Drift-Scale Radionuclide Transport

International Nuclear Information System (INIS)

Houseworth, J.

2004-01-01

The purpose of this model report is to document the drift scale radionuclide transport model, taking into account the effects of emplacement drifts on flow and transport in the vicinity of the drift, which are not captured in the mountain-scale unsaturated zone (UZ) flow and transport models ''UZ Flow Models and Submodels'' (BSC 2004 [DIRS 169861]), ''Radionuclide Transport Models Under Ambient Conditions'' (BSC 2004 [DIRS 164500]), and ''Particle Tracking Model and Abstraction of Transport Process'' (BSC 2004 [DIRS 170041]). The drift scale radionuclide transport model is intended to be used as an alternative model for comparison with the engineered barrier system (EBS) radionuclide transport model ''EBS Radionuclide Transport Abstraction'' (BSC 2004 [DIRS 169868]). For that purpose, two alternative models have been developed for drift-scale radionuclide transport. One of the alternative models is a dual continuum flow and transport model called the drift shadow model. The effects of variations in the flow field and fracture-matrix interaction in the vicinity of a waste emplacement drift are investigated through sensitivity studies using the drift shadow model (Houseworth et al. 2003 [DIRS 164394]). In this model, the flow is significantly perturbed (reduced) beneath the waste emplacement drifts. However, comparisons of transport in this perturbed flow field with transport in an unperturbed flow field show similar results if the transport is initiated in the rock matrix. This has led to a second alternative model, called the fracture-matrix partitioning model, that focuses on the partitioning of radionuclide transport between the fractures and matrix upon exiting the waste emplacement drift. The fracture-matrix partitioning model computes the partitioning, between fractures and matrix, of diffusive radionuclide transport from the invert (for drifts without seepage) into the rock water. The invert is the structure constructed in a drift to provide the floor of the

Abstraction of Drift Seepage

International Nuclear Information System (INIS)

J.T. Birkholzer

2004-01-01

This model report documents the abstraction of drift seepage, conducted to provide seepage-relevant parameters and their probability distributions for use in Total System Performance Assessment for License Application (TSPA-LA). Drift seepage refers to the flow of liquid water into waste emplacement drifts. Water that seeps into drifts may contact waste packages and potentially mobilize radionuclides, and may result in advective transport of radionuclides through breached waste packages [''Risk Information to Support Prioritization of Performance Assessment Models'' (BSC 2003 [DIRS 168796], Section 3.3.2)]. The unsaturated rock layers overlying and hosting the repository form a natural barrier that reduces the amount of water entering emplacement drifts by natural subsurface processes. For example, drift seepage is limited by the capillary barrier forming at the drift crown, which decreases or even eliminates water flow from the unsaturated fractured rock into the drift. During the first few hundred years after waste emplacement, when above-boiling rock temperatures will develop as a result of heat generated by the decay of the radioactive waste, vaporization of percolation water is an additional factor limiting seepage. Estimating the effectiveness of these natural barrier capabilities and predicting the amount of seepage into drifts is an important aspect of assessing the performance of the repository. The TSPA-LA therefore includes a seepage component that calculates the amount of seepage into drifts [''Total System Performance Assessment (TSPA) Model/Analysis for the License Application'' (BSC 2004 [DIRS 168504], Section 6.3.3.1)]. The TSPA-LA calculation is performed with a probabilistic approach that accounts for the spatial and temporal variability and inherent uncertainty of seepage-relevant properties and processes. Results are used for subsequent TSPA-LA components that may handle, for example, waste package corrosion or radionuclide transport

Waste package emplacement borehole option study

International Nuclear Information System (INIS)

Streeter, W.S.

