposal on 36 students of a postgraduate course was conducted. They have been studying civil and rock engineering, so they belong to 'the Group' that acquires high education, culture and faculty to understand the science in geological disposal of HLW. First of all the awareness of danger or safety to HLW disposal was examined. Some 23% regard HLW disposal as safe, on the contrary 60% feel danger. This is similar to the awareness of the average public. And some 72% think that HLW should be disposal, but only 6% agree the repository in their town. It shows that the Group of high education has a tendency of calmly understand the necessity of disposal, but they also have a nature so-called 'not in my back yard (NIMBY)'. After that, the students were divided in two groups. Then, one group received information from the promoter, and another received information from opponents. The result of second questionnaire shows that the awareness of danger is affected strongly by given information even in this Group, but they become thoughtful and prudent in their opinion and decision-making as increasing information. Finally in this paper it is studied that 'what is the role of education of civil engineering?' and 'what is key issue in R and D of HLW disposal?' considering Public Acceptance. (author)

An evaluation of environmental effects of the DOE HLW repository siting and characterization program at Yucca Mountain, Nevada

International Nuclear Information System (INIS)

Winsor, M.F.; Ulland, L.M.

1989-01-01

This paper presents highlights of the Nevada Nuclear Waste Project Office (NWPO) environmental investigations in progress on the environmental effects of past and proposed activities of the Department of Energy (DOE) at the Yucca Mountain repository. The environmental investigations refer to those studies specifically related to resource evaluation, impact assessment and mitigation planning for the repository program; it is defined to exclude consideration of technical suitability determinations, socioeconomics and transportation. This paper addresses the question of what are the disturbances created by past and proposed DOE activities related to repository siting and characterization at Yucca Mountain. It discusses considerations in linking disturbance to the potential for significant adverse environmental impacts

Economic appraisal of deployment schedules for high-level radioactive waste repositories

International Nuclear Information System (INIS)

Hoai Linh Doan, Phuong; Duquesnoy, T.; Devezeaux de Lavergne, J.G.

2017-01-01

The deep geological repository (DGR) is considered as the definitive management solution for high-level radioactive waste (HLW). Countries defined different DGR implementation schedules, depending on their national context and political choices. We raise the question of the economic grounds of such political decisions by providing an economic analysis of different DGR schedules. We investigate the optimal timing for DGR commissioning based on available Nuclear Energy Agency (NEA) data (2013). Two scenarios are considered: -1) rescheduling the deployment of a DGR with the same initial operational period, and -2) rescheduling the deployment of a DGR with a shorter operational period, i.e. initial closure date. Given the long timescales of such projects, we also take into account the discounting effect. The first finding is that it appears more economically favorable to extend the interim storage than to dispose of the HLW immediately. Countries which chose 'immediate' disposal are willing to accept higher costs to quickly solve the problem. Another interesting result is that there is an optimal solution with respect to the length of DGR operational period and the waste flow for disposal. Based on data provided by the Organisation for Economic Cooperation and Development (OECD)/Nuclear Energy Agency (NEA), we find an optimal operating period of about 15 years with a flow of 2000 tHM/year. (authors)

Economic appraisal of deployment schedules for high-level radioactive waste repositories

Directory of Open Access Journals (Sweden)

Doan Phuong Hoai Linh

2017-01-01

Full Text Available The deep geological repository (DGR is considered as the definitive management solution for high-level waste (HLW. Countries defined different DGR implementation schedules, depending on their national context and political choices. We raise the question of the economic grounds of such political decisions by providing an economic analysis of different DGR schedules. We investigate the optimal timing for DGR commissioning based on available Nuclear Energy Agency (NEA data (2013. Two scenarios are considered: (1 rescheduling the deployment of a DGR with the same initial operational period, and (2 rescheduling the deployment of a DGR with a shorter operational period, i.e. initial closure date. Given the long timescales of such projects, we also take into account the discounting effect. The first finding is that it appears more economically favorable to extend the interim storage than to dispose of the HLW immediately. Countries which chose “immediate” disposal are willing to accept higher costs to quickly solve the problem. Another interesting result is that there is an optimal solution with respect to the length of DGR operational period and the waste flow for disposal. Based on data provided by the Organisation for Economic Cooperation and Development (OECD/Nuclear Energy Agency (NEA, we find an optimal operating period of about 15 years with a flow of 2000 tHM/year.

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

International Nuclear Information System (INIS)

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

1982-01-01

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

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

International Nuclear Information System (INIS)

Jacobs, R.A.

1998-01-01

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

HLW disposal in Germany - R and D achievements and outlook

International Nuclear Information System (INIS)

Steininger, W.

2006-01-01

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

Concept development for HLW disposal research tunnel

International Nuclear Information System (INIS)

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

2003-01-01

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

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

International Nuclear Information System (INIS)

Zheng Hualing; Chen Shi; Sun Donghui

2007-01-01

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

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

Science.gov (United States)

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

2016-12-01

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

Overview of adaptive phased management repository design development

International Nuclear Information System (INIS)

Russell, S.

2011-01-01

The Nuclear Waste Management Organization is implementing Adaptive Phased Management, Canada's plan for long-term management of used nuclear fuel. The organization is proceeding with the process for selecting a site in partnership with an informed and willing host community to safely and securely container and isolate used nuclear fuel in a deep geological repository in a suitable rock formation. Adaptive Phased Management is the culmination of more than 30 years of research, development and demonstration of repository concepts in Canada. Adaptive Phased Management uses a phased and adaptive step-wise approach to the multi-barrier system which is consistent with the long-term waste management approaches being developed in many other countries with nuclear power programs such as Sweden, Finland, Switzerland, the United Kingdom and France. The Nuclear Waste Management Organization is examining and developing conceptual designs for a deep geological repository and associated facilities for the placement of used nuclear fuel in long-lived containers. This paper will examine two of these generic conceptual designs which have recently been refined and updated. These conceptual designs will be used to support a pre-project review of repository design and safety by the Canadian Nuclear Safety Commission. (author)

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

International Nuclear Information System (INIS)

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

1999-01-01

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

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

International Nuclear Information System (INIS)

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

2005-01-01

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

HLW immobilization in glass

International Nuclear Information System (INIS)

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

1992-01-01

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

The safety case for a HLW repository in Opalinus clay: aims, methodology, first results

International Nuclear Information System (INIS)

Zuidema, Piet

2002-01-01

Piet Zuidema (Nagra, Switzerland) described the development of the safety case for a high level waste repository in Opalinus clay in which canisters would be placed in large vaults. The current phase of work was concerned with demonstrating the feasibility of the disposal concept. The Safety Case is taken to mean a set of arguments to support a statement that the proposed facility will meet relevant safety criteria and will include arguments giving the basis for confidence that those arguments are correct and properly taking account of uncertainties. The safety strategy was concerned both with the inherent robustness of the disposal concept and the adequacy of the assessment capability. As regards the former, the arguments being advanced were primarily qualitative. Key issues in terms of the documentation of the Safety Case were traceability and transparency of information, including how to ensure that key arguments did not become obscured because of the need to make available very large quantities of information

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

International Nuclear Information System (INIS)

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

1983-01-01

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

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

International Nuclear Information System (INIS)

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

1999-01-01

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

Development of an object-oriented simulation code for repository performance assessment

International Nuclear Information System (INIS)

Tsujimoto, Keiichi; Ahn, J.

1999-01-01

As understanding for mechanisms of radioactivity confinement by a deep geologic repository improves at the individual process level, it has become imperative to evaluate consequences of individual processes to the performance of the whole repository system. For this goal, the authors have developed a model for radionuclide transport in, and release from, the repository region by incorporating multiple-member decay chains and multiple waste canisters. A computer code has been developed with C++, an object-oriented language. By utilizing the feature that a geologic repository consists of thousands of objects of the same kind, such as the waste canister, the repository region is divided into multiple compartments and objects for simulation of radionuclide transport. Massive computational tasks are distributed over, and executed by, multiple networked workstations, with the help of parallel virtual machine (PVM) technology. Temporal change of the mass distribution of 28 radionuclides in the repository region for the time period of 100 million yr has been successfully obtained by the code

The RAGE Game Software Components Repository for Supporting Applied Game Development

Directory of Open Access Journals (Sweden)

Krassen Stefanov

2017-09-01

Full Text Available This paper presents the architecture of the RAGE repository, which is a unique and dedicated infrastructure that provides access to a wide variety of advanced technology components for applied game development. The RAGE project, which is the principal Horizon2020 research and innovation project on applied gaming, develops up to three dozens of software components (RAGE software assets that are reusable across a wide diversity of game engines, game platforms and programming languages. The RAGE repository provides storage space for assets and their artefacts and is designed as an asset life-cycle management system for defining, publishing, updating, searching and packaging for distribution of these assets. It will be embedded in a social platform for asset developers and other users. A dedicated Asset Repository Manager provides the main functionality of the repository and its integration with other systems. Tools supporting the Asset Manager are presented and discussed. When the RAGE repository is in full operation, applied game developers will be able to easily enhance the quality of their games by including selected advanced game software assets. Making available the RAGE repository system and its variety of software assets aims to enhance the coherence and decisiveness of the applied game industry.

Thermal, operational, and economic aspects of repository design alternatives

International Nuclear Information System (INIS)

Closs, K.D.; Papp, R.; Bechthold, W.; Engelmann, H.J.; Hartje, B.

1989-01-01

This paper discusses how a broad spectrum of heat-generating nuclear waste types like reprocessing waste (HLW, ILW), spent LWR, and spent HTR fuel will be disposed of in a future German repository located in a salt dome. Different package and emplacement concepts for the various waste forms are feasible, ranging from pure borehole emplacement of canisters and drums to pure drift emplacement of heavily shielded casks as well as combinations of both concepts. Optimization of the whole back-end system (waste treatment, interim storage, and disposal) is performed taking into account thermal, operational, and economic aspects. From a radiological viewpoint, the drift emplacement concept is superior to all concepts in as much as the occupational dose is lowest and zero release from the casks is attainable during normal and anomalous repository operation. As far as economics are concerned, the cost of the heavily shielded casks necessary for drift emplacement contributes markedly to the overall cost of the whole back-end system. Drift emplacement can compete economically with the other concepts only if the casks are used both for interim storage and disposal

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

International Nuclear Information System (INIS)

Kitamoto, A.; Mulyanto

1993-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2013-11-13

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

Open and closed fuel cycle of HWR and PWR. How large is the high-level radioactive wastes repository; Ciclos de combustible abierto y cerrado con HWR y PWR. ?Cuanto mas grande es el repositorio de residuos radiactivos de alta actividad?

Energy Technology Data Exchange (ETDEWEB)

Bevilacqua, Arturo M. [Comision Nacional de Energia Atomica, San Carlos de Bariloche (Argentina). Centro Atomico Bariloche

1996-07-01

A conceptual analysis was carried out on the size of a high-level wastes (HLW) repository for the waste arising from once-through and closed fuel cycles with (HLW) and PWR. The mass, the activity and thermal loading was calculated with the ORIGEN2.1 computer code for the spent fuel and for the high-level liquid wastes. It was considered a minimum burnup of 7.000 MW.d/t U and 33.000 MW.d/t U for HWR and PWR respectively, cooling times of 20 and 55 years, reprocessing recovery ratios of 99% and 99.7% and a total electricity production of 81.6 GW(e).a. It was concluded that the cooling time is the most important repository size reproduction parameter for the closed cycles. On the other hand, the spent fuel mass for the once-through cycles does not depend on the cooling time what prevents repository size reduction once a cooling time of 55 years is reached. The repository size reduction in the case of HWR is larger than in the case of PWR, owing to the larger fuel mass required to produce the specific electricity amount. (author)

Some inspirations drawn from the SKB license-application report for the construction of the final repository for spent nuclear fuel in Sweden

International Nuclear Information System (INIS)

Xu Guoqing

2014-01-01

The following problems are described in this paper: the current state of nuclear energy and spent fuel in Sweden, the brief process of repository site selection and the concise geological settings of candidate site Forsmark, the option of repository design, the data presented in license-application report on repository, construction, the review work presented by OECD/NEA international peer review team (IRT) and some considerations noted by the author: 1) Why the design of underground engineering, also known as 5 shafts + l ramp, is adopted? 2) Why the timescale for the assessment is one million years? 3) Why the computer codes used in performance assessment at Forsmark candidate repository site are different from those used at Yucca Mountain candidate repository site? 4) How to carry out the public involvement, support from local government and volumtary siting in China? 5) What is the most important in the R and D program of HLW disposal? 6) What shall we do for the future license-application report for the construction of the final repository in China? (author)

PROCESS FOR LICENSE APPLICATION DEVELOPMENT FOR THE GEOLOGIC REPOSITORY

International Nuclear Information System (INIS)

DOUGLAS M. FRANKS AND NORMAN C. HENDERSON

1997-01-01

The Department of Energy (DOE), specifically the Office of Civilian Radioactive Waste Management (OCRWM) has been charged by the U.S. Congress, through the Nuclear Waste Policy Act (NWPA), with the responsibility for obtaining a license to develop a geologic repository. The NRC is the licensing authority for geologic disposal, and its regulations pertinent to construction authorization and license application are specified in 10 CFR Part 60, Disposal of High-Level Radioactive Wastes in Geologic Repositories, (section)60.21ff and (section)60.31ff. This paper discusses the process the Yucca Mountain Site Site Characterization Project (YMP) will use to identify and apply regulatory and industry guidance to development of the license application (LA) for a geologic repository at Yucca Mountain, Nevada. This guidance will be implemented by the ''Technical Guidance Document for Preparation of the License Application'' (TGD), currently in development

Process for license application development for the geologic repository

International Nuclear Information System (INIS)

Franks, D.M.; Henderson, N.C.

1998-01-01

The Department of Energy (DOE), specifically the Office of Civilian Radioactive Waste Management (OCRWM) has been charged by the US Congress, through the Nuclear Waste Policy Act (NWPA), with the responsibility for obtaining a license to develop a geologic repository. The NRC is the licensing authority for geologic disposal, and its regulations pertinent to construction authorization and license application are specified in 10 CFR Part 60, Disposal of High-Level Radioactive Wastes in Geologic Repositories, section 60.21ff and section 60.31ff. This paper discusses the process the Yucca Mountain Site Characterization Project (YMP) will use to identify and apply regulatory and industry guidance to development of the license application (LA) for a geologic repository at Yucca Mountain, Nevada. This guidance will be implemented by the Technical Guidance Document for Preparation of the License Application (TGD), currently in development

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

Energy Technology Data Exchange (ETDEWEB)

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

2011-12-29

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

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

International Nuclear Information System (INIS)

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

2011-01-01

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

Regional groundwater chemical characteristics of Aqishan pre-selected site for high level radioactive waste repository and its hydrogeological significance

International Nuclear Information System (INIS)

Guo Yonghai; Dong Jiannan; Liu Shufen; Zhou Zhichao

2014-01-01

Aqishan area located in Xinjiang Uygur Automonous Region is one of the main preselected site of disposal repository for high-level radioactive waste (HLW) in our country. Groundwater chemical feature is one of the most important consideration factors in the siting and site evaluation for high-level radioactive waste repository, From 2012 to 2013, the regional field hydrogeochemical investigation was carried out in study area and more than 30 groundwater samples were collected. According to the measurement data, the groundwater chemical features for different subareas are discussed in the paper. Furthermore, the location of discharge area of groundwater in Aqishan area was estimated according to the chemical features of different subareas. (authors)

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

International Nuclear Information System (INIS)

Bastiaens, W.

2007-01-01

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

A control volume scheme for three-dimensional transport: buffer and matrix effect on a decay chain transport in the repository

International Nuclear Information System (INIS)

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

2002-01-01

Using a three-dimensional numerical code, B3R developed for nuclide transport of an arbitrary length of decay chain in the buffer between the canister and adjacent rock in a high-level radioactive waste repository by adopting a finite difference method utilizing the control-volume scheme, some illustrative calculations have been done. A linear sorption isotherm, nuclide transport due to diffusion in the buffer and the rock matrix, and advection and dispersion along thin rigid parallel fractures existing in a saturated porous rock matrix as well as diffusion through the fracture wall into the matrix is assumed. In such kind of repository, buffer and rock matrix are known to be important physico-chemical barriers in nuclide retardation. To show effects of buffer and rock matrix on nuclide transport in HLW repository and also to demonstrate usefulness of B3R, several cases of breakthrough curves as well as three-dimensional plots of concentration isopleths associated with these two barriers are introduced for a typical case of decay chain of 234 U→ 230 Th→ 226 Ra, which is the most important chain as far as the human environment is concerned

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

International Nuclear Information System (INIS)

Yamada, Fumitaka; Sonobe, Hitoshi; Igarashi, Hiroshi

2008-02-01

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

An analysis of the dispute process regarding high-level nuclear waste repository siting in Toyo-cho, Japan: Decisive factors in the dispute and roles of the governments and experts

Energy Technology Data Exchange (ETDEWEB)

Komatsuzaki, Shunsaku; Horii, Hideyuki (Dept. of Civil Engineering, Univ. of Tokyo (Japan)); Saigo, Takahiro (Mitsubishi Research Institute, Inc. (Japan))

2010-09-15

The siting policy of HLW repository in Japan was 'application-based' until 2007 and Toyo-cho is the only municipality which applied for the Literature Survey. In Toyo-cho, however, a serious antagonism among citizens occurred and the application was withdrawn after the mayor was replaced by election. Our detailed analysis of the process based on the methods of political science and psychology shows five decisive factors: 1) opposing activists both in the town and from outside successfully changed citizens' perceptions of HLW by rhetorical expressions, 2) the mayor lacks careful actions and effective policy adviser, 3) NUMO, an organization which runs HLW projects, didn't effectively coordinate Toyo-cho and stakeholders, 4) the municipal government and council exercised very limited influences on the dispute despite their political authority, and 5) the existence of grant adversely influenced the citizens since it causes criticism that Toyo-cho applies a repository for grant. We finally conclude that the substantial problems, caused by the five decisive factors, were the propagation of enthusiastic opposition and the lack of peaceful deliberation based on local governance. In order to avoid enthusiastic opposition and to realize responsible decision making, or negotiation, we suggest that A) active and prompt response of experts, especially political/administrative ones, to radical opposing activities, B) solution to the adverse influence of the grant by the government's agenda

An analysis of the dispute process regarding high-level nuclear waste repository siting in Toyo-cho, Japan: Decisive factors in the dispute and roles of the governments and experts

International Nuclear Information System (INIS)

Komatsuzaki, Shunsaku; Horii, Hideyuki; Saigo, Takahiro

2010-09-01

The siting policy of HLW repository in Japan was 'application-based' until 2007 and Toyo-cho is the only municipality which applied for the Literature Survey. In Toyo-cho, however, a serious antagonism among citizens occurred and the application was withdrawn after the mayor was replaced by election. Our detailed analysis of the process based on the methods of political science and psychology shows five decisive factors: 1) opposing activists both in the town and from outside successfully changed citizens' perceptions of HLW by rhetorical expressions, 2) the mayor lacks careful actions and effective policy adviser, 3) NUMO, an organization which runs HLW projects, didn't effectively coordinate Toyo-cho and stakeholders, 4) the municipal government and council exercised very limited influences on the dispute despite their political authority, and 5) the existence of grant adversely influenced the citizens since it causes criticism that Toyo-cho applies a repository for grant. We finally conclude that the substantial problems, caused by the five decisive factors, were the propagation of enthusiastic opposition and the lack of peaceful deliberation based on local governance. In order to avoid enthusiastic opposition and to realize responsible decision making, or negotiation, we suggest that A) active and prompt response of experts, especially political/administrative ones, to radical opposing activities, B) solution to the adverse influence of the grant by the government's agenda

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

International Nuclear Information System (INIS)

Lang-Lenton, J.