1992-03-01

This study evaluates the cost and thermal effects of various waste package emplacement configurations that differ in emplacement orientation, number of containers per borehole, and standoff distance at the potential Yucca Mountain nuclear waste repository. In this study, eight additional alternatives to the vertical and horizontal orientation options presented in the Site Characterization Plan Conceptual Design Report are considered. Typical panel layout configurations based on thermal analysis of the waste and cost estimates for design and construction, operations, and closure and decommissioning were made for each emplacement option. For the thermal analysis average waste 10 years out of reactor and the SIM code were used to determine whether the various configurations temperatures would exceed the design criteria for temperature. This study does not make a recommendation for emplacement configuration, but does provide information for comparison of alternatives

DRIFT-SCALE COUPLED PROCESSES (DST AND TH SEEPAGE) MODELS

International Nuclear Information System (INIS)

J.T. Birkholzer; S. Mukhopadhyay

2005-01-01

The purpose of this report is to document drift-scale modeling work performed to evaluate the thermal-hydrological (TH) behavior in Yucca Mountain fractured rock close to waste emplacement drifts. The heat generated by the decay of radioactive waste results in rock temperatures elevated from ambient for thousands of years after emplacement. Depending on the thermal load, these temperatures are high enough to cause boiling conditions in the rock, giving rise to water redistribution and altered flow paths. The predictive simulations described in this report are intended to investigate fluid flow in the vicinity of an emplacement drift for a range of thermal loads. Understanding the TH coupled processes is important for the performance of the repository because the thermally driven water saturation changes affect the potential seepage of water into waste emplacement drifts. Seepage of water is important because if enough water gets into the emplacement drifts and comes into contact with any exposed radionuclides, it may then be possible for the radionuclides to be transported out of the drifts and to the groundwater below the drifts. For above-boiling rock temperatures, vaporization of percolating water in the fractured rock overlying the repository can provide an important barrier capability that greatly reduces (and possibly eliminates) the potential of water seeping into the emplacement drifts. In addition to this thermal process, water is inhibited from entering the drift opening by capillary forces, which occur under both ambient and thermal conditions (capillary barrier). The combined barrier capability of vaporization processes and capillary forces in the near-field rock during the thermal period of the repository is analyzed and discussed in this report

Tabulated In-Drift Geometric and Thermal Properties Used In Drift-Scale Models for TSPA-SR

International Nuclear Information System (INIS)

N.D. Francis

2000-01-01

The objective of this calculation is to provide in-drift physical properties required by the drift-scale models (both two- and three-dimensional) used in total system performance assessments (TSPA). The physical properties include waste package geometry, waste package thermal properties, emplacement drift geometry including backfill and invert geometry and properties (both thermal and hydrologic), drip shield geometry and thermal properties, all tabulated in a single source

Concept of Operations for Waste Transport, Emplacement, and Retrieval

International Nuclear Information System (INIS)

Raczka, Norman T.

2001-01-01

The preparation of this technical report has two objectives. The first objective is to discuss the base case concepts of waste transport, emplacement, and retrieval operations and evaluate these operations relative to a lower-temperature repository design. Aspects of the operations involved in waste transport, emplacement and retrieval may be affected by the lower-temperature operating schemes. This report evaluates the effects the lower-temperature alternatives may have on the operational concepts involved in emplacing and retrieving waste. The second objective is to provide backup material for the design description, in a traceable and defensible format, for Section 2 of the Waste Emplacement/Retrieval System Description Document

Thermal simulation of drift emplacement. Geotechnical and geophysical investigations in and around backfilled galleries

International Nuclear Information System (INIS)

Schneefub, J.U.; Gommlich, G.E.