2007-01-01

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

Development and application of a quality assurance system for waste repositories

International Nuclear Information System (INIS)

Beckmerhagen, I.; Berg, H.P.

1996-01-01

A large set of requirements are developed for the structures, systems and components of a waste repository. These requirements cover different regulations concerning planning, design, construction, operation and decommissioning of a waste repository as well as nuclear-specific requirements on the design of the plant. The implementation of a quality assurance (QA) system is an essential task in order to ensure that the requirements are fulfilled by systematic means. The QA system for the German repositories for radioactive waste is built up as a hierarchic system and described in more detail in particular covering aspects like maintenance of QA relevant systems, structures and components as well as the procedure in case of changes. For the operational phase of a repository a separate site-specific manual had to be developed. Such a manual has been elaborated for the operation of the Morsleben repository; main topics of this manual are presented

Enhanced HLW glass formulations for the waste treatment and immobilization plant

Energy Technology Data Exchange (ETDEWEB)

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

2013-07-01

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

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

International Nuclear Information System (INIS)

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

2005-01-01

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

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

International Nuclear Information System (INIS)

Sasaki, Noriaki

1994-01-01

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

An open repositories network development for medical teaching resources.

Science.gov (United States)

Soula, Gérard; Darmoni, Stefan; Le Beux, Pierre; Renard, Jean-Marie; Dahamna, Badisse; Fieschi, Marius

2010-01-01

The lack of interoperability between repositories of heterogeneous and geographically widespread data is an obstacle to the diffusion, sharing and reutilization of those data. We present the development of an open repositories network taking into account both the syntactic and semantic interoperability of the different repositories and based on international standards in this field. The network is used by the medical community in France for the diffusion and sharing of digital teaching resources. The syntactic interoperability of the repositories is managed using the OAI-PMH protocol for the exchange of metadata describing the resources. Semantic interoperability is based, on one hand, on the LOM standard for the description of resources and on MESH for the indexing of the latter and, on the other hand, on semantic interoperability management designed to optimize compliance with standards and the quality of the metadata.

A new approach to performance assessment of barriers in a repository. Executive summary, draft, technical appendices. Final report

International Nuclear Information System (INIS)

Mueller-Hoeppe, N.; Krone, J.; Niehues, N.; Poehler, M.; Raitz von Frentz, R.; Gauglitz, R.

1999-06-01

Multi-barrier systems are accepted as the basic approach for long term environmental safe isolation of radioactive waste in geological repositories. Assessing the performance of natural and engineered barriers is one of the major difficulties in producing evidence of environmental safety for any radioactive waste disposal facility, due to the enormous complexity of scenarios and uncertainties to be considered. This report outlines a new methodological approach originally developed basically for a repository in salt, but that can be transferred with minor modifications to any other host rock formation. The approach is based on the integration of following elements: (1) Implementation of a simple method and efficient criteria to assess and prove the tightness of geological and engineered barriers; (2) Using the method of Partial Safety Factors in order to assess barrier performance at certain reasonable level of confidence; (3) Integration of a diverse geochemical barrier in the near field of waste emplacement limiting systematically the radiological consequences from any radionuclide release in safety investigations and (4) Risk based approach for the assessment of radionuclide releases. Indicative calculations performed with extremely conservative assumptions allow to exclude any radiological health consequences from a HLW repository in salt to a reference person with a safety level of 99,9999% per year. (orig.)

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

International Nuclear Information System (INIS)

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

2003-02-01

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

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

International Nuclear Information System (INIS)

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

1997-01-01

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

Gas formation in ILW and HLW repositories, evaluation and modelling of the production rates and consequences on the safety of the repository

International Nuclear Information System (INIS)

Besnus, F.

1990-01-01

This paper summarizes the main gas formation mechanisms in deep radioactive waste repositories. Production rates and overall gas volumes were estimated and showed predominance of hydrogen production by anoxic corrosion and radiolysis for French wastes. Gas evolution in the near field has been modeled. First results issued from a sensitivity analysis showed desaturation of the storage cavities for a wide range of parameter values

Evaluation of alternative shaft-sinking techniques for high-level nuclear waste (HLW) deep geologic repositories. Final report (Task 3), June 1981-July 1982

International Nuclear Information System (INIS)

Gonano, L.; Findley, D.; Wildanger, W.; Gates, R.; Phillips, S.

1983-03-01

This report represents the results of Task 3 of US Nuclear Regulatory Commission (NRC) Contract, Technical Assistance for Repository Design. The purpose of the complete project is to provide NRC with technical assistance for the following reasons: To enable the focused, adequate review by NRC of aspects related to design and construction of an in situ test facility and final geologic repository, as presented in US Department of Energy (DOE) Site Characterization Reports (SRC), To ascertain that the DOE site characterization program will provide, as far as possible, all the information necessary to permit a review to be conducted by NRC of a license application for construction authorization. It is assumed that the Site Characterization Report and License Application will describe the exploratory shaft and concept designs for the repository shafts. This report provides a comparative evaluation of various shaft sinking techniques for production shafts for a repository. The primary comparative evaluation has been conducted for 14-ft internal diameter shafts developed in two composite media using four different methods of sinking/lining. The technical, cost and schedule comparisons draw a major distinction between shafts sunk blind and those which utilize bottom access. Based on the system of ranking introduced to grade the significant attributes of each method and the resulting design, it is concluded that for application to repository access, no one particular method of sinking exhibits a clear overall superiority. When a specific site is made available for a study of the most suitable shaft sinking methods, it will be necessary to establish actual geological conditions and technological capabilities and the comparisons presented herein reviewed accordingly

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

International Nuclear Information System (INIS)

Kitamoto, Asashi; Mulyanto.

1995-01-01

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

Policy and practice of radioactive waste management in India

International Nuclear Information System (INIS)

Sunder Radzhan, N.S.

1986-01-01

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

Development of the SIOPE DIPG network, registry and imaging repository

DEFF Research Database (Denmark)

Veldhuijzen van Zanten, Sophie E M; Baugh, Joshua; Chaney, Brooklyn

2017-01-01

was developed, The SIOPE DIPG Registry and Imaging Repository, to centrally collect data of DIPG patients. As for April 2016, clinical data as well as MR-scans of 694 patients have been entered into the SIOPE DIPG Registry/Imaging Repository. The median progression free survival is 6.0 months (95% Confidence...

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

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2010-01-04

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

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

International Nuclear Information System (INIS)

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

2010-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2013-11-13

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

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

Radioactive waste disposal programme and siting regions for geological deep repositories. Executive summary. November 2008; Entsorgungsprogramm und Standortgebiete fuer geologische Tiefenlager. Zusammenfassung. November 2008

Energy Technology Data Exchange (ETDEWEB)

NONE

2008-11-15

the first stage, the producers of the radioactive waste have to propose some geological siting regions to the Federal Office of Energy (SFOE). On behalf of the waste producers NAGRA evaluated geological regions everywhere in Switzerland. Based on this process three regions are proposed to the SFOE for the repository for high-level radioactive wastes (HLW). For the repository for low- and intermediate-level radioactive wastes (L/ILW) six sites were retained. The regions proposed for a HLW repository could also match the requirements for a L/ILW repository.

Study of an applicability of technologies developed in the conventional industries from the view point of developing the geological disposal system

International Nuclear Information System (INIS)

Ushio, Kazuhiro; Ando, Yasumasa; Kubota, Kazuo; Sokejima, Susumu

1999-02-01

The geological disposal study of HLW (High Level Wastes) is being developed in Japan. Especially, JNC has played the central role to proceed this project, while in the industries, from the viewpoint of the environmental measures, various technologies and materials have been developed. Some of them might be applied into the geological disposal. The purpose of this study is to investigate such technologies and their applicability to the geological disposal system. Firstly, the environmental technologies used for the repository of industrial wastes were studied. The concepts of management and the regulations for the repository are summarized, and compared with the current geological disposal concept. Secondly, concerning structural and durable materials, their properties and usage were overviewed and their applicability to the current geological disposal concept was studied. (J.P.N.)

Historical fuel reprocessing and HLW management in Idaho

International Nuclear Information System (INIS)

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

1997-01-01

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

Sectoral Plan 'Deep Geological Disposal', Stage 2. Proposed site areas for the surface facilities of the deep geological repositories as well as for their access infrastructure. General report

International Nuclear Information System (INIS)

2011-12-01

In line with the provisions of the nuclear energy legislation, the sites for deep geological disposal of Swiss radioactive waste are selected in a three-stage Sectoral Plan process (Sectoral Plan for Deep Geological Disposal). The disposal sites are specified in Stage 3 of the selection process with the granting of a general licence in accordance with the Nuclear Energy Act. The first stage of the process was completed on 30 th November 2011, with the decision of the Federal Council to incorporate the six geological siting regions proposed by the National Cooperative for the Disposal of Radioactive Waste (NAGRA) into the Sectoral Plan for Deep Geological Disposal, for further evaluation in Stage 2. The decision also specifies the planning perimeters within which the surface facilities and shaft locations for the repositories will be constructed. In the second stage of the process, at least two geological siting regions each will be specified for the repository for low- and intermediate-level waste (L/ILW) and for the high-level waste (HLW) repository and these will undergo detailed geological investigation in Stage 3. For each of these potential siting regions, at least one location for the surface facility and a corridor for the access infrastructure will also be specified. NAGRA is responsible, at the beginning of Stage 2, for submitting proposals for potential locations for the surface facilities and their access infrastructure to the Federal Office of Energy (SFOE); these are then considered by the regional participation bodies in the siting regions. The present report and its annexes volume document these proposals. In Stage 2, under the lead of the SFOE, socio-economic-ecological studies will also be carried out to investigate the impact of a repository project on the environment, economy and society. The present reports also contain the input data to be provided by NAGRA for the generic (site-independent) part of these impact studies. A meaningful discussion

Considerations for developing seismic design criteria for nuclear waste storage repositories

International Nuclear Information System (INIS)

Owen, G.N.; Yanev, P.I.; Scholl, R.E.

1980-04-01

The function of seismic design criteria is to reduce the potential for hazards that may arise during various stages of the repository life. During the operational phase, the major concern is with the possible effects of earthquakes on surface facilities, underground facilities, and equipment. During the decommissioned phase, the major concern is with the potential effects of earthquakes on the geologic formation, which may result in a reduction in isolation capacity. Existing standards and guides or criteria used for the static and seismic design of licensed nuclear facilities were reviewed and evaluated for their applicability to repository design. This report is directed mainly toward the development of seismic design criteria for the underground structures of repositories. An initial step in the development of seismic design criteria for the underground structures of repositories is the development of performance criteria, or minimum standards of acceptable behavior. A number of possible damage modes are identified for the operating phase of the repository; however, no damage modes are foreseen that would perturb the long-term function of the repository, except for the possibility of increased permeability within the rock mass. Subsequent steps in formulating acceptable seismic design criteria for the underground structures involve the quantification of the design process. The report discusses the necessity of specifying the form of ground motion that would be needed for seismic analysis and the procedures that may be used for making ground motion predictions. Further discussions outline what is needed for analysis, including rock properties, failure criteria, modeling techniques, seismic hardening criteria for the host rock mass, and probabilistic considerations

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.

Assessment of national systems for obtaining local siting acceptance of nuclear waste management facilities (October 1, 1985). Volume 2. Summary of principal new (April 1, 1983-October 1, 1985) developments relating to the siting of waste management facilities

International Nuclear Information System (INIS)

Paige, H.W.; Numark, N.J.

1985-01-01

This report is the fourth in a series of periodic surveys of approaches and progress in other countries in dealing with the problems of obtaining local acceptance for siting of waste management facilities. Of the countries visited (Belgium, FRG, Finland, Sweden, Switzerland, and the UK) all have been engaged in recent years in the process of selecting and obtaining state and local acceptance of sites for new LLW repositories. Only Sweden has been successful thus far. Success has been understandably even more elusive in the siting of HLW repositories. Although there is also one country, FRG, that has gotten provisional site approval by the state and local governments for a HLW repository, the political process by which this was achieved does not appear to be one that could be duplicated elsewhere, and all other countries are still years away from making a site-specific selection or recommendation. Fortunately this need not create a serious safety, political or logistical problem. For those countries not having their spent fuel reprocessed, the spent fuel storage cask concept is available for safe storage of spent fuel at the point of origin for as long as needed until a HLW repository is available. For those countries which will be having to dispose of HLW resulting from reprocessing, air cooled and water cooled surface storage facilities are proven and acceptable options for interim long-term (decades) storage awaiting permanent disposal in repositories when available. One country has recently successfully sited a new reprocessing plant. After several years of rejection by state authorities, FRG now has two states willing and anxious to have a reprocessing plant. Construction is now underway at one of the sites

Initial demonstration of the NRC`s capability to conduct a performance assessment for a High-Level Waste Repository

Energy Technology Data Exchange (ETDEWEB)

Codell, R.; Eisenberg, N.; Fehringer, D.; Ford, W.; Margulies, T.; McCartin, T.; Park, J.; Randall, J.

1992-05-01

In order to better review licensing submittals for a High-Level Waste Repository, the US Nuclear Regulatory Commission staff has expanded and improved its capability to conduct performance assessments. This report documents an initial demonstration of this capability. The demonstration made use of the limited data from Yucca Mountain, Nevada to investigate a small set of scenario classes. Models of release and transport of radionuclides from a repository via the groundwater and direct release pathways provided preliminary estimates of releases to the accessible environment for a 10,000 year simulation time. Latin hypercube sampling of input parameters was used to express results as distributions and to investigate model sensitivities. This methodology demonstration should not be interpreted as an estimate of performance of the proposed repository at Yucca Mountain, Nevada. By expanding and developing the NRC staff capability to conduct such analyses, NRC would be better able to conduct an independent technical review of the US Department of Energy (DOE) licensing submittals for a high-level waste (HLW) repository. These activities were divided initially into Phase 1 and Phase 2 activities. Additional phases may follow as part of a program of iterative performance assessment at the NRC. The NRC staff conducted Phase 1 activities primarily in CY 1989 with minimal participation from NRC contractors. The Phase 2 activities were to involve NRC contractors actively and to provide for the transfer of technology. The Phase 2 activities are scheduled to start in CY 1990, to allow Sandia National Laboratories to complete development and transfer of computer codes and the Center for Nuclear Waste Regulatory Analyses (CNWRA) to be in a position to assist in the acquisition of the codes.

Compas project stress analysis of HLW containers intermediate testwork

International Nuclear Information System (INIS)

Ove Arup and Partners London

1990-01-01

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

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

International Nuclear Information System (INIS)

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

2011-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2011-12-29

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

A regulatory perspective on design and performance requirements for engineered systems in high-level waste

International Nuclear Information System (INIS)

Bernero, R.M.

1992-01-01

For engineered systems, this paper gives an overview of some of the current activities at the U.S. Nuclear Regulatory Commission (NRC), with the intent of elucidating how the regulatory process works in the management of high-level waste (HLW). Throughout the waste management cycle, starting with packaging and transportation, and continuing to final closure of a repository, these activities are directed at taking advantage of the prelicensing consultation period, a period in which the NRC, DOE and others can interact in ways that will reduce regulatory, technical and institutional uncertainties, and open the path to development and construction of a deep geologic repository for permanent disposal of HLW. Needed interactions in the HLW program are highlighted. Examples of HLW regulatory activities are given in discussions of a multipurpose-cask concept and of current NRC work on the meaning of the term substantially complete containment

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

International Nuclear Information System (INIS)

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

2012-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2012-12-13

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

Economic analysis of waste management alternatives for reprocessing wastes

International Nuclear Information System (INIS)

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

1984-02-01

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

An overview on the national strategy to implement a deep geological repository in Romania

International Nuclear Information System (INIS)

Negut, G.; Ghitescu, P.; Dupleac, D.; Prisecaru, I.

2010-01-01

Since 1996 in Romania was started operation Candu 700 MW Unit 1 Cernavoda Nuclear Power Station and in 2007 begun operation of the Candu 700 MW Unit 2. The energy produced by nuclear units is accompanied by radioactive waste production. According with European Union requirements in Romania was created National Agency for Radioactive Waste (ANDRAD) in 2003. ANDRAD business is radioactive waste management. ANDRAD, together with the stakeholders, worked the law of great radioactive waste generators contribution for radioactive waste management, which was approved by Governmental Ordinance in September 2007. ANDRAD is responsible manager of this fund. ANDRAD is responsible, also, with the National Strategy for radioactive waste management. Romania's National Strategy for Energy approved in 2007 by Government Ordinance says that a deep geological repository for spent fuel (SF) and High Level Waste (HLW) is to be put in operation around 2055. IAEA supported ANDRAD in a Technical Cooperation Project for a concept of a geological repository of radioactive waste. A strategy to implement o geological repository in Romania was drafted. There are problems with potential rocks and sites to host a geological repository. There are problems for funding this project and also sensitive and serious problems connected with social and political issues. Paper presents this strategy and all the problems arisen by implantation of this strategy. (authors)

Spent fuel and HLW transportation the French experience

International Nuclear Information System (INIS)

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

1995-01-01

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

Process mining software repositories: do developers work as expected?

NARCIS (Netherlands)

Poncin, W.; Serebrenik, A.; Brand, van den M.G.J.

2012-01-01

Modern software development commonly makes use of a multitude of software repositories. How can these help us to understand the on-going development process? Researchers of Eindhoven University of Technology design new methods revealing how software has been developed.

Development of the Canadian used fuel repository engineered barrier system

Energy Technology Data Exchange (ETDEWEB)

Hatton, C., E-mail: chatton@nwmo.ca [Nuclear Waste Management Organization, Toronto, ON (Canada)

2015-07-01

The Nuclear Waste Management Organization (NWMO) is responsible for the implementation of Adaptive Phased Management (APM), the federally-approved plan for the 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. In implementing APM, the NWMO is committed to ensure consistency with international best practices in the development of its repository system, including any advances in technology. In 2012, the NWMO undertook an optimization study to look at both the design and manufacture of its engineered barriers. This study looked at current technologies for the design and manufacture of used fuel containers, placement technologies, repository design, and buffer and sealing systems, while taking into consideration the state of the art worldwide in repository design and acceptance. The result of that study is the current Canadian engineered barrier system, consisting of a 2.7 tonne used fuel container with a carbon-steel core, copper-coated surface and welded spherical heads. The used fuel container is encapsulated in a bentonite buffer box at the surface and then transferred underground. Once underground, the used fuel is placed into a repository room which is cut into the rock using traditional drill-and-blast technologies. This paper explains the logic for the selection of the container and sealing system design and the development of innovative technologies for their manufacture including the use of laser welding, cold spray and pulsed-electrodeposition copper coating for the manufacture of the used fuel container, isostatic presses for the production of the one-piece bentonite blocks, and slip-skid technologies for placement into the repository. (author)

Development of the Canadian used fuel repository engineered barrier system

International Nuclear Information System (INIS)

Hatton, C.

2015-01-01

The Nuclear Waste Management Organization (NWMO) is responsible for the implementation of Adaptive Phased Management (APM), the federally-approved plan for the 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. In implementing APM, the NWMO is committed to ensure consistency with international best practices in the development of its repository system, including any advances in technology. In 2012, the NWMO undertook an optimization study to look at both the design and manufacture of its engineered barriers. This study looked at current technologies for the design and manufacture of used fuel containers, placement technologies, repository design, and buffer and sealing systems, while taking into consideration the state of the art worldwide in repository design and acceptance. The result of that study is the current Canadian engineered barrier system, consisting of a 2.7 tonne used fuel container with a carbon-steel core, copper-coated surface and welded spherical heads. The used fuel container is encapsulated in a bentonite buffer box at the surface and then transferred underground. Once underground, the used fuel is placed into a repository room which is cut into the rock using traditional drill-and-blast technologies. This paper explains the logic for the selection of the container and sealing system design and the development of innovative technologies for their manufacture including the use of laser welding, cold spray and pulsed-electrodeposition copper coating for the manufacture of the used fuel container, isostatic presses for the production of the one-piece bentonite blocks, and slip-skid technologies for placement into the repository. (author)

Current status of system development to provide databases of nuclides migration

International Nuclear Information System (INIS)

Sasamoto, Hiroshi; Yoshida, Yasushi; Isogai, Takeshi; Suyama, Tadahiro; Shibata, Masahiro; Yui, Mikazu; Jintoku, Takashi

2005-01-01

JNC has developed databases of nuclides migration for safety assessment of high-level radioactive waste (HLW) repository, and they have been used in the second progress report to present the technical reliability of HLW geological disposal system in Japan. The technical level and applicability of databases have been highly evaluated even overseas. To provide the databases broadly over the world and to promote the use of the databases, we have performed the followings: 1) development of tools to convert the database format from geochemical code PHREEQE to PHREEQC, GWB and EQ3/6 and 2) set up a web site (http://migrationdb.jnc.go.jp) which enables the public to access to the databases. As a result, the number of database users has significantly increased. Additionally, a number of useful comments from the users can be applied to modification and/or update of databases. (author)

Development Support Environment of Business ApplicationsBased on a Multi-Grain-Size Repository

Science.gov (United States)

Terai, Koichi; Izumi, Noriaki; Yamaguchi, Takahira

In order to build the Web-based application as a shopping site on the Web, various ideas from the different viewpoints are required, such as enterprise modeling, workflow modeling, software development, and so on. From the above standpoint, this paper proposes an integrated environment to support the whole development process of analysis, design and implementation of business application. In order to reuse know-hows of various ideas in the business application development, we device a multi-grain-size repository, which consists of coarse-, middle-, and fine-grain-size repositories that correspond to the enterprise models, workflow models, and software models, respectively. We also provide a methodology that rebuilds heterogeneous information resources required for the business applications development into a multi-grain-size repository based on ontologies. The contents of the repositories are modeled by the is-a, has-a, and E-R relations, and described by the XML language. We have implemented Java-based prototype environment with the tools dealing with the multi-layered repository and confirmed that it supports us in various phases of business application development including business model manifestation, detailed business model definition and an implementation of business software applications.

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

International Nuclear Information System (INIS)

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

2010-01-01

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

Waste-Mixes Study for space disposal

International Nuclear Information System (INIS)

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

1983-01-01

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

Implementation of a geological disposal facility (GDF) in the UK by the NDA Radioactive Waste Management Directorate (RWMD): the potential for interaction between the co-located ILW/LLW and HLW/SF components of a GDF - 16306

International Nuclear Information System (INIS)

Towler, George; Hicks, Tim; Watson, Sarah; Norris, Simon

2009-01-01

for ILW/LLW and HLW/SF modules to be co-located without compromising key safety functions of different barrier components, and this reflects international precedents, e.g. the Andra and Nagra repository designs. There are two key technical issues that need to be managed in designing the geometry of the co-located GDF: (i) the heat flux from the HLW/SF module interacting with the ILW/LLW module, and (ii) the potential for development of an alkaline plume from the ILW/LLW module interacting with the HLW/SF module; particularly within fractured host rocks. (authors)

Development and design application of cerium (IV) decontamination process

International Nuclear Information System (INIS)

Bray, L.A.; Seay, J.M.

1988-01-01

A simple and effective method was developed for decontamination of high-level waste canisters. This method of chemical decontamination is applicable to a wide variety of contaminated equipment found in the nuclear industry. Conceptual design of the cerium [Ce(IV)] decontamination process equipment has been completed for the West Valley Demonstration project (WVDP) vitrification facility. This remote equipment, which is the first engineering scale application of this technology, will remove surface contamination from stainless-steel (SS) containers containing high-level waste (HLW) glass prior to placing them into temporary storage and ultimate shipment to a U.S. Department of Energy (DOE) repository for disposal. The objective of the development and design study was to identify an effective chemical process and to design equipment to decontaminate the HLW glass canisters to limits that meet U.S. DOE requirements. The equipment includes canister-capping and smear stations in addition to the decontamination module and associated services

Development and design application of cerium (IV) decontamination process

International Nuclear Information System (INIS)

Bray, L.A.; Seay, J.M.

1988-10-01

A simple and effective method was developed for decontamination of high-level waste canisters. This method of chemical decontamination is applicable to a wide variety of contaminated equipment found in the nuclear industry. Conceptual design of the cerium [Ce(IV)] decontamination process equipment has been completed for the West Valley Demonstration Project (WVDP) vitrification facility. This remote equipment, which is the first engineering scale application of this technology, will remove surface contamination from stainless-steel (SS) containers containing high-level waste (HLW) glass prior to placing them into temporary storage and ultimate shipment to a US Department of Energy (DOE) repository for disposal. The objective of the development and design study was to identify an effective chemical process and to design equipment to decontaminate the HLW glass canisters to limits that meet USDOE requirements. The equipment includes canister-capping and smear stations in addition to the decontamination module and associated services. 2 refs., 1 fig

Sectoral Plan 'Deep Geological Disposal', Stage 2. Proposed site areas for the surface facilities of the deep geological repositories as well as for their access infrastructure. Annexes

International Nuclear Information System (INIS)

2011-12-01

In line with the provisions of the nuclear energy legislation, the sites for deep geological disposal of Swiss radioactive waste are selected in a three-stage Sectoral Plan process (Sectoral Plan for Deep Geological Disposal). The disposal sites are specified in Stage 3 of the selection process with the granting of a general licence in accordance with the Nuclear Energy Act. The first stage of the process was completed on 30 th November 2011, with the decision of the Federal Council to incorporate the six geological siting regions proposed by the National Cooperative for the Disposal of Radioactive Waste (NAGRA) into the Sectoral Plan for Deep Geological Disposal, for further evaluation in Stage 2. The decision also specifies the planning perimeters within which the surface facilities and shaft locations for the repositories will be constructed. In the second stage of the process, at least two geological siting regions each will be specified for the repository for low- and intermediate-level waste (L/ILW) and for the high-level waste (HLW) repository and these will undergo detailed geological investigation in Stage 3. For each of these potential siting regions, at least one location for the surface facility and a corridor for the access infrastructure will also be specified. NAGRA is responsible, at the beginning of Stage 2, for submitting proposals for potential locations for the surface facilities and their access infrastructure to the Federal Office of Energy (SFOE); these are then considered by the regional participation bodies in the siting regions. The general report and the present annexes volume document these proposals. In Stage 2, under the lead of the SFOE, socio-economic-ecological studies will also be carried out to investigate the impact of a repository project on the environment, economy and society. The present reports also contain the input data to be provided by NAGRA for the generic (site-independent) part of these impact studies. A meaningful

Numerical and experimental investigations on the time dependent behavior of a salt dome with a high-level waste repository

International Nuclear Information System (INIS)

Prij, J.; Vons, L.H.

1984-01-01

Results are presented of in-situ measurements, performed in a 300 m deep dry-drilled borehole, in the ASSE-mine. Convergence measurements at ambient as well as elevated temperatures and pressure measurements at elevated temperatures are discussed. Creep equations derived from these experiments are used for the numerical analysis of the time dependent behavior of a salt dome with a HLW repository. The analyses show that the total stresses in the salt remain compressive with deviatoric components smaller than 3 MPa. 9 references, 6 figures, 1 table

Counter current decantation washing of HLW sludge

International Nuclear Information System (INIS)

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

1997-01-01

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

Performance assessment development for a LILW repository in Slovenia

International Nuclear Information System (INIS)

Zeleznik, N.; Mele, I.

2001-01-01

Simultaneously with the site selection process for a low and intermediate level radioactive waste (LILW) repository, the preliminary assessment of the influence of the specific disposal concept on the environment and on the population was developed. The performance assessment team, organized in 1997 by ARAO, prepared several basic studies in order to clarify the objectives of the performance and safety assessment (PA/SA) procedure. In 1999 also the first performance assessment of two safety cases (surface and underground) for generic site for a LILW repository was realized. In the year 2000 activities on PA/SA analyses continued. A systematic, generic list was prepared of all possible features, events and processes (FEP list) predictable for surface or underground LILW disposal in Slovenia. Recommended and selected were the most reliable scenarios with conceptual models for LILW disposal in normal and altered evolution conditions. New verification of the obtained results was done with more powerful and accurate models for the surface repository over an aquifer of lower water permeability and an underground repository in a plastic rock. The results for both generic cases under normal evolution scenarios showed that there is a negligible dose influence on members of the critical population due to the migration of radionuclides from the foreseen LILW repository. The results of the already performed work as well as plans for the future activities are presented in the paper.(author)

A Methodology to analyze the biosphere in the assessment of deep geological repositories for high level radioactive waste

International Nuclear Information System (INIS)

Pinedo, P.; Smith, G.

1996-11-01

This report summarizes the work done and the achievements reached within the R and D Project that IMA/CIEMAT has had with ENRESA during 1993-1995. The overall R and D Project has a wide radiological protection context, but the work reported here relates only to the development of a Methodology for considering the Biosphere sub-system in the assessments of deep geological repositories for high radioactive wastes (HLW). The main areas concerned within the Methodology have to do with the Biosphere structure and morphology in the long-term relevant to deep disposal of HLW: in the contexts of the assessment of these systems, and appropriate modelling of the behaviour of radionuclides released to the biosphere system and with the associated human exposure. This document first provides a review of the past and present international and national concerns about the biosphere modelling and its importance in relation to the definition of safety criteria. A joint ENRESA/ANDRA/IPSN/CIEMAT study about the definition and practical descriptions of the biosphere systems under different climatic states is then summarized. The Methodology developed by IMA/CIEMAT is outlined with an illustration of the way it works. Different steps and procedures are included for a better practical understanding of the software tools developed within the project to support the application of the Methodology. This methodology is widely based on an international working group on ''Reference Biospheres'', part of the BIOMOVS II Project. Specific software developments have been carried out in collaboration with Qunti Sci Itd and with the Polytechnical University of Madrid. (Author)

Learning Object Repositories

Science.gov (United States)

Lehman, Rosemary

2007-01-01

This chapter looks at the development and nature of learning objects, meta-tagging standards and taxonomies, learning object repositories, learning object repository characteristics, and types of learning object repositories, with type examples. (Contains 1 table.)

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

International Nuclear Information System (INIS)

Fujiwara, Osamu; Sanga, Tomoji; Moriya, Toshifumi

2005-01-01

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

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

International Nuclear Information System (INIS)

Kitamoto, A.; Mulyanto

1993-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2009-12-30

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

Progress towards developing an international repository for radioactive waste

International Nuclear Information System (INIS)

Kurzeme, M.

2001-01-01

Pangea Resources was established to develop and promote the concept of international repositories. The concept was based on repositories located in geologically stable, flat, arid environments meeting the 'high isolation' criteria developed by Pangea Resources. The high isolation concept of Pangea Resources can be applied in various regions of the world, and these have been made public. The pilot project chosen for exploring feasibility was for an international repository in Australia. This study was in preparation for public release, when the existence of the proposal was prematurely revealed to the media. The general media and public reaction to the concept was understandably negative. However, there were also some opinion pieces reported suggesting that the concept of an international repository should be thought about seriously, before it is accepted or rejected. In addition, thoughtful reactions were received from the business and scientific communities. In particular a group of very eminent scientists, the Scientific Review Group, undertook to review independently the quality of the work done. Nevertheless, the reaction from the state and federal governments, was also negative. The Australian Government stated its policy not to permit importation or disposal of foreign nuclear waste. During the last two years Pangea Resources has been developing the technical aspects of the concept and has produced and published several scientific reports on transport, storage and disposal of radioactive materials. All the scientific studies of Pangea Resources have been reviewed by the Scientific Review Group. The work has also been presented in scientific conferences, to the media (including a dedicated television documentary) and to the public. The technical work is generic or is based on diverse potential siting areas around the world and the disposal and storage technologies studied are being broadened beyond the specific high-isolation concept. Most importantly, it has

Terminology development towards harmonizing multiple clinical neuroimaging research repositories.

Science.gov (United States)

Turner, Jessica A; Pasquerello, Danielle; Turner, Matthew D; Keator, David B; Alpert, Kathryn; King, Margaret; Landis, Drew; Calhoun, Vince D; Potkin, Steven G; Tallis, Marcelo; Ambite, Jose Luis; Wang, Lei

2015-07-01

Data sharing and mediation across disparate neuroimaging repositories requires extensive effort to ensure that the different domains of data types are referred to by commonly agreed upon terms. Within the SchizConnect project, which enables querying across decentralized databases of neuroimaging, clinical, and cognitive data from various studies of schizophrenia, we developed a model for each data domain, identified common usable terms that could be agreed upon across the repositories, and linked them to standard ontological terms where possible. We had the goal of facilitating both the current user experience in querying and future automated computations and reasoning regarding the data. We found that existing terminologies are incomplete for these purposes, even with the history of neuroimaging data sharing in the field; and we provide a model for efforts focused on querying multiple clinical neuroimaging repositories.

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

Energy Technology Data Exchange (ETDEWEB)

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

2013-11-13

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

A new integrated approach to demonstrate the safe disposal of high-level radioactive waste and spent nuclear fuel in a geological repository

International Nuclear Information System (INIS)

Mueller-Hoeppe, N.; Krone, J.; Niehues, N.; Raitz von Frentz, R.

2000-01-01

Multi-barrier systems are accepted as the basic approach for long term environmental safe isolation of radioactive waste in geological repositories. Assessing the performance of natural and engineered barriers is one of the major difficulties in producing evidence of environmental safety for any radioactive waste disposal facility, due to the enormous complexity of scenarios and uncertainties to be considered. This paper outlines a new methodological approach originally developed basically for a repository in salt, but that can be transferred with minor modifications to any other host rock formation. The approach is based on the integration of following elements: (1) Implementation of a simple method and efficient criteria to assess and prove the tightness of geological and engineered barriers; (2) Using the method of Partial Safety Factors in order to assess barrier performance at certain reasonable level of confidence; (3) Integration of a diverse geochemical barrier in the near field of waste emplacement limiting systematically the radiological consequences from any radionuclide release in safety investigations and (4) Risk based approach for the assessment of radionuclide releases. Indicative calculations performed with extremely conservative assumptions allowed to exclude any radiological health consequences from a HLW repository in salt to a reference person with a safety level of 99,9999% per year. (author)

Aspects of igneous activity significant to a repository at Yucca Mountain, Nevada

International Nuclear Information System (INIS)

Krier, D.J.; Perry, F.V.