1988-01-01

The concept for the direct disposal of spent fuel in rock salt foresees the emplacement of large waste canisters on the floor of a disposal gallery. Subsequent to emplacement the drift is backfilled with salt grit. In a demonstration test six cylindrical containments are to be emplaced within distances of 3 m from each other in two parallel galleries of 14 m 2 , separated by a pillar of 10 m thickness. They will be heated up by a power output of approx. 1 kW/m to 200 degree C surface temperature by electrical heaters. The thermal and mechanical response of the salt rock and the backfilling to the artificial heating is to be investigated as follows: (1) measurement of the temperature field at the contained surface, in the backfill and in the rock salt; (2) deformation measurements of the salt rock around the heated drifts; (3) measurements of tunnel convergence in heated and unheated sections; (4) compaction measurements of the backfilling in heated and unheated areas; (5) measurement of rock stresses in areas very close to the galleries; and (6) pressure measurements in the backfilling, between backfilling and rock, and between backfilling and containers. The rock burst monitoring system of the Asse salt mine will be expanded to this special field of the mine to detect seismic events due to thermomechanical effects. Another seismoacoustic system will be installed to observe the compaction of the backfilling. This system must be calibrated for the relation between the density and velocities of seismic waves. It is planned to monitor the density by gamma-gamma log measurements. The changes in overall density during the entire experiment will be observed by gravimeter measurements of high precision

Drift Scale THM Model

International Nuclear Information System (INIS)

Rutqvist, J.

2004-01-01

This model report documents the drift scale coupled thermal-hydrological-mechanical (THM) processes model development and presents simulations of the THM behavior in fractured rock close to emplacement drifts. The modeling and analyses are used to evaluate the impact of THM processes on permeability and flow in the near-field of the emplacement drifts. The results from this report are used to assess the importance of THM processes on seepage and support in the model reports ''Seepage Model for PA Including Drift Collapse'' and ''Abstraction of Drift Seepage'', and to support arguments for exclusion of features, events, and processes (FEPs) in the analysis reports ''Features, Events, and Processes in Unsaturated Zone Flow and Transport and Features, Events, and Processes: Disruptive Events''. The total system performance assessment (TSPA) calculations do not use any output from this report. Specifically, the coupled THM process model is applied to simulate the impact of THM processes on hydrologic properties (permeability and capillary strength) and flow in the near-field rock around a heat-releasing emplacement drift. The heat generated by the decay of radioactive waste results in elevated rock temperatures for thousands of years after waste emplacement. Depending on the thermal load, these temperatures are high enough to cause boiling conditions in the rock, resulting in water redistribution and altered flow paths. These temperatures will also cause thermal expansion of the rock, with the potential of opening or closing fractures and thus changing fracture permeability in the near-field. Understanding the THM coupled processes is important for the performance of the repository because the thermally induced permeability changes potentially effect the magnitude and spatial distribution of percolation flux in the vicinity of the drift, and hence the seepage of water into the drift. This is important because a sufficient amount of water must be available within a

Drift-scale thermomechanical analysis for the retrievability systems study

International Nuclear Information System (INIS)

Tsai, F.C.

1996-01-01

A numerical method was used to estimate the stability of potential emplacement drifts without considering a ground support system as a part of the Thermal Loading Systems Study for the Yucca Mountain Site Characterization Project. The stability of the drift is evaluated with two variables: the level of thermal loading and the diameter of the emplacement drift. The analyses include the thermomechanical effects generated by the excavation of the drift, subsequently by the thermal loads from heat-emitting waste packages, and finally by the thermal reduction resulting from rapid cooling ventilation required for the waste retrieval if required. The Discontinuous Deformation Analysis (DDA) code was used to analyze the thermomechanical response of the rock mass of multiple blocks separated by joints. The result of this stability analysis is used to discuss the geomechanical considerations for the advanced conceptual design (ACD) with respect to retrievability. In particular, based on the rock mass strength of the host rock described in the current version of the Reference Information Base, the computed thermal stresses, generated by 111 MTU/acre thermal loads in the near field at 100 years after waste emplacement, is beyond the criterion for the rock mass strength used to predict the stability of the rock mass surrounding the emplacement drift

Igneous Intrusion Impacts on Waste Packages and Waste Forms

International Nuclear Information System (INIS)

P. Bernot

2004-01-01

The purpose of this model report is to assess the potential impacts of igneous intrusion on waste packages and waste forms in the emplacement drifts at the Yucca Mountain Repository. The model is based on conceptual models and includes an assessment of deleterious dynamic, thermal, hydrologic, and chemical impacts. This constitutes the waste package and waste form impacts submodel of the Total System Performance Assessment for the License Application