2004-01-01

Location, timing, volume, and eruptive style of post-Miocene volcanoes have defined the volcanic hazard significant to a proposed high-level radioactive waste (HLW) and spent nuclear fuel (SNF) repository at Yucca Mountain, Nevada, as a low-probability, high-consequence event. Examination of eruptive centers in the region that may be analogueues to possible future volcanic activity at Yucca Mountain have aided in defining and evaluating the consequence scenarios for intrusion into and eruption above a repository. The probability of a future event intersecting a repository at Yucca Mountain has a mean value of 1.7 x 10 -8 per year. This probability comes from the Probabilistic Volcanic Hazard Assessment (PVHA) completed in 1996 and updated to reflect change in repository layout. Since that time, magnetic anomalies representing potential buried volcanic centers have been identified fiom magnetic surveys; however these potential buried centers only slightly increase the probability of an event intersecting the repository. The proposed repository will be located in its central portion of Yucca Mountain at approximately 300m depth. The process for assessing performance of a repository at Yucca Mountain has identified two scenarios for igneous activity that, although having a very low probability of occurrence, could have a significant consequence should an igneous event occur. Either a dike swarm intersecting repository drifts containing waste packages, or a volcanic eruption through the repository could result in release of radioactive material to the accessible environment. Ongoing investigations are assessing the mechanisms and significance of the consequence scenarios. Lathrop Wells Cone (∼80,000 yrs), a key analogue for estimating potential future volcanic activity, is the youngest surface expression of apparent waning basaltic volcanism in the region. Cone internal structure, lavas, and ash-fall tephra have been examined to estimate eruptive volume, eruption

Developing multinational radioactive waste repositories: Infrastructural framework and scenarios of cooperation

International Nuclear Information System (INIS)

2004-10-01

Currently the management of radioactive wastes centres on national strategies for collection, treatment, interim storage and disposal. This tendency to focus exclusively on national strategies reflects the fact that radioactive waste is a sensitive political issue, making cooperation among countries difficult. It is consistent with the accepted principle that a country that enjoys the benefit of nuclear energy, or the utilization of nuclear technology, should also take full responsibility for managing the generated radioactive waste. However, there are countries whose radioactive waste volumes do not easily justify a national repository, and/or countries that do not have the resources or favourable natural conditions for waste disposal to dedicate to a national repository project or would prefer to collaborate in shared initiatives because of their economic advantages. In such cases it may be appropriate for these countries to engage in a multinational collaborative effort to ensure that they have access to a common repository, in order that they can fulfil their responsibilities for their managing wastes safely. In response to requests from several Member States expressing an interest in multinational disposal options, the IAEA produced in 1998 a TECDOC outlining the important factors to be taken into account in the process of realizing such options. These factors include for example, technical (safety), institutional (legal, safeguards), economic (financial) socio-political (public acceptance) and ethical considerations. The present report reviews the work done in the previous study, taking into account developments since its publication as well as current activities in the field of multinational repositories. The report attempts to define the concepts involved in the creation of multinational repositories, to explore the likely scenarios, to examine the conditions for successful implementation, and to point out the benefits and challenges inherent to

Waste acceptance product specifications for vitrified high-level waste forms

International Nuclear Information System (INIS)

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

1993-01-01

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

The experiment of affective web risk communication on HLW geological disposal

International Nuclear Information System (INIS)

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

2006-01-01

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

Modelling of radionuclide transport along the underground access structures of deep geological repositories

Energy Technology Data Exchange (ETDEWEB)

Poller, A. [National Cooperative for the Disposal of Radioactive Waste (NAGRA), Wettingen (Switzerland); Smith, P. [SAM Switzerland GmbH, Zuerich (Switzerland); Mayer, G.; Hayek, M. [AF-Consult Switzerland AG, Baden (Switzerland)

2014-08-15

The arrangement and sealing of the access routes to a deep geological repository for radioactive waste should ensure that any radionuclide release from the emplacement rooms during the post closure phase does not by-pass the geological barriers of the repository system to a significant extent. The base case of the present study, where realistic values for the hydraulic properties of the seals and the associated excavation damage zones were assumed, assesses to what extent this is actually the case for different layout variants (ramp and shaft access and shaft access only). Furthermore, as a test of robustness of system performance against uncertainties related to such seals and the associated excavation damage zones, the present study also considers a broad spectrum of calculation cases including the hypothetical possibility that the seals perform much more poorly than expected and to check whether, consequently, the repository tunnel system and the access structures may provide significant release pathways. The study considers a generic repository system for high-level waste (HLW repository) and for low- and intermediate-level waste (L/ILW repository), both with Opalinus Clay as the host rock. It also considers the alternative possibilities of a ramp or a shaft as the access route for material transport (waste packages, etc.) to the underground facilities. Additional shafts, e.g. for the transport of persons and for ventilation, are included in both cases. The overall modelling approach consists of three broad steps: (a) the network of tunnels and access structures is implemented in a flow model, which serves to calculate water flow rates along the tunnels and through the host rock; (b) all relevant transport paths are implemented in a radionuclide release and transport model, the water flow rates being obtained from the preceding flow model calculations; (c) individual effective dose rates arising from the radionuclides released from the considered repository

Dating of ground water: an evaluation of its use in the assessment of HLW repositories

International Nuclear Information System (INIS)

Davis, S.N.; Bentley, H.W.; Zito, R.

1984-01-01

Dating of ground water is potentially useful in the evaluation of the hydrogeologic hazards associated with proposed repositories for high-level radioactive waste in the following ways: (1) identification of areas of static ground water where regional migration of radionuclides should be minimal; (2) help with the calibration of numerical transport models; (3) estimation of water velocities; (4) help with the prediction of future natural changes in the chemistry of ground water; and (5) help with the interpretation of the Pleistocene history of a region which will have a bearing on the development of hazards not related directly to ground water. Methods judged to be most useful for dating water are hydrodynamic calculations, the use of atmospherically derived radionuclides, and the measurement of the accumulated products of the decay of certain radionuclides. Thus far, the most useful atmospherically derived radionuclide is chlorine-36 with a half-life of about 3 x 10 5 years. Iodine-129, with a half-life of 1.6 x 10 7 years, is potentially useful in dating and tracing some waters older than 5 x 10 6 years. Further study is needed, however

Preliminary report on longevity of montmorillonite clay under repository-related conditions

International Nuclear Information System (INIS)

Pusch, R.; Karnland, O.

1990-12-01

This report summarizes current work on developing complete models for predicting the chemical stability of montmorillonitic bentonite buffer materials in HLW repositories. The major degrading mechanism is the conversion of montmorillonite to hydrous mica ('illite'), the basis of the derivation of tentative models being offered by ongoing systematic hydrothermal tests, which indicate that such conversion takes place by dissolution, yielding very mobile silica, and aluminium that tends to stay in the clay voids. The key mechanism in the transformation to hydrous mica at temperatures below about 130 degrees C is concluded to be neoformation of the latter mineral, the controlling parameter being the access to potassium. For temperatures ranging between 130 and 200 degrees C, excess silica is released by beidellitization and this yields I/S mixed layer minerals on uptake of potassium, and precipitation of cementing silica in the form of amorphous silica and cristobalite. The rate of transformation of montmorillonite to hydrous mica is still controlled by the access to potassium, i.e. by the rate at which potassium enters the clay. 10 Aa minerals that can be mistaken for hydrous mica can be formed by cementing collapsed stacks together. The models appear to give montmorillonite transformation rates that are in agreement with lab results and geological examples but further calibration, primarily by using the results of ongoing hydrothermal tests in the Stripa study, are required to make them applicable to repository conditions. (au)

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

Safety of HLW shipments

International Nuclear Information System (INIS)

1998-01-01

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

Stress analysis of HLW containers advanced test work Compas project

International Nuclear Information System (INIS)

Ove Arup and Partners

1990-01-01

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

Socioeconomic impacts of large-scale developments: implications for high-level nuclear waste repositories

International Nuclear Information System (INIS)

Murdock, S.H.; Leistritz, F.L.; Hamm, R.R.

1983-01-01

High-level nuclear waste repositories will likely be located in sparsely settled rural areas in the US. These projects will significantly affect the economic, demographic, public service, fiscal, and social (the socioeconomic) dimensions of those rural areas. The impacts and means of mitigating them thus require careful analysis. This paper examines some of the potential socioeconomic impacts likely to occur in rural areas as a result of repository siting and development, and it describes some of the characteristics of mitigation programs that are likely to be necessary, if the impacts are to be addressed. Both (1) standard impacts, those resulting from the fact that, like many other large-scale developments, repositories will involve a substantial number of new workers and residents (relative to the size of existing communities), and (2) special impacts, those resulting from the fact that repositories store radioactive materials, are examined. The discussion indicates that economic, demographic, public service, fiscal, and social impacts (standard and special) of these repositories will be substantial and problematic in many cases. Unless the impacts are effectively addressed with carefully planned and well financed mitigation efforts that insure that high-quality planning information is provided to local residents, and that local residents are involved in impact planning and management throughout the siting and development process, repository siting is unlikely to be effectively and equitably achieved. 44 references

Analytical methods of heat transfer compared with numerical methods as related to nuclear waste repositories

International Nuclear Information System (INIS)

Estrada-Gasca, C.A.

1986-01-01

Analytical methods were applied to the prediction of the far-field thermal impact of a nuclear waste repository. Specifically, the transformation of coordinates and the Kirchhoff transformation were used to solve one-dimensional nonlinear heat conduction problems. Calculations for the HLW and TRU nuclear waste with initial areal thermal loadings of 12 kW/acre and 0.7 kW/acre, respectively, are carried out for various models. Also, finite difference and finite element methods are applied. The last method is used to solve two-dimensional linear and nonlinear heat conduction problems. Results of the analysis are temperature distributions and temperature histories. Explicit analytical expressions of the maximum temperature rise as a function of the system parameters are presented. The theoretical approaches predict maximum temperature increases in the overburden with an error of 10%. When the finite solid one-dimensional NWR thermal problem is solved with generic salt and HLW thermal load as parameters, the maximum temperature rises predicted by the finite difference and finite element methods had maximum errors of 2.6 and 6.7%, respectively. In all the other cases the finite difference method also gave a smaller error than the finite element method

Development of the design and reliability analysis of a seabed repository system

International Nuclear Information System (INIS)

1987-06-01

This study examines the seabed repository scheme proposed in 1979 for the long term disposal of heat generating radio-active waste and develops it to a standard sufficient to compare its reliability with the drilled emplacement and penetrator schemes. The reinforced concrete repositories contain 324 waste canisters and weigh 982 tonnes fully loaded in water. The repositories are transported up to 6000 km to the disposal area by a special purpose ship and lowered 5.5 km to the seabed on six braided nylon ropes by traction winches. Reliability of the seabed repository system, measured in terms of accidents per year involving loss of one or more canisters, was comparable with the other systems. (author)

Site Characterization and Preliminary Performance Assessment Calculation Applied To JAEA-Horonobe URL Site of Japan

International Nuclear Information System (INIS)

Lim, Doo Hyun; Hatanaka, Koichiro; Ishii, Eiichi

2010-01-01

JAEA-Horonobe Underground Research Laboratory (URL) is designed for research and development on high-level radioactive waste (HLW) repository in sedimentary rock. For a potential HLW repository, understanding and implementing fracturing and faulting system, with data from the site characterization, into the performance assessment is essential because fracture and fault will be the major conductors or barriers for the groundwater flow and radionuclide release. The objectives are i) quantitative derivation of characteristics and correlation of fracturing/faulting system with geologic and geophysics data obtained from the site characterization, and ii) preliminary performance assessment calculation with characterized site information

Site Characterization and Preliminary Performance Assessment Calculation Applied To JAEA-Horonobe URL Site of Japan

Energy Technology Data Exchange (ETDEWEB)

Lim, Doo Hyun [NE Union Hill Road, Suite 200, WA 98052 (United States); Hatanaka, Koichiro; Ishii, Eiichi [Japan Atomic Energy Agency, Hokkaido (Japan)

2010-10-15

JAEA-Horonobe Underground Research Laboratory (URL) is designed for research and development on high-level radioactive waste (HLW) repository in sedimentary rock. For a potential HLW repository, understanding and implementing fracturing and faulting system, with data from the site characterization, into the performance assessment is essential because fracture and fault will be the major conductors or barriers for the groundwater flow and radionuclide release. The objectives are i) quantitative derivation of characteristics and correlation of fracturing/faulting system with geologic and geophysics data obtained from the site characterization, and ii) preliminary performance assessment calculation with characterized site information

Technology overview of mined repositories

International Nuclear Information System (INIS)

Gimera, R.; Thirumalai, K.

1982-01-01

Mined repositories present an environmentally viable option for permanent disposal of nuclear waste. This paper reviews the state-of-the-art mining technologies and identifies technological issues and developments necessary to mine a repository in basalt. The thermal loading, isolation, and retrieval requirements of a repository present unique technological challenges unknown to conventional mining practice. The technology issues and developments required in the areas of excavation, roof and ground support, equipment development, instrumentation development, and sealing are presented. Performance assessment methods must be developed to evaluate the adequacies of technologies developed to design, construct, operate, and decommission a repository. A stepwise test-and-development approach is used in the Basalt Waste Isolation Project to develop cost-effective technologies for a repository

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

International Nuclear Information System (INIS)

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

2011-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2011-01-05

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

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

International Nuclear Information System (INIS)

Hoorelbeke, J.M.; Dourthe, M.

1986-01-01

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

Time for action. Disposal of radioactive wastes is a task regarding all of us; Zeit zum Handeln. Die Entsorgung radioaktiver Abfaelle geht uns alle an

Energy Technology Data Exchange (ETDEWEB)

NONE

2008-11-15

This report helps to the popularization of the Nagra works accomplished for the management and disposal of the radioactive wastes in Switzerland. In Switzerland the disposal of radioactive wastes is clearly regulated. The disposal programme describes the process according to which the long-term storage of radioactive wastes has to be implemented to guaranty safety for mankind and environment. The disposal of radioactive wastes in deep geological repositories is a task of national importance. Until a deep repository can get into operation, several licence stages are necessary. After the overall licence given by the Federal Council is accepted by the parliament, it may still be the subject of an optional referendum. There are three different types of radioactive wastes: the low- and intermediate-level wastes (L/ILW) and the high-level wastes (HLW). Two different deep geological repositories are needed, one for the L/ILW wastes and another one for the HLW wastes. Before the repositories begin operation, the wastes are stored on the ground in interim storage facilities. The general licence for the construction of the repositories is expected to be granted before 2020. Afterwards the construction of both repositories can be undertaken. The repository for L/ILW wastes will not start operation before 2030. The start of operation of the repository for HLW wastes is not expected before 2040. This repository will be designed in such a way that the spent fuel assemblies which are stored as HLW wastes, could be withdrawn even after the closure of the repository.

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

International Nuclear Information System (INIS)

Rutledge, V.J.; Maio, V.

2013-01-01

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

Development of database systems for safety of repositories for disposal of radioactive wastes

Energy Technology Data Exchange (ETDEWEB)

Lee, Yeong Hun; Han, Jeong Sang; Shin, Hyeon Jun; Ham, Sang Won; Kim, Hye Seong [Yonsei Univ., Seoul (Korea, Republic of)

1999-03-15

In the study, GSIS os developed for the maximizing effectiveness of the database system. For this purpose, the spatial relation of data from various fields that are constructed in the database which was developed for the site selection and management of repository for radioactive waste disposal. By constructing the integration system that can link attribute and spatial data, it is possible to evaluate the safety of repository effectively and economically. The suitability of integrating database and GSIS is examined by constructing the database in the test district where the site characteristics are similar to that of repository for radioactive waste disposal.

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

International Nuclear Information System (INIS)

Piccolo, S.F.

1999-01-01

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

Application of ArcGIS to the geoscience data management of the preselected site in Beishan high-level radioactive waste repository

International Nuclear Information System (INIS)

Zhong Xia; Wang Ju; Huang Shutao; Wang Shuhong; Gao Min

2010-01-01

The site selection of a high-level radioactive waste (HLW) repository is long-term and complicated system project. In the study, the information and data are not only related to different research fields, but also of large volume, diverse types and dispersed storage. In order to wholly manage and effectively make use of the information, the authors study how to set up the platform of the geoscience database to store, manage and apply the diverse type, based on ArcGIS, and present the data management and sharing solution. (authors)

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

International Nuclear Information System (INIS)

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

2010-12-01

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

COMSOL Multiphysics Model for HLW Canister Filling

Energy Technology Data Exchange (ETDEWEB)

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

2016-04-11

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2011-12-29

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

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

International Nuclear Information System (INIS)

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

2011-01-01

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

Geological disposal of high level radioactive waste in China: progress during 1985-2004

International Nuclear Information System (INIS)

Wang Ju; Xu Guoqing; Zheng Hualing; Fan Xianhua; Wang Chengzu; Fan Zhiwen

2005-01-01

Safe disposal of high level radioactive waste (HLW) is a challenging issue for the sustainable development of nuclear energy. The studies for the disposal of HLW in China started in 1985, the proposed goal was to build China's high level waste repository by mid-21st Century, while the waste to be disposed of will be vitrified waste, transuranic waste and small amount of spent fuel. The proposed repository was a shaft-tunnel-silo model hosted by granite in saturated zone. In the period of 1985 to 2004, progress was made in China's HLW disposal program. It was decided that 'deep geological disposal' will be used to dispose of China's HLW, while the technical strategy for the development of repository will a 3-step strategy, that includes steps of site selection and site evaluation, construction of underground research laboratory, and construction of repository. Based on nation wide screening, the Beishan area, Gansu Province, northwestern China, located in Gobi desert area with few inhabitants, integral crust structure and favorable geological and hydrogeological conditions, was selected as the most potential area for China's repository. In early 1990's, site selection for underground research laboratory was conducted, 2 sites in the suburb of Beijing were preliminarily selected as the potential sites for a 'generic underground research laboratory'. It was determined to use bentonite as backfill material for the repository, while the bentonite from Gaomiaozi deposit in Inner Mongolia was selected as potential buffer and backfill material for China's repository. The studies on the mineralogical, geotechnical, physico-mechanical and thermal properties of the Gaomiaozi bentonite have been conducting. Some parameters such as sorption radio, diffusion coefficient and dispersion coefficient of radionuclides (Np, Pu and Tc) in Beishan granite and bentonite have been obtained. A low-oxygen glove box and a device simulating the temperature, pressure and redox potential of

Towards Content Development For Institutional Digital Repository ...

African Journals Online (AJOL)

The growth in Information and Communication Technology has lead to the emergence of Institutional Digital Repository, a digital archive for the preservation and dissemination of institutional research outputs. Institutional Digital Repositories make possible global dissemination of research outputs through the use of the ...

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

On area-specific underground research laboratory for geological disposal of high-level radioactive waste in China

Directory of Open Access Journals (Sweden)

Ju Wang

2014-04-01

Full Text Available Underground research laboratories (URLs, including “generic URLs” and “site-specific URLs”, are underground facilities in which characterisation, testing, technology development, and/or demonstration activities are carried out in support of the development of geological repositories for high-level radioactive waste (HLW disposal. In addition to the generic URL and site-specific URL, a concept of “area-specific URL”, or the third type of URL, is proposed in this paper. It is referred to as the facility that is built at a site within an area that is considered as a potential area for HLW repository or built at a place near the future repository site, and may be regarded as a precursor to the development of a repository at the site. It acts as a “generic URL”, but also acts as a “site-specific URL” to some extent. Considering the current situation in China, the most suitable option is to build an “area-specific URL” in Beishan area, the first priority region for China's high-level waste repository. With this strategy, the goal to build China's URL by 2020 may be achieved, but the time left is limited.

Time-frames and the demonstration of safety for HLW disposal

International Nuclear Information System (INIS)

Watkins, B.; Kessler, J.

1999-01-01

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

Reactive transport model and apparent Kd of Ni in the near field of a HLW repository in granite

Science.gov (United States)

Lu, Chuanhe; Samper, Javier; Luis Cormenzana, José; Ma, Hongyun; Montenegro, Luis; Ángel Cuñado, Miguel

2012-12-01

Current performance assessment models for radionuclide migration through the near field of high-level radioactive waste repositories often rely on the assumption of a constant Kd for sorption. The validity of such assumption is evaluated here with a reactive transport model for Ni2+ in the near field of a repository in granite. Model results show that Ni2+ sorbs mainly by surface complexation on weak sorption sites. The apparent Kd of Ni2+, Kda, depends on the concentration of dissolved Ni and pH and is constant only when the concentration of dissolved Ni is smaller than 10-6 mol/L. The results of the sensitivity runs show that Kda is sensitive to the water flux at the bentonite-granite interface, the effective diffusion of the bentonite and the concentration of weak sorption sites of the bentonite. The competition of other nuclides such as Cs+ on Ni2+ sorption is not important. Corrosion products, however, affect significantly the sorption of Ni2+ on the bentonite. The model with a constant Kd does not reproduce the release rates of Ni2+ from the bentonite into the granite. A model with a variable Kd which depends on the concentration of dissolved Ni2+ and pH may provide an acceptable surrogate of the multicomponent reactive transport model for the conditions of the repository considered in our model. Simulations using the Kd-approach were performed with GoldSim based on the interpolation in the pH and concentration table, while the reactive transport model simulations were performed with CORE2D which incorporates multisite surface complexation.

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

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

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

International Nuclear Information System (INIS)

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

2007-02-01

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

Establishment of research and development priorities regarding the geologic disposal of nuclear waste in the United States and strategies for international collaboration

International Nuclear Information System (INIS)

McMahon, Kevin; Swift, Peter; Nutt, Mark; Peters, Mark; Williams, Jeff; Voegele, Michael; Birkholzer, Jens

2011-01-01

The U.S. Department of Energy Office of Nuclear Energy (DOE-NE), Office of Fuel Cycle Technologies (OFCT) has established the Used Fuel Disposition Campaign (UFDC) to conduct research and development (R and D) activities related to storage, transportation and disposal of low level waste (LLW), used nuclear fuel (UNF) and high level radioactive waste (HLW). The U.S. has, for the past twenty-plus years, focused efforts on disposing spent nuclear fuel (SNF) and HLW in a geologic repository at Yucca Mountain Nevada. The recent decision by the U.S. DOE to no longer pursue the development of that repository has necessitated investigating alternative concepts for the disposal of SNF and HLW that exists today and that could be generated under future fuel cycles. The disposal of SNF and HLW in a range of geologic media has been investigated internationally. Considerable progress has been made by in the U.S and other nations, but gaps in knowledge still exist. The U.S. national laboratories have participated in these programs and have conducted R and D related to these issues to a limited extent. However, a comprehensive R and D program investigating a variety of storage, geologic media and disposal concepts has not been a part of the U.S. waste management program since the mid 1980s. Such a comprehensive R and D program has been developed in the UFDC using a systematic approach to identify potential R and D opportunities. This paper will describe the process used by the UFDC and summarize the R and D being pursued. The U.S. DOE has cooperated and collaborated with other countries in many different 'arenas' including the Nuclear Energy Agency (NEA) within the Organisation for Economic Co-operation and Development (OECD), the International Atomic Energy Agency (IAEA), and through bilateral agreements with other countries. These international activities benefited the DOE through the acquisition and exchange of information, database development, and peer reviews by experts from

Repository simulation model: Final report

International Nuclear Information System (INIS)

1988-03-01

This report documents the application of computer simulation for the design analysis of the nuclear waste repository's waste handling and packaging operations. The Salt Repository Simulation Model was used to evaluate design alternatives during the conceptual design phase of the Salt Repository Project. Code development and verification was performed by the Office of Nuclear Waste Isolation (ONWL). The focus of this report is to relate the experience gained during the development and application of the Salt Repository Simulation Model to future repository design phases. Design of the repository's waste handling and packaging systems will require sophisticated analysis tools to evaluate complex operational and logistical design alternatives. Selection of these design alternatives in the Advanced Conceptual Design (ACD) and License Application Design (LAD) phases must be supported by analysis to demonstrate that the repository design will cost effectively meet DOE's mandated emplacement schedule and that uncertainties in the performance of the repository's systems have been objectively evaluated. Computer simulation of repository operations will provide future repository designers with data and insights that no other analytical form of analysis can provide. 6 refs., 10 figs

Life cycle assessment of geological repositories for the final disposal of spent fuel in Finland and Sweden

International Nuclear Information System (INIS)

Puhrer, A.; Bauer, C.

2014-01-01

This paper presents a Life Cycle Assessment (LCA) of the geological repositories for the final disposal of spent nuclear fuel in Finland and Sweden. A separate LCA has been performed for the geological spent fuel repository in each country and the results have been compared. A further benchmark comparison has been made with the LCA of the Swiss geological repository for high-level waste and spent fuel. The life cycle inventory (LCI) product system boundaries include the spent fuel repository and encapsulation facility in each country. All materials, processes, consumed utilities and transport associated with the construction, operation and closure of the repositories for spent fuel are included in the LCI. The life cycle impact assessment (LCIA) is performed using two methods: IPCC 2007 Climate Change and ReCiPe. These assessment methods return results pertaining to global warming potential (GWP) as well as a number of environmental impact categories such as human toxicity and natural land transformation. Results indicate that the use of copper for disposal canister fabrication and bentonite for repository backfilling are the causes for most of the environmental impact of the spent fuel repositories in Finland and Sweden. Alternate, less bentonite-intensive backfilling scenarios may mitigate this impact. While the Swiss bentonite consumption is lower and no copper is used for canister fabrication, the Swiss electricity and fuel consumption associated with final disposal of high-level waste and spent fuel is significantly higher than in Finland or Sweden. Approximately 1 g CO 2 -eq is emitted due to the final disposal of spent fuel and HLW per kWh of nuclear generated electricity. This represents some 10% of the emissions due to the entire nuclear energy chain and is practically negligible in the context of GHG emissions of other energy technologies. (authors)

5th Total System Performance Assessment Workshop

International Nuclear Information System (INIS)

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

2009-07-01

Research items on safety assessment of high-level waste repository have been proposed by external invited experts outside KAERI and discussed extensively during the annual 5th performance assessment workshop prepared by safety assessment group in KAERI. This could be useful to set up R and D plans necessary for the next phase of mid- and long-term reaserch area regarding the safety assessment of high-level waste repository. Through the research and the presentation, HLW-related research and development area including such specific research items as current status of HLW safety assessment research, current requirement for the licensing of the repository system, priority on research area, data base building for the safety assessment, source-term modeling as well as safety case, among many others, have been discussed and summarized

Robust record preservation system on geological repository

International Nuclear Information System (INIS)

Ohuchi, J.; Torata, S.; Tsuboya, T.

2004-01-01

Long-term record preservation system on geological disposal of High Level Radioactive Wastes (HLW) has been investigated as the institutional control by RWMC, Japan. Geological disposal of HLW, being based on the passive safe concept, has been considered not to necessitate the human controls to maintain its long-term safety. However how to complement the safety case on geological disposal is an important issue in each countries to progress the repository program with the step-wise decisions process during the long-term period up to several hundreds years. Although we cannot predict the future society, we need to realize the robust and redundant system for preserving records, which should be accessible, retrievable and understandable for the unpredicted future generations. First of all, we held a Rome workshop in January 2003 to exchange views on the matter, resulted in the suggestion directing the discussion on the record management and long-term preservation and retrieval of information regarding radioactive waste. Second, we considered the balance of active and passive system to strengthen the robustness. Another significance of long-term record preservation is to send current generation an implicit message, 'doing our best for future generations', in addition to aiming at both warning and their own decision-making. We call it 'meta-signal' to current generation. Thirdly, we demonstrated the laser-engraving technology to have converted five hundreds pages of an A4 sized report with human readable font sizes to 42 square silicon carbide plates, 10cm x10cm and 1mm in thickness. Silicon carbide would be an alternative to paper and might be possible to be an alternative to microfilm utilized as digital recording media. Another case study is the future generations' accessibility to the preserved records. (author)

The repository ecology an approach to understanding repository and service interactions

CERN Document Server

CERN. Geneva; Hagemann, Melissa

2007-01-01

An increasing number of university institutions and other organisations are deciding to deploy repositories and a growing number of formal and informal distributed services are supporting or capitalising on the information these repositories provide. Despite reasonably well understood technical architectures, early majority adopters may struggle to articulate their place within the actualities of a wider information environment. The idea of a repository ecology provides developers and administrators with a useful way of articulating and analysing their place in the information environment, and the technical and organisational interactions they have, or are developing, with other parts of such an environment. This presentation will provide an overview of the concept of a repository ecology and examine some examples from the domains of scholarly communications and elearning.

Characteristics of potential repository wastes

International Nuclear Information System (INIS)

1992-07-01

This document, and its associated appendices and microcomputer (PC) data bases, constitutes the reference OCRWM data base of physical and radiological characteristics data of radioactive wastes. This Characteristics Data Base (CDB) system includes data on spent nuclear fuel and high-level waste (HLW), which clearly require geologic disposal, and other wastes which may require long-term isolation, such as sealed radioisotope sources. The data base system was developed for OCRWM by the CDB Project at Oak Ridge National Laboratory. Various principal or official sources of these data provided primary information to the CDB Project which then used the ORIGEN2 computer code to calculate radiological properties. The data have been qualified by an OCRWM-sponsored peer review as suitable for quality-affecting work meeting the requirements of OCRWM's Quality Assurance Program. The wastes characterized in this report include: light-water reactor (LWR) spent fuel and immobilized HLW

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

International Nuclear Information System (INIS)

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

2011-01-01

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

In-situ test programs related to design and construction of high-level nuclear waste (HLW) deep geologic repositories. Final report (Task 2), June 1981-November 1982

International Nuclear Information System (INIS)

Roberds, W.; Bauhof, F.; Gonano, L.

1983-03-01

The media and sites considered include (1) basalt at Hanford, Washington; (2) tuff at Yucca Mountain, Nevada Test Site; (3) domal salt at specific Gulf Coast sites; (4) bedded salt at an unspecified site; and (5) granite at an unspecified site. A licensing perspective is outlined and a defensible rationale developed and utilized for the test selection process. This rationale essentially consists of: establishing the information needs for construction authorization; assessing the relevant capabilities of available tests; and matching the capabilities of specific tests to the perceived information needs. The information needs at any time consist of the additional information (if any) needed in order to predict satisfactory repository system performance with the required level of confidence, and thus are a function of: the significance of the repository engineered components and site characteristics to system performance; the currently available information, which may be supplemented with time; and the acceptable level of confidence in satisfactory performance for each licensing step. Determination of the acceptable levels of confidence and the significance of repository system components is outside the scope of this report. Suitable assumptions have thus been made regarding the development of information needs for construction authorization by the time of initial site submittals. Tests which are available and respond to the perceived media/site specific information needs, either by simulation or assessment of site characteristics, are identified and their capabilities assessed. Specific in situ tests are investigated and described in detail. Research and development which might be effective in improving test capabilities have been recommended

Development of the geologic waste disposal programme in the United States of America

International Nuclear Information System (INIS)

Coffman, F.E.; Ballard, W.W.; Carbiener, W.A.

1983-01-01

Although alternative concepts are being studied as future options, over at least the next few decades the United States of America is committed to the disposal of commercially generated high-level and transuranic nuclear waste (HLW and TRU) in mined geologic repositories. A 10,000-year minimum isolation period is sought. Responsibility for the management and disposal of United States nuclear waste, in accordance with standards and regulations established, respectively, by the Environmental Protection Agency (EPA) and the Nuclear Regulatory Commission (NRC), resides with the Department of Energy (DOE). The DOE National Waste Terminal Storage (NWTS) Program has been implemented to provide the facilities and develop the requisite technology for the disposal of HLW and TRU. The NWTS Program is highly structured, adequately funded, and realistically scheduled. The timely realization of its objectives is basic to the furtherance of the new national energy policy being defined by President Reagan and the United States Congress. The first NWTS repository is scheduled to be operational as early as 1998. The host-rock formation, selected on the basis of the results of at-depth investigations via exploratory shafts to be sunk in 1983-1985 at three potential sites previously extensively characterized by surface techniques, will be either basalt, volcanic tuff, or domed or bedded salt. Selection of one site in these formations will not necessarily disqualify others. Also, screening studies of granitic formations in the United States for the siting of later, regionally located repositories are currently being conducted. Each NWTS repository will be licensed by the NRC. The first application for a construction authorization will probably be submitted in 1988. The application will be submitted for a site to be selected in 1987

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

International Nuclear Information System (INIS)

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

1998-01-01

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

Testing to evaluate the suitability of waste forms developed for electrometallurgically treated spent sodium-bonded nuclear fuel for disposal in the Yucca Mountain reporsitory.

Energy Technology Data Exchange (ETDEWEB)

Ebert, W. E.

2006-01-31

The results of laboratory testing and modeling activities conducted to support the development of waste forms to immobilize wastes generated during the electrometallurgical treatment of spent sodium-bonded nuclear fuel and their qualification for disposal in the federal high-level radioactive waste repository are summarized in this report. Tests and analyses were conducted to address issues related to the chemical, physical, and radiological properties of the waste forms relevant to qualification. These include the effects of composition and thermal treatments on the phase stability, radiation effects, and methods for monitoring product consistency. Other tests were conducted to characterize the degradation and radionuclide release behaviors of the ceramic waste form (CWF) used to immobilize waste salt and the metallic waste form (MWF) used to immobilize metallic wastes and to develop models for calculating the release of radionuclides over long times under repository-relevant conditions. Most radionuclides are contained in the binder glass phase of the CWF and in the intermetallic phase of the MWF. The release of radionuclides from the CWF is controlled by the dissolution rate of the binder glass, which can be tracked using the same degradation model that is used for high-level radioactive waste (HLW) glass. Model parameters measured for the aqueous dissolution of the binder glass are used to model the release of radionuclides from a CWF under all water-contact conditions. The release of radionuclides from the MWF is element-specific, but the release of U occurs the fastest under most test conditions. The fastest released constituent was used to represent all radionuclides in model development. An empirical aqueous degradation model was developed to describe the dependence of the radionuclide release rate from a MWF on time, pH, temperature, and the Cl{sup -} concentration. The models for radionuclide release from the CWF and MWF are both bounded by the HLW glass

The repository ecology: an approach to understanding repository and service interactions

CERN Multimedia

CERN. Geneva

2007-01-01

An increasing number of university institutions and other organisations are deciding to deploy repositories and a growing number of formal and informal distributed services are supporting or capitalising on the information these repositories provide. Despite reasonably well understood technical architectures, early majority adopters may struggle to articulate their place within the actualities of a wider information environment. The idea of a repository ecology provides developers and administrators with a useful way of articulating and analysing their place in the information environment, and the technical and organisational interactions they have, or are developing, with other parts of such an environment. This presentation will provide an overview of the concept of a repository ecology and examine some examples from the domains of scholarly communications and elearning. View John Robertson's biography

Repository operational criteria analysis

International Nuclear Information System (INIS)

Hageman, J.P.; Chowdhury, A.H.

1992-08-01

The objective of the ''Repository Operational Criteria (ROC) Feasibility Studies'' (or ROC task) was to conduct comprehensive and integrated analyses of repository design, construction, and operations criteria in 10 CFR Part 60 regulations, considering the interfaces and impacts of any potential changes to those regulations. The study addresses regulatory criteria related to the preclosure aspects of the geologic repository. The study task developed regulatory concepts or potential repository operational criteria (PROC) based on analysis of a repository's safety functions and other regulations for similar facilities. These regulatory concepts or PROC were used as a basis to assess the sufficiency and adequacy of the current criteria in 10 CFR Part 60. Where the regulatory concepts were same as current operational criteria, these criteria were referenced. The operations criteria referenced or the PROC developed are given in this report. Detailed analyses used to develop the regulatory concepts and any necessary PROC for those regulations that may require a minor change are also presented. The results of the ROC task showed a need for further analysis and possible major rule change related to the design bases of a geologic repository operations area, siting, and radiological emergency planning

Repository Rodeo Redux

CERN Document Server

Anez, Melissa; Donohue, Tim; Fyson, Will; Simko, Tibor; Wilcox, David

2017-01-01

You’ve got more repository questions and we’ve got more answers! Last year’s Repository Rodeo panel was a huge success, so we’re taking the show on the road to Brisbane for OR2017. Join representatives from the DSpace, Eprints, Fedora, Hydra, and Islandora communities as we (briefly) explain what each of our repositories actually does. We'll also talk about the directions of our respective technical and community developments, and related to the conference theme of Open: Innovation Knowledge Repositories, offer brief observations about the latest, most promising and/or most surprising innovations in our space. This panel will be a great opportunity for newcomers to Open Repositories to get a crash course on the major repository options and meet representatives from each of their communities. After a brief presentation from each representative, we'll open the session up for questions from the audience.

Managing and Evaluating Digital Repositories

Science.gov (United States)

Zuccala, Alesia; Oppenheim, Charles; Dhiensa, Rajveen

2008-01-01

Introduction: We examine the role of the digital repository manager, discuss the future of repository management and evaluation and suggest that library and information science schools develop new repository management curricula. Method: Face-to-face interviews were carried out with managers of five different types of repositories and a Web-based…

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

International Nuclear Information System (INIS)

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

2015-01-01

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

A literature review of coupled thermal-hydrologic-mechanical-chemical processes pertinent to the proposed high-level nuclear waste repository at Yucca Mountain

International Nuclear Information System (INIS)

Manteufel, R.D.; Ahola, M.P.; Turner, D.R.; Chowdhury, A.H.

1993-07-01

A literature review has been conducted to determine the state of knowledge available in the modeling of coupled thermal (T), hydrologic (H), mechanical (M), and chemical (C) processes relevant to the design and/or performance of the proposed high-level waste (HLW) repository at Yucca Mountain, Nevada. The review focuses on identifying coupling mechanisms between individual processes and assessing their importance (i.e., if the coupling is either important, potentially important, or negligible). The significance of considering THMC-coupled processes lies in whether or not the processes impact the design and/or performance objectives of the repository. A review, such as reported here, is useful in identifying which coupled effects will be important, hence which coupled effects will need to be investigated by the US Nuclear Regulatory Commission in order to assess the assumptions, data, analyses, and conclusions in the design and performance assessment of a geologic reposit''. Although this work stems from regulatory interest in the design of the geologic repository, it should be emphasized that the repository design implicitly considers all of the repository performance objectives, including those associated with the time after permanent closure. The scope of this review is considered beyond previous assessments in that it attempts with the current state-of-knowledge) to determine which couplings are important, and identify which computer codes are currently available to model coupled processes

A literature review of coupled thermal-hydrologic-mechanical-chemical processes pertinent to the proposed high-level nuclear waste repository at Yucca Mountain

Energy Technology Data Exchange (ETDEWEB)

Manteufel, R.D.; Ahola, M.P.; Turner, D.R.; Chowdhury, A.H. [Southwest Research Inst., San Antonio, TX (United States). Center for Nuclear Waste Regulatory Analyses

1993-07-01

A literature review has been conducted to determine the state of knowledge available in the modeling of coupled thermal (T), hydrologic (H), mechanical (M), and chemical (C) processes relevant to the design and/or performance of the proposed high-level waste (HLW) repository at Yucca Mountain, Nevada. The review focuses on identifying coupling mechanisms between individual processes and assessing their importance (i.e., if the coupling is either important, potentially important, or negligible). The significance of considering THMC-coupled processes lies in whether or not the processes impact the design and/or performance objectives of the repository. A review, such as reported here, is useful in identifying which coupled effects will be important, hence which coupled effects will need to be investigated by the US Nuclear Regulatory Commission in order to assess the assumptions, data, analyses, and conclusions in the design and performance assessment of a geologic reposit``. Although this work stems from regulatory interest in the design of the geologic repository, it should be emphasized that the repository design implicitly considers all of the repository performance objectives, including those associated with the time after permanent closure. The scope of this review is considered beyond previous assessments in that it attempts with the current state-of-knowledge) to determine which couplings are important, and identify which computer codes are currently available to model coupled processes.

Effects of post-disposal gas generation in a repository for spent fuel, high-level waste and long-lived intermediate level waste sited in opalinus clay

International Nuclear Information System (INIS)

Johnson, L.; Marschall, P.; Zuidema, P.; Gribi, P.

2004-07-01

This comprehensive report issued by the Swiss National Cooperative for the Disposal of Radioactive Waste NAGRA takes a look at post-disposal gas generation in a repository for spent fuel and highly radioactive wastes in Opalinus clay strata. This study provides a comprehensive treatment of the issue of gas generation in a repository for spent fuel (SF), vitrified high-level waste (HLW) and long-lived intermediate-level waste (ILW), sited in the Opalinus clay of the Zuercher Weinland in northern Switzerland. The issue of how gas generation in and transport from waste repositories may influence disposal system performance has been under study for many years, both at Nagra and internationally. The report consists of three main parts: (i) A synthesis of basic information on the host rock and on details of repository construction; (ii) A discussion on gas transport characteristics of the engineered barrier system and the geosphere; (iii) A discussion on the effects of gas on system performance, based on the available information on gas generation, gas transport properties and gas pathways provided in the previous parts of the report. Simplified model calculations based on a mass balance approach for the gas generated within the repository are presented and discussed

HLW immobilization in glass: industrial operation and product quality

International Nuclear Information System (INIS)

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

1992-01-01

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

Extreme ground motions and Yucca Mountain

Science.gov (United States)

Hanks, Thomas C.; Abrahamson, Norman A.; Baker, Jack W.; Boore, David M.; Board, Mark; Brune, James N.; Cornell, C. Allin; Whitney, John W.

2013-01-01

Yucca Mountain is the designated site of the underground repository for the United States' high-level radioactive waste (HLW), consisting of commercial and military spent nuclear fuel, HLW derived from reprocessing of uranium and plutonium, surplus plutonium, and other nuclear-weapons materials. Yucca Mountain straddles the western boundary of the Nevada Test Site, where the United States has tested nuclear devices since the 1950s, and is situated in an arid, remote, and thinly populated region of Nevada, ~100 miles northwest of Las Vegas. Yucca Mountain was originally considered as a potential underground repository of HLW because of its thick units of unsaturated rocks, with the repository horizon being not only ~300 m above the water table but also ~300 m below the Yucca Mountain crest. The fundamental rationale for a geologic (underground) repository for HLW is to securely isolate these materials from the environment and its inhabitants to the greatest extent possible and for very long periods of time. Given the present climate conditions and what is known about the current hydrologic system and conditions around and in the mountain itself, one would anticipate that the rates of infiltration, corrosion, and transport would be very low—except for the possibility that repository integrity might be compromised by low-probability disruptive events, which include earthquakes, strong ground motion, and (or) a repository-piercing volcanic intrusion/eruption. Extreme ground motions (ExGM), as we use the phrase in this report, refer to the extremely large amplitudes of earthquake ground motion that arise at extremely low probabilities of exceedance (hazard). They first came to our attention when the 1998 probabilistic seismic hazard analysis for Yucca Mountain was extended to a hazard level of 10-8/yr (a 10-4/yr probability for a 104-year repository “lifetime”). The primary purpose of this report is to summarize the principal results of the ExGM research program

Radioactive waste repository of high ecological safety

International Nuclear Information System (INIS)

Sobolev, I.; Barinov, A.; Prozorov, L.

2000-01-01

With the purpose to construct a radioactive waste repository of high ecological safety and reliable containment, MosNPO 'Radon' specialists have developed an advanced type repository - large diameter well (LBD) one. A project is started for the development of a technology for LDW repository construction and pilot operation of the new repository for 25-30 years. The 2 LDW repositories constructed at the 'Radon' site and the developed monitoring system are described

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

International Nuclear Information System (INIS)

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

1986-01-01

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

NWTS program criteria for mined geologic disposal of nuclear waste: repository performance and development criteria. Public draft

Energy Technology Data Exchange (ETDEWEB)

None

1982-07-01

This document, DOE/NWTS-33(3) is one of a series of documents to establish the National Waste Terminal Storage (NWTS) program criteria for mined geologic disposal of high-level radioactive waste. For both repository performance and repository development it delineates the criteria for design performance, radiological safety, mining safety, long-term containment and isolation, operations, and decommissioning. The US Department of Energy will use these criteria to guide the development of repositories to assist in achieving performance and will reevaluate their use when the US Nuclear Regulatory Commission issues radioactive waste repository rules.

NWTS program criteria for mined geologic disposal of nuclear waste: repository performance and development criteria. Public draft

International Nuclear Information System (INIS)

1982-07-01

This document, DOE/NWTS-33(3) is one of a series of documents to establish the National Waste Terminal Storage (NWTS) program criteria for mined geologic disposal of high-level radioactive waste. For both repository performance and repository development it delineates the criteria for design performance, radiological safety, mining safety, long-term containment and isolation, operations, and decommissioning. The US Department of Energy will use these criteria to guide the development of repositories to assist in achieving performance and will reevaluate their use when the US Nuclear Regulatory Commission issues radioactive waste repository rules

Development of repository-wide radionuclide transport model considering the effects of multiple sources

International Nuclear Information System (INIS)

Hatanaka, Koichiro; Watari, Shingo; Ijiri, Yuji

1999-11-01

Safety assessment of the geological isolation system according to the groundwater scenario has traditionally been conducted based on the signal canister configuration and then the safety of total system has been evaluated based on the dose rates which were obtained by multiplying the migration rates released from the engineered barrier and/or the natural barrier by dose conversion factors and total number of canisters disposed in the repository. The dose conversion factors can be obtained from the biosphere analysis. In this study, we focused on the effect of multiple sources due to the disposal of canisters at different positions in the repository. By taking the effect of multiple sources into consideration, concentration interference in the repository region is possible to take place. Therefore, radionuclide transport model/code considering the effect of concentration interference due to the multiple sources was developed to make assessments of the effect quantitatively. The newly developed model/code was verified through the comparison analysis with the existing radionuclide transport analysis code used in the second progress report. In addition, the effect of the concentration interference was evaluated by setting a simple problem using the newly developed analysis code. This results shows that the maximum park value of the migration rates from the repository was about two orders of magnitude lower than that based on single canister configuration. Since the analysis code was developed by assuming that all canisters disposed of along the one-dimensional groundwater flow contribute to the concentration interference in the repository region, the assumption should be verified by conducting two or three-dimensional analysis considering heterogeneous geological structure as a future work. (author)

The Yucca Mountain Repository - Too Little, Too Late

International Nuclear Information System (INIS)

Eriksson, L.G.; Pentz, D.L.

2009-01-01

In 2008, the U.S. Department of Energy (US DOE) announced that the nation's first (and only pursued) deep geological disposal system (repository) for 70,000 metric tonnes of spent nuclear fuel (SNF) and other high-level radioactive waste (HLW) at the Yucca Mountain (YM) site in Nevada would: 1. Not be able to accommodate the projected stockpile of utility-generated SNF beyond 2010. 2. Open no earlier than in 2020, i.e., more than 22 years behind the statutory-mandated opening date. In the meantime, the US DOE is legally obligated to compensate the utilities from January 31, 1998, until it takes title to the utilities' SNF. In 2005 when the YM SNF repository was projected to open in 2010, the utilities estimated that, depending upon how close to 2010 the YM repository opened, the 'breach-of-contract' compensation could be in the range of between 100 billion and 300 billion U.S. dollars ($300 B), which would exceed the 2008 projected life-cycle cost of $96 B for the YM repository. It thus seems appropriate to look beyond the YM repository and call upon the U.S. Congress to promptly act and open new avenues allowing the US DOE to more timely and cost-effectively take title to both existing and pending SNF the current fleet of 104 reactors will generate through the next 60 years. Options for SNF arising from an additional 50 reactors should also be provided. Based on our more than 60 years of combined involvement in nuclear waste management in the USA and abroad, we submit the following industrial-scale-proven, repository-related, nuclear-waste-management and disposition solutions for prompt Congressional consideration and action: 1. An increase in the disposal capacity (and perhaps mission) of the YM repository. 2. Prompt establishment of at least one large federal monitored retrievable storage (MRS) facility for utility-generated SNF. 3. Continued research in reprocessing options of existing and pending SNF with defined milestones. 4. Resurrection of a second

Methodology for the biosphere analysis in the evaluation of deep geological repositories for high radioactive waste

International Nuclear Information System (INIS)

Cancio, D.; Pinedo, P.; Aguero, A.; Simon, I.; Torres, C.; Robles, B.; Smith, G.M.; Little, R.; Watkings, B.; Brice, A.; Jaen, J.A.; Coronado, S.

1997-01-01

This report summarizes the work done and the achievements reached within the R and D Project that IMA/CIEMAT has had with ENRESA during 1993-1995. The overal R and D Project has a wide radiological protection context, but the work reported here relates only to the development of a Methodology for considering the Biosphere sub-system in the assessments of deep geological repositories for high radioactive wastes (HLW). The main areas concerned within the Methodology have to do with Biosphere structure and morphology in the long-term relevant to deep disposal of HLW: in the contexts of the assessment of these systems, and appropiate modelling of the behaviour of radionuclides released to the biosphere system and with the associated human exposure. This document first provides a review of the past and present international and national concerns about the biosphere modelling and its importance in relation to the definition of safety criteria. A joint ENRESA/ANDRA/IPSN/CIEMAT study about the definition and proactical descriptions of the biosphere systems under different climatic states is then summarized. The Methodology developed by IMA/CIEMAT is outlined with an illustration of the way it works. Different steps and procedures are included for a better proactical understanding of the software tools developed within the project to support the application of the Methologoy. This Methodology is widely based on an international working group on Reference Biospheres part national work for ENRESA has been supported under a collaborative agreement with QuantiSci Ltd. Specific software development have been carried out in collaboration with QuantiSci Ltd and with the Polytechnical University of Madrid. Most of the items included within the Methodology and moreover the Methodology as a whole, follows a continuos progressive development. It is increasinaly recognized that assessment capabilities, establisment of safety criteria and regulatory framework and the steps in a

EEI/UWASTE oversight of the DOE Repository Program by the Repository Information Exchange Team

International Nuclear Information System (INIS)

Henkel, C.J.; Supko, E.M.; Schwartz, M.H.

1993-01-01

The Utility Nuclear Waste and Transportation Program of the Edison Electric Institute (EEI/UWASTE) has conducted reviews of the US DOE's repository program through its Repository Information Exchange Team (RIET or Team). Eight such reviews have been conducted since 1985 covering topics that include repository program management and control; repository schedule; repository budget; quality assurance; site characterization; repository licensing; environmental issues; and institutional and public information activities. The utility industry has used these repository program reviews as a forum for providing DOE's Office of Civilian Radioactive Waste Management (OCRWM) with comments on the direction of the repository program, advice for future actions regarding quality assurance activities and repository licensing, and suggestions for management and control of the Repository Program. The most significant recommendations made by the utility industry through the RIET are discussed along with any subsequent action by OCRWM in response to or subsequent to utility industry recommendations. The process used by the RIET to develop its recommendations to OCRWM regarding the repository program is also discussed

Effects of temperature, temperature gradients, stress, and irradiation on migration of brine inclusions in a salt repository

International Nuclear Information System (INIS)

Jenks, G.H.

1979-07-01

Available experimental and theoretical information on brine migration in bedded salt are reviewed and analyzed. The effects of temperature, thermal gradients, stress, irradiation, and pressure in a salt repository are among the factors considered. The theoretical and experimental (with KCl) results of Anthony and Cline were used to correlate and explain the available data for rates of brine migration at temperatures up to 250 0 C in naturally occurring crystals of bedded salt from Lyons and Hutchinson, Kansas. Considerations of the effects of stressing crystals of bedded salt on the migratin properties of brine inclusions within the crystals led to the conclusion that the most probable effects are a small fractional increase in the solubility of the salt within the liquid and a concomitant and equal fractional increase in the rate of the thermal gradient-induced migration of the brine. The greatest uncertainty relative to the prediction of rates of migration of brine into a waste emplacement cavity in bedded salt is associated with questions concerning the effects of the grain boundaries (within the aggregates of single crystals which comprise a bedded salt deposit) on brine migration through the deposit. The results of some of the estimates of rates and total amounts of brine inflow to HLW and SURF waste packages emplaced in bedded salt were included to illustrate the inflow volumes which might occur in a repository. The results of the brine inflow estimates for 10-year-old HLW emplaced at 150 kW/acre indicated inflow rates starting at 0.7 liter/year and totaling 12 liters at 30 years after emplacement. The results of the estimates for 10-year-old PWR SURF emplaced at 60 kW/acre indicated a constant inflow of 0.035 liter/year for the first 35 years after emplacement

Geochemical Processes and compacted bentonite FEBEX with a thermohydraulic gradient with a thermohydraulic gradient; Procesos geoquimicos y modificaciones texturales en bentonita FEBEX compactada sometida a un gradiente termohidraulico

Energy Technology Data Exchange (ETDEWEB)

Leguey Jimenez, S; Cuevas Rodriguez, J; Martin Barca, M; Vigil de la Villa Mencia, R.; Ramirez Martin, S; Garcia Gimenez, R [Universidad Autonoma de Madrid (Spain)

2002-07-01

At present, the main source of High Level radioactive Waste (HLW) is the electrical energy production during all sep of developing. In almost all the countries with nuclear programs, the option for the final management of HLW is the Deep Geological Repository (DGR), based on the concept of multi barrier. According to this concept, the wastes is isolated from biosphere by the interposition of confinement barrier. In the context of an investigation of the near field for a repository of HLW, the FEBEX Project, a set of laboratory test has been designed to give a better understanding of the thermo-hydro-mechanical and geochemical behaviour of the compacted bentonite as a confinement barrier. The object of these work is to analyse the properties of the bentonite and its behaviour under conditions that will be found in a repository. The precipitation of mineral phases, due to local changes in the chemical equilibrium and the hydration itself, can produce changes in the salinity of the interstitial water and in the microstructural organisation of the clay particles. the hydraulic and mechanical properties of the bentonite can be modified by the special conditions of the barrier. (Author)

Geochemical processes and compacted bentonite FEBEX with a thermohydraulic gradient with a thermohydraulic gradient

International Nuclear Information System (INIS)

Leguey Jimenez, S.; Cuevas Rodriguez, J.; Martin Barca, M.; Vigil de la Villa Mencia, R.; Ramirez Martin, S.; Garcia Gimenez, R.

2002-01-01

At present, the main source of High Level radioactive Waste (HLW) is the electrical energy production during all sep of developing. In almost all the countries with nuclear programs, the option for the final management of HLW is the Deep Geological Repository (DGR), based on the concept of multi barrier. According to this concept, the wastes is isolated from biosphere by the interposition of confinement barrier. In the context of an investigation of the near field for a repository of HLW, the FEBEX Project, a set of laboratory test has been designed to give a better understanding of the thermo-hydro-mechanical and geochemical behaviour of the compacted bentonite as a confinement barrier. The object of these work is to analyse the properties of the bentonite and its behaviour under conditions that will be found in a repository. The precipitation of mineral phases, due to local changes in the chemical equilibrium and the hydration itself, can produce changes in the salinity of the interstitial water and in the microstructural organisation of the clay particles. the hydraulic and mechanical properties of the bentonite can be modified by the special conditions of the barrier. (Author)

Characteristics of potential repository wastes

International Nuclear Information System (INIS)

1992-07-01

The LWR spent fuels discussed in Volume 1 of this report comprise about 99% of all domestic non-reprocessed spent fuel. In this report we discuss other types of spent fuels which, although small in relative quantity, consist of a number of diverse types, sizes, and compositions. Many of these fuels are candidates for repository disposal. Some non-LWR spent fuels are currently reprocessed or are scheduled for reprocessing in DOE facilities at the Savannah River Site, Hanford Site, and the Idaho National Engineering Laboratory. It appears likely that the reprocessing of fuels that have been reprocessed in the past will continue and that the resulting high-level wastes will become part of defense HLW. However, it is not entirely clear in some cases whether a given fuel will be reprocessed, especially in cases where pretreatment may be needed before reprocessing, or where the enrichment is not high enough to make reprocessing attractive. Some fuels may be canistered, while others may require special means of disposal. The major categories covered in this chapter include HTGR spent fuel from the Fort St. Vrain and Peach Bottom-1 reactors, research and test reactor fuels, and miscellaneous fuels, and wastes generated from the decommissioning of facilities

In-situ test programs related to design and construction of high-level nuclear waste (HLW) deep geologic repositories. Appendices. Final report (Task 2), June 1981-November 1982

International Nuclear Information System (INIS)

Roberds, W.; Bauhof, F.; Gonano, L.

1983-03-01

The media and sites considered include: (1) basalt at Hanford, Washington; (2) tuff at Yucca Mountain, Nevada Test Site; (3) domal salt at specific Gulf Coast sites; (4) bedded salt at an unspecified site; and (5) granite at an unspecified site. A licensing perspective is outlined and a defensible rationale developed and utilized for the test selection process. This rationale essentially consists of: establishing the information needs for construction authorization; assessing the relevant capabilities of available tests; and matching the capabilities of specific tests to the perceived information needs. The information needs at any time consist of the additional information (if any) needed in order to predict satisfactory repository system performance with the required level of confidence, and thus are a function of: the significance of the repository engineered components and site characteristics to system performance; the currently available information, which may be supplemented with time; and the acceptable level of confidence in satisfactory performance for each licensing step. Determination of the acceptable levels of confidence and the significance of repository system components is outside the scope of this report. Suitable assumptions have thus been made regarding the development of information needs for construction authorization by the time of initial site characterization reports submittals. Tests which are available and respond to the perceived media/site specific information needs, either by simulation or assessment of site characteristics, are identified and their capabilities assessed. Specific in situ tests are investigated and described in detail. Research and development which might be effective in improving tests capabilities have been recommended

Development of an international safeguards approach to the final disposal of spent fuel in geological repositories

International Nuclear Information System (INIS)

Murphey, W.M.; Moran, B.W.; Fattah, A.

1996-01-01

The International Atomic Energy Agency (IAEA) is currently pursuing development of an international safeguards approach for the final disposal of spent fuel in geological repositories through consultants meetings and through the Program for Development of Safeguards for Final Disposal of Spent Fuel in Geological Repositories (SAGOR). The consultants meetings provide policy guidance to IAEA; SAGOR recommends effective approaches that can be efficiently implemented by IAEA. The SAGOR program, which is a collaboration of eight Member State Support Programs (MSSPs), was initiated in July 1994 and has identified 15 activities in each of three areas (i.e. conditioning facilities, active repositories, and closed repositories) that must be performed to ensure an efficient, yet effective safeguards approach. Two consultants meetings have been held: the first in May 1991 and the last in November 1995. For nuclear materials emplaced in a geological repository, the safeguards objectives were defined to be (1) to detect the diversion of spent fuel, whether concealed or unconcealed, from the repository and (2) to detect undeclared activities of safeguards concern (e.g., tunneling, underground reprocessing, or substitution in containers)

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)

Developing criteria to establish Trusted Digital Repositories

Science.gov (United States)

Faundeen, John L.

2017-01-01

This paper details the drivers, methods, and outcomes of the U.S. Geological Survey’s quest to establish criteria by which to judge its own digital preservation resources as Trusted Digital Repositories. Drivers included recent U.S. legislation focused on data and asset management conducted by federal agencies spending $100M USD or more annually on research activities. The methods entailed seeking existing evaluation criteria from national and international organizations such as International Standards Organization (ISO), U.S. Library of Congress, and Data Seal of Approval upon which to model USGS repository evaluations. Certification, complexity, cost, and usability of existing evaluation models were key considerations. The selected evaluation method was derived to allow the repository evaluation process to be transparent, understandable, and defensible; factors that are critical for judging competing, internal units. Implementing the chosen evaluation criteria involved establishing a cross-agency, multi-disciplinary team that interfaced across the organization. 

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

International Nuclear Information System (INIS)

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

2009-01-01

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

State technical review of the HLNW program and the peer review process

International Nuclear Information System (INIS)

Jacob, G.R.

1986-01-01

Millions of dollars are being spent on state governments' review of the Department of Energy (DOE) high level waste (HLW) repository program. A significant portion of the review efforts focus on technical issues surrounding the development and installment of HLW disposal technologies. Some view the states' technical review efforts as part of a peer review process. However, this interpretation reveals a misunderstanding of the concept of peer review and the purposes of state technical review

Quality assurance records system for research and development activities in support of geologic repository programs

International Nuclear Information System (INIS)

Smith, J.W.; Ryder, D.E.

1987-01-01

The Pacific Northwest Laboratory (PNL), which is operated by Battelle Memorial Institute for the Department of Energy, is conducting site-specific research for all three candidate sites for the first geologic high-level waste repository, as well as generic research for the second repository. In conjunction with this effort, PNL has developed a quality assurance (QA) program that is applicable to all organizations that are performing research and development (R and D) activities in support of the repository programs. This QA program meets the basic and supplemental requirements of ANSI/ASME NQA-1-1983 and the Nuclear Regulatory Commission (NRC) Review Plan for QA Programs for Site Characterization of High Level Nuclear Waste Repositories. A key part of this program is the handling of QA records that may ultimately support the licensing process for the repository. This paper describes a QA records system that is flexible enough to accommodate several types of research, such as paper studies, test method development, site characterization studies, software development, and hardware design. In addition, the QA records system is acceptable to a variety of sponsors who have licensing concerns. The QA procedures and their relation to the requirements are described. Most important is the discussion on the approaches used to assure that the records are organized such that the user can readily recreate or defend data, conclusions, and recommendations resulting from the research

The cathedral and the bazaar of e-repository development: encouraging community engagement with moving pictures and sound

Directory of Open Access Journals (Sweden)

Denis Wong

2008-12-01

Full Text Available This paper offers an insight into the development, use and governance of e-repositories for learning and teaching, illustrated by Eric Raymond's bazaar and cathedral analogies and by a comparison of collection strategies that focus on content coverage or on the needs of users. It addresses in particular the processes that encourage and achieve community engagement. This insight is illustrated by one particular e-repository, the Education Media On-Line (EMOL service. This paper draws analogies between the bazaar approach for open source software development and its possibilities for developing e-repositories for learning and teaching. It suggests in particular that the development, use and evaluation of online moving pictures and sound objects for learning and teaching can benefit greatly from the community engagement lessons provided by the development, use and evaluation of open source software. Such lessons can be underpinned by experience in the area of learning resource collections, where repositories have been classified as ‘collections-based' or ‘user-based'. Lessons from the open source movement may inform the development of e-repositories such as EMOL in the future.

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

Engineering Review Group (ERG) and Geologic Review Group (GRG) report on brine migration at the Deaf Smith County site salt repository horizon

International Nuclear Information System (INIS)

1986-01-01

In April 1986, ONWI requested the ERG and GRG to jointly address the status of current knowledge of, and ONWI approach to further characterization of, the geohydrology of the candidate repository horizon of the potential site in Deaf Smith County, Texas. Specifically, the ERG-GRG was asked to evaluate the status of understanding of the hydrogeology of the Lower San Andres Unit 4 (LSA-4) evaporite section and identify any major gaps in the data; evaluate the current understanding of the chemistry and movement of brines in the LSA-4 salt and associated interbeds; develop recommendations for estimating the upper limit quantity of brines, and modeling the brine movement, with respect to the emplaced HLW packages; and identify questions concerning the chemistry of the brines and recommend a technical approach to addressing these questions. 19 refs

Confidence in the long-term safety of deep geological repositories. Its development and communication

International Nuclear Information System (INIS)

1999-01-01

The technical aspects of confidence have been the subject of considerable debate, especially the concept of model validation. The safety case that is provided at a particular stage in the planning, construction, operation or closure of a deep geological repository is a part of a broader decision basis that guides the repository-development process. The basic steps for deriving the safety case at various stages of repository development involve: a safety assessment; and the documentation of the safety assessment, a statement of confidence in the safety indicated by the assessment, and the confirmation of the appropriateness of the safety strategy. The approaches to establish confidence in the evaluation of safety should aim to ensure that the decisions taken within the incremental process of repository development are well-founded. Various aspects of confidence in the evaluation of safety, and their integration within a safety case, are presented in detail in the present report. When communicating confidence in the findings of a safety assessment, clarity in the communication of concepts is always required. Consistent with this requirement, key concepts are specifically defined in the main text of the report. (R.P.)

Reference concepts for the final disposal of LWR spent fuel and other high activity wastes in Spain

International Nuclear Information System (INIS)

Huertas, F.; Ulibarri, A.

1993-01-01

Studies over the last three years have been recently concluded with the selection of a reference repository concept for the final disposal of spent fuel and other high activity wastes in deep geological formations. Two non-site specific preliminary designs, at a conceptual level, have been developed; one considers granite as the host rock and the other rock salt formations. The Spanish General Radioactive Waste Program also considers clay as a potential host rock for HLW deep disposal; conceptualization for a deep repository in clay is in the initial phase of development. The salt repository concept contemplates the disposal of the HLW in self-shielding casks emplaced in the drifts of an underground facility, excavated at a depth of 850 m in a bedded salt formation. The Custos Type I(7) cask admits up to seven intact PWR fuel assemblies or 21 of BWR type. The final repository facilities are planned to accept a total of 20,000 fuel assemblies (PWR and BWR) and 50 vitrified waste canisters over a period of 25 years. The total space needed for the surface facilities amounts to 322,000 m 2 , including the rock salt dump. The space required for the underground facilities amounts to 1.2 km 2 , approximately. The granite repository concept contemplates the disposal of the HLW in carbon steel canisters, embedded in a 0.75 m thick buffer of swelling smectite clay, in the drifts of an underground facility, excavated at a depth of 55 m in granite. Each canister can host 3 PWR or 9 BWR fuel assemblies. For this concept the total number of canisters needed amounts to 4,860. The space required for the surface and underground facilities is similar to that of the salt concept. The technical principles and criteria used for the design are discussed, and a description of the repository concept is presented

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

International Nuclear Information System (INIS)

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

2000-01-01

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

Development of site suitability criteria for the high level waste repository for Lawrence Livermore Laboratories

International Nuclear Information System (INIS)

1977-06-01

Results of our mining, geological and geotechnical studies provided in support of the development of site suitability criteria for the high level waste repository are presented. The primary purpose of the work was the identification and development of appropriate geotechnical descriptors and coefficients required for the Site Suitability Repository Model. This model was developed by The Analytic Sciences Corporation (TASC) of Reading, Massachusetts and is not described in this report

Development of site suitability criteria for the high level waste repository for Lawrence Livermore Laboratories

Energy Technology Data Exchange (ETDEWEB)

1977-06-01

Results of our mining, geological and geotechnical studies provided in support of the development of site suitability criteria for the high level waste repository are presented. The primary purpose of the work was the identification and development of appropriate geotechnical descriptors and coefficients required for the Site Suitability Repository Model. This model was developed by The Analytic Sciences Corporation (TASC) of Reading, Massachusetts and is not described in this report.

Office of Crystalline Repository Development FY 83 technical project plan

International Nuclear Information System (INIS)

1983-03-01

The technical plan for FY 83 activities of the Office of Crystalline Repository Development is presented in detail. Crystalline Rock Project objectives are discussed in relation to the National Waste Terminal storage (NWTS) program. The plan is in full compliance with requirements mandated by the Nuclear Waste Policy Act of 1982. Implementation will comply with the requirements and criteria set forth in the Nuclear Regulatory Commission regulations (10 CFR 60) and the Environmental Protection Agency standard (40 CFR 191). Technical approaches and the related milestones and schedules are presented for each of the Level 3 NWTS work Breakdown Structure Tasks. These are: Systems, Waste Package, Site, Repository, Regulatory and Institutional, Test Facilities and Excavations, Land Acquisition, and Program Management

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

International Nuclear Information System (INIS)

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

1984-04-01

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

Microbial effects on high-level waste disposal. Research review and perspective

Energy Technology Data Exchange (ETDEWEB)

Ohnuki, Toshihiko [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2002-09-01

Various microorganisms have been observed in deep geologic formation. The effects of such microorganisms on the performance of HLW disposal are still unknown. This paper reviews the studies of microbial effects on the long-term containment of HLW disposal, and discusses the future work to be carried out. Microbial reduction and oxidation and byproducts derived from microbial activities affect performance of HLW repository and have a potential to enhance actinides migration in geologic formation (degradation of the materials of repository, complex-formation, dissolution of actinides precipitates and occurrence of nm scale colloid formation). Potential microbial perturbation of performance of the barriers may enhance confinement of actinides by biomineralization, bioadsorption, bioaccumulation and precipitation. These studies indicate that further experiments are required to elucidate microbial effects on the performance of HLW disposal. (author)

Consortial routes to effective repositories

OpenAIRE

Moyle, M.; Proudfoot, R.

2009-01-01

A consortial approach to the establishment of repository services can help a group of Higher Education Institutions (HEIs) to share costs, share technology and share expertise. Consortial repository work can tap into existing structures, or it can involve new groupings of institutions with a common interest in exploring repository development. This Briefing Paper outlines some of the potential benefits of collaborative repository activity, and highlights some of the technical and organisation...

SUBSURFACE REPOSITORY INTEGRATED CONTROL SYSTEM DESIGN

International Nuclear Information System (INIS)

Randle, D.C.

2000-01-01

The primary purpose of this document is to develop a preliminary high-level functional and physical control system architecture for the potential repository at Yucca Mountain. This document outlines an overall control system concept that encompasses and integrates the many diverse process and communication systems being developed for the subsurface repository design. This document presents integrated design concepts for monitoring and controlling the diverse set of subsurface operations. The Subsurface Repository Integrated Control System design will be composed of a series of diverse process systems and communication networks. The subsurface repository design contains many systems related to instrumentation and control (I andC) for both repository development and waste emplacement operations. These systems include waste emplacement, waste retrieval, ventilation, radiological and air monitoring, rail transportation, construction development, utility systems (electrical, lighting, water, compressed air, etc.), fire protection, backfill emplacement, and performance confirmation. Each of these systems involves some level of I andC and will typically be integrated over a data communications network throughout the subsurface facility. The subsurface I andC systems will also interface with multiple surface-based systems such as site operations, rail transportation, security and safeguards, and electrical/piped utilities. In addition to the I andC systems, the subsurface repository design also contains systems related to voice and video communications. The components for each of these systems will be distributed and linked over voice and video communication networks throughout the subsurface facility. The scope and primary objectives of this design analysis are to: (1) Identify preliminary system-level functions and interfaces (Section 6.2). (2) Examine the overall system complexity and determine how and on what levels the engineered process systems will be monitored

Influence analysis of Github repositories.

Science.gov (United States)

Hu, Yan; Zhang, Jun; Bai, Xiaomei; Yu, Shuo; Yang, Zhuo

2016-01-01

With the support of cloud computing techniques, social coding platforms have changed the style of software development. Github is now the most popular social coding platform and project hosting service. Software developers of various levels keep entering Github, and use Github to save their public and private software projects. The large amounts of software developers and software repositories on Github are posing new challenges to the world of software engineering. This paper tries to tackle one of the important problems: analyzing the importance and influence of Github repositories. We proposed a HITS based influence analysis on graphs that represent the star relationship between Github users and repositories. A weighted version of HITS is applied to the overall star graph, and generates a different set of top influential repositories other than the results from standard version of HITS algorithm. We also conduct the influential analysis on per-month star graph, and study the monthly influence ranking of top repositories.

Demonstration of pyrometallurgical processing for metal fuel and HLW

International Nuclear Information System (INIS)

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

2001-01-01

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

Assessment of the safety reserve offered by a concrete buffer in case of a geological repository in clay

International Nuclear Information System (INIS)

Govaerts, Joan; Weetjens, Eef; Marivoet, Jan

2012-01-01

Performance assessment calculations have been performed to investigate if the sorption of 14 C, 36 Cl and 129 I on the cementitious materials occurring in the near field of the repository on the diffusion would offer an extra safety reserve to deep disposal of vitrified HLW. Four cases have been studied: a reference case with no cementitious material and three cases in which the considered concrete region was subsequently extended to the buffer, backfill and gallery liner. The results show a beneficial impact on peak dose and residence time of the three radionuclides. The effect on total released fractions is very high for 14 C, moderate for 36 Cl and small for 129 I

Development, evaluation, and selection of candidate high-level waste forms

International Nuclear Information System (INIS)

Bernadzikowski, T.A.; Allender, J.S.; Gordon, D.E.; Gould, T.H. Jr.

1982-01-01

The seven candidate waste forms, evaluated as potential media for the immobilization and gelogic disposal of high-level nuclear wastes were borosilicate glass, SYNROC, tailored ceramic, high-silica glass, FUETAP concrete, coated sol-gel particles, and glass marbles in a lead matrix. The evaluation, completed on August 1, 1981, combined preliminary waste form evaluations conducted at Department of Energy (DOE) defense waste-sites and at independent laboratories, peer review assessments, a product performance evaluation, and a processability analysis. Based on the combined results of these four inputs, two of the seven forms, borosilicate glass and a titanate-based ceramic, SYNROC, were selected as the reference and alternative forms, respectively, for continued development and evaluation in the National HLW Program. The borosilicate glass and ceramic forms were further compared during FY-1982 on the basis of risk assessments, cost comparisons, properties comparisons, and conformance with proposed regulatory and repository criteria. Both the glass and ceramic forms are viable candidates for use at DOE defense HLW sites; they are also candidates for immobilization of commercial reprocessing wastes. This paper describes the waste form screening process, discusses each of the four major inputs considered in the selection of the two forms in 1981, and presents a brief summary of the comparisons of the two forms during 1982 and the selection process to determine the final form for SRP defense HLW

TWRS HLW interim storage facility search and evaluation

Energy Technology Data Exchange (ETDEWEB)

Calmus, R.B., Westinghouse Hanford

1996-05-16

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

A design concept of underground facilities for the deep geologic disposal of spent fuel

International Nuclear Information System (INIS)

Lee, Jong Youl; Choi, Heui Joo; Choi, Jong Won; Hahn, Pil Soo

2005-01-01

Spent nuclear fuel from nuclear power plants can be disposed in the underground repository. In this paper, a concept of Korean Reference HLW disposal System (KRS-1) design is presented. Though no site for the underground repository has been specified in Korea, but a generic site with granitic rock is considered for reference spent fuel repository design. To implement the concept, design requirements such as spent fuel characteristics and capacity of the repository and design principles were established. Then, based on these requirements and principles, a concept of the disposal process, the facilities and the layout of the repository was developed

Pre-disposal storage, transport and handling of vitrified high level waste

International Nuclear Information System (INIS)

Kempe, T.F.; Martin, A.

1981-05-01

The objectives of the study were to review non site-specific engineering features of the storage, transport and handling of vitrified high level radioactive waste prior to its transfer into an underground repository, and to identify those features which require validation or development. Section headings are: introduction (historical and technical background); characteristics and arisings of vitrified high level waste; overpacks (additional containment barrier, corrosion resistant); interim storage of HLW; transport of HLW; handling; conclusions and recommendations. (U.K.)

NOx AND HETEROGENEITY EFFECTS IN HIGH LEVEL WASTE (HLW)

International Nuclear Information System (INIS)

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

2000-01-01

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

Process mining software repositories

NARCIS (Netherlands)

Poncin, W.; Serebrenik, A.; Brand, van den M.G.J.

2011-01-01

Software developers' activities are in general recorded in software repositories such as version control systems, bug trackers and mail archives. While abundant information is usually present in such repositories, successful information extraction is often challenged by the necessity to

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

High Pressure Soxhlet Type Leachability testing device and leaching test of simulated high-level waste glass at high temperature

International Nuclear Information System (INIS)

Senoo, Muneaki; Banba, Tsunetaka; Tashiro, Shingo; Shimooka, Kenji; Araki, Kunio

1979-11-01

A High Pressure Soxhlet Type Leachability Testing Device (HIPSOL) was developed to evaluate long-period stability of high-level waste (HLW) solids. For simulated HLW solids, temperature dependency of the leachability was investigated at higher temperatures from 100 0 C to 300 0 C at 80 atm. Leachabilities of cesium and sodium at 295 0 C were 20 and 7 times higher than at 100 0 C, respectively. In the repository, the temperatures around solidified products may be hundred 0 C. It is essential to test them at such elevated temperatures. HIPSOL is also usable for accelerated test to evaluate long-period leaching behavior of HLW products. (author)

Developing of Radioactive Wastes Management Safety at Baldone Repository Radons

International Nuclear Information System (INIS)

Abramenkovs, A.; Abramenkova, G.; Klavins, M.

2008-01-01

The near surface radioactive wastes repository Radons near the Baldone city was put in operation in 1962. The safety assessment of repository was performed during 2000-2001 under the PHARE project to evaluate the recommended upgrades of repository. The outline design for new vaults and interim storage for long lived radioactive wastes was elaborated during 2003-2004 years. The Environmental Impact Assessment (EIA) for upgrade of Baldone repository was performed during 2004-2005 years. Additional evaluations of radioactive wastes management safety were performed during 2006 year by the experts of ENRESA, Spain. It was shown, that the additional efforts were spent for improving of radioactive wastes cementation in concrete containers. The results of tritium and Cs 137 leaching studies are presented and discussed. It was shown, that additives can significantly reduce the migration of radionuclides in ground water. The leaching coefficients for tritium and Cs 137 were determined to supply with the necessary data the risk assessment calculations for operation of Baldone repository R adons

The Efficacy of Institutional Repositories: Reflections on the Development of a Personalised Collection on UPSpace

Science.gov (United States)

Olivier, Elsabe

2007-01-01

There is much speculation that the development of institutional repositories will impact on or even change the traditional scholarly communication process. The purpose of this conversation is to introduce the reader to the use of and response to institutional repositories which were initiated by the Open Access Initiative. The concept of…

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

Energy Technology Data Exchange (ETDEWEB)

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

2012-11-13

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

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

International Nuclear Information System (INIS)

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

2012-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2008-07-01

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

Ramifications of defining high-level waste

International Nuclear Information System (INIS)

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

1987-01-01

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

Development of Spherical Near Field Model for Geological Radioactive Waste Repository

International Nuclear Information System (INIS)

Kim, S. Y.; Lee, K. J.; Chang, S. H.; Lee, K. J.; Chang, S. H.

2012-01-01

Modeling for geological radioactive waste repository can be divided into 3 parts. They are near field modeling related to engineered barrier, far field modeling related to natural barrier and biosphere modeling. In order to make the general application for safety assessment of geological waste repository, spherical geometry near field model has been developed. This model can be used quite extensively when users calculate equivalent spherical geometry for specific engineered barrier like equivalent waste radius, equivalent barrier radius and etc. Only diffusion was considered for general purpose but advection part can be updated. Goldsim and Goldsim Radionuclide Transport (RT) module were chosen and used as developing tool for the flexible modeling. Developer can freely make their own model with developer friendly graphic interface by using Goldsim. Furthermore, model with user friendly graphic interface can be developed by using Goldsim Dashboard Authoring module. The model has been validated by comparing the result with that of another model, inserting similar inputs and conditions. The model has been proved to be reasonably operating from the comparison result by validation process. Cylindrical model can be developed as a further work based on the knowledge and experience from this research

Development of Spherical Near Field Model for Geological Radioactive Waste Repository

Energy Technology Data Exchange (ETDEWEB)

Kim, S. Y.; Lee, K. J.; Chang, S. H. [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Lee, K. J.; Chang, S. H. [Khalifa Univ. of Science/Technology and Research, Abu Dhabi (United Arab Emirates)

2012-03-15

Modeling for geological radioactive waste repository can be divided into 3 parts. They are near field modeling related to engineered barrier, far field modeling related to natural barrier and biosphere modeling. In order to make the general application for safety assessment of geological waste repository, spherical geometry near field model has been developed. This model can be used quite extensively when users calculate equivalent spherical geometry for specific engineered barrier like equivalent waste radius, equivalent barrier radius and etc. Only diffusion was considered for general purpose but advection part can be updated. Goldsim and Goldsim Radionuclide Transport (RT) module were chosen and used as developing tool for the flexible modeling. Developer can freely make their own model with developer friendly graphic interface by using Goldsim. Furthermore, model with user friendly graphic interface can be developed by using Goldsim Dashboard Authoring module. The model has been validated by comparing the result with that of another model, inserting similar inputs and conditions. The model has been proved to be reasonably operating from the comparison result by validation process. Cylindrical model can be developed as a further work based on the knowledge and experience from this research.

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

Energy Technology Data Exchange (ETDEWEB)

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

2013-11-13

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

System studies in PA: Development of process influence diagram (PID) for SFR-1 repository near-field + far-field

International Nuclear Information System (INIS)

Stenhouse, M.J.; Miller, W.M.; Chapman, N.A.

2001-05-01

Scenario development is a key component of the performance assessment (PA) process for radioactive waste disposal, the primary objective being to ensure that all relevant factors associated with the future evolution of the repository system are properly considered in PA. As part of scenario development, a list of features, events and processes (FEPs) are identified and assembled, representing the Process System, with interactions/influences between FEPs incorporated in a Process Influence Diagram (PID). This report documents the technical work conducted between 1997 and the end of 1999 under the Systems Studies Project. The overall objective of this project has been the construction of a PID for the SFR-1 repository (final repository for reactor waste), this PID being the first stage in the identification of scenarios to describe future evolution of this repository. The PIDs discussed in this report have been created using two software applications: existing commercial software (Business Modeller, Infotool AB. Stockholm, Sweden) and, more recently, a newly developed software tool SPARTA (Enviros QuantiSci, Henley, U.K.). Although the focus of this report is on the application of SPARTA to PID development, it is important to document the work carried out prior to SPARTA being available, in order to provide a complete record of the entire SFR-1 PID development effort as well as preserving the context of the multi-year project. Following a description of the different disposal sections of the SFR-1 and the various near-field barriers, the sequential development (i.e. near-field of Silo, BMA, BLA, BTF sections; far-field; integrated near-field + far-field) of the PID for SFR-1 repository system using Business Modeller is described. Owing to the complexity of the repository, in terms of number of both different disposal sections (Silo, BLA, BMA, BTF) and barriers associated with each section, the two-dimensional (2D) PID created for SFR-1 using Business Modeller is

Centralized mouse repositories.

Science.gov (United States)

Donahue, Leah Rae; Hrabe de Angelis, Martin; Hagn, Michael; Franklin, Craig; Lloyd, K C Kent; Magnuson, Terry; McKerlie, Colin; Nakagata, Naomi; Obata, Yuichi; Read, Stuart; Wurst, Wolfgang; Hörlein, Andreas; Davisson, Muriel T

2012-10-01

Because the mouse is used so widely for biomedical research and the number of mouse models being generated is increasing rapidly, centralized repositories are essential if the valuable mouse strains and models that have been developed are to be securely preserved and fully exploited. Ensuring the ongoing availability of these mouse strains preserves the investment made in creating and characterizing them and creates a global resource of enormous value. The establishment of centralized mouse repositories around the world for distributing and archiving these resources has provided critical access to and preservation of these strains. This article describes the common and specialized activities provided by major mouse repositories around the world.

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

International Nuclear Information System (INIS)

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

2012-01-01

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

IAEA safeguards for geological repositories

International Nuclear Information System (INIS)

Moran, B.W.

2005-01-01

In September. 1988, the IAEA held its first formal meeting on the safeguards requirements for the final disposal of spent fuel and nuclear material-bearing waste. The consensus recommendation of the 43 participants from 18 countries at this Advisory Group Meeting was that safeguards should not terminate of spent fuel even after emplacement in, and closure of, a geologic repository.' As a result of this recommendation, the IAEA initiated a series of consultants' meetings and the SAGOR Programme (Programme for the Development of Safeguards for the Final Disposal of Spent Fuel in Geologic Repositories) to develop an approach that would permit IAEA safeguards to verify the non-diversion of spent fuel from a geologic repository. At the end of this process, in December 1997, a second Advisory Group Meeting, endorsed the generic safeguards approach developed by the SAGOR Programme. Using the SAGOR Programme results and consultants' meeting recommendations, the IAEA Department of Safeguards issued a safeguards policy paper stating the requirements for IAEA safeguards at geologic repositories. Following approval of the safeguards policy and the generic safeguards approach, the Geologic Repository Safeguards Experts Group was established to make recommendations on implementing the safeguards approach. This experts' group is currently making recommendations to the IAEA regarding the safeguards activities to be conducted with respect to Finland's repository programme. (author)

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

International Nuclear Information System (INIS)

Jiang Yaozhong; Tang Baolong; Zhang Baoshan; Zhou Hui

1995-01-01

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

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

International Nuclear Information System (INIS)

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

1997-01-01

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

Tc Chemistry in HLW: Role of Organic Complexants

International Nuclear Information System (INIS)

Hess, Nancy S.; Conradsen, Steven D.

2003-01-01

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

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

International Nuclear Information System (INIS)

Nagano, Koji

2003-01-01

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

Defense High Level Waste Disposal Container System Description Document

International Nuclear Information System (INIS)

Pettit, N. E.

2001-01-01

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

Korean Reference HLW Disposal System

Energy Technology Data Exchange (ETDEWEB)

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

2008-03-15

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

Evaluasi Website Repositori Institusi Universitas Surabaya

Directory of Open Access Journals (Sweden)

Amirul Ulum

2015-11-01

Full Text Available The  development  of  institutional  repositories  (IRs  in Indonesia within the broader framework of open access has significant impact on preserving intellectual capital and scholarly communication. Institutional  repositories  play  a  fundamental  role  in  centralizing, preserving,  and  making  accessible  institution’s  intellectual  capital. Evaluation of the system is to determine the functionality the system to meet the users need. Using a descriptive analysis this study wants to evaluate institutional repositories of University of Surabaya. The result is usefull for institution to develop the repository systems.

The RAGE Advanced Game Technologies Repository for Supporting Applied Game Development

NARCIS (Netherlands)

Georgiev, Atanas; Grigorov, Alexander; Bontchev, Boyan; Boytchev, Pavel; Stefanov, Krassen; Westera, Wim; Prada, Rui; Hollins, Paul; Moreno Ger, Pablo

2016-01-01

This paper describes the structural architecture of the RAGE repository, which is a unique and dedicated infrastructure that provides access to a wide variety of advanced technologies (RAGE software assets) for applied game development. These software assets are reusable across a wide diversity of

Preliminary waste form characteristics report Version 1.0. Revision 1

International Nuclear Information System (INIS)

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

1991-01-01

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

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

Safety analysis in subsurface repositories

International Nuclear Information System (INIS)

1985-06-01

The development of mathematical models to represent the repository-geosphere-biosphere system, and the development of a structure for data acquisition, processing, and use to analyse the safety of subsurface repositories, are presented. To study the behavior of radionuclides in geosphere a laboratory to determine the hydrodynamic dispersion coefficient was constructed. (M.C.K.) [pt

Thermodynamic modelling of bentonite-groundwater interaction and implications for near field chemistry in a repository for spent fuel

International Nuclear Information System (INIS)

Wanner, H.; Wersin, P.; Sierro, N.

1992-11-01

Predictions of near field geochemistry are made using a thermodynamic model for bentonite/ground interaction. This model is a refinement and extension of the model developed by the senior author. It is based on recent experiments performed at high solid/water ratio and adapted to the Swedish type of HLW repository design. Thus, from the obtained experimental results on solution composition, the model includes chemical reactions resulting from both the impurities and the main clay fraction within the bentonite. Ion exchange reactions are treated both with and without the contribution of edge sites. Due to its thermodynamic basis, the model exhibits prediction capability over a wide range of conditions in terms of solid/water ratio. The modelling of repository conditions implies, due to the lack of experimental information, simplifications with regard to thermodynamic properties of the bentonite. This mainly involves the non-consideration of the temperature effects and of the acid/base properties of the solid. Nevertheless, our results yield insight into important processes affecting porewater chemistry. Thus, the model suggests that proton exchange reactions may exert a strong control on calcite dissolution within highly compacted bentonite. Estimations of chemical changes over time in the bentonite were done in the basis of a mixing tank model. These results indicate transformation of Na-bentonite to Ca-bentonite over time. The extent of this process, however, critically depends on the amount of carbonate present in the bentonite. (authors) (34 refs.)

MAJOR REPOSITORY DESIGN ISSUES

International Nuclear Information System (INIS)

JACK N. BAILEY, DWAYNE CHESTNUT, JAMES COMPTON AND RICHARD D. SNELL

1997-01-01

The Yucca Mountain Project is focused on producing a four-part viability assessment in late FY98. Its four components (design, performance assessment, cost estimate, and licensing development plan) must be consistent. As a tool to compare design and performance assessment options, a series of repository pictures were developed for the sequential time phases of a repository. The boundaries of the time phases correspond to evolution in the engineered barrier system (EBS)

National high-level waste systems analysis plan

International Nuclear Information System (INIS)

Kristofferson, K.; Oholleran, T.P.; Powell, R.H.; Thiel, E.C.

1995-05-01

This document details the development of modeling capabilities that can provide a system-wide view of all US Department of Energy (DOE) high-level waste (HLW) treatment and storage systems. This model can assess the impact of budget constraints on storage and treatment system schedules and throughput. These impacts can then be assessed against existing and pending milestones to determine the impact to the overall HLW system. A nation-wide view of waste treatment availability will help project the time required to prepare HLW for disposal. The impacts of the availability of various treatment systems and throughput can be compared to repository readiness to determine the prudent application of resources or the need to renegotiate milestones

Influence of Iltization on the Ion-sorbing Capacity of Smectitic Clay

International Nuclear Information System (INIS)

Lee, Jae Owan; Cho, Won Jin

2008-01-01

A high-level waste(HLW) repository uses smectitic clay as a buffer material to inhibit the penetration of groundwater and to retard the release of radionuclides from the radioactive wastes to the surrounding environment. However, when the smectitic clay is exposed to an elevated temperature due to radioactive decay heat and geochemical conditions for a long time, its physicochemical and mineralogical properties may be degradated and thus lose its barrier functions. It has been known in literature that the degradation of these properties of the smectitic clay occurs by a illitization in which the smectite transforms into illite. Therefore, an understanding of the illitization is essential to evaluate the long-term barrier performance of smectitic clay for the buffer of a HLW repository. This paper will carry out hydrothermal reaction tests with domestic smectitic clay which will be favorably considered for the buffer material of a Korean HLW repository, and also investigate the influence of illization on the ion-sorbing capacity of the smectitic clay

Developing an Automatic Crawling System for Populating a Digital Repository of Professional Development Resources: A Pilot Study

Science.gov (United States)

Park, Jung-ran; Yang, Chris; Tosaka, Yuji; Ping, Qing; Mimouni, Houda El

2016-01-01

This study is a part of the larger project that develops a sustainable digital repository of professional development resources on emerging data standards and technologies for data organization and management in libraries. Toward that end, the project team developed an automated workflow to crawl for, monitor, and classify relevant web objects…

Corrosion of high purity copper as engineering barrier in deep geological repositories

International Nuclear Information System (INIS)

Ochoa, Maité; Rodríguez Martín, A.; Farina Silvia, B.

2013-01-01

Pure copper with oxygen content below 5 ppm (to minimize segregation at grain boundaries) and doped with phosphorus (to increase creep resistance) is the chosen material for the corrosion-resistant barrier of the High Level Radioactive 2 Wastecontainers in the Swedish and Finnish repository models. These models include the construction of the repository below the water table, which is a reducing environment in which copper has excellent resistance to general and localized corrosion in aqueous electrolytes. The aim of this work is contribute to determine the durability of the material, given that deep geological repositories of HLW are designed to ensure the protection of the environment for periods of hundreds of thousands years. As a first step in a more general analysis the effects of chloride, one of the main aggressive species of corrosion, are evaluated. To this purpose corrosion potential was determined and anodic polarization curves were performed in deaerated solutions varying the chloride concentration between 0.01 and 1M and the temperature between 30 and 90°C. Several electrochemical techniques were used: the evolution of corrosion potential was measured, anodic polarization curves were obtained and electrochemical impedance tests were performed. The analysis was complemented with microscopic observations of the type of corrosive attack, as well as determinations of the eventual corrosion products formed using Energy-Dispersive X-ray Analysis (EDS). Results show that the corrosion potential decreases with the increase of temperature and with the increase of chloride concentration. A correlation of the corrosion potential as a function of temperature and chloride concentration was obtained, with the purpose of making predictions in variable conditions.The current density increases both with temperature and with chloride concentration. A pitting potential is observed in certain conditions. (author)

Characteristics of potential repository wastes. Volume 2

Energy Technology Data Exchange (ETDEWEB)

1992-07-01

The LWR spent fuels discussed in Volume 1 of this report comprise about 99% of all domestic non-reprocessed spent fuel. In this report we discuss other types of spent fuels which, although small in relative quantity, consist of a number of diverse types, sizes, and compositions. Many of these fuels are candidates for repository disposal. Some non-LWR spent fuels are currently reprocessed or are scheduled for reprocessing in DOE facilities at the Savannah River Site, Hanford Site, and the Idaho National Engineering Laboratory. It appears likely that the reprocessing of fuels that have been reprocessed in the past will continue and that the resulting high-level wastes will become part of defense HLW. However, it is not entirely clear in some cases whether a given fuel will be reprocessed, especially in cases where pretreatment may be needed before reprocessing, or where the enrichment is not high enough to make reprocessing attractive. Some fuels may be canistered, while others may require special means of disposal. The major categories covered in this chapter include HTGR spent fuel from the Fort St. Vrain and Peach Bottom-1 reactors, research and test reactor fuels, and miscellaneous fuels, and wastes generated from the decommissioning of facilities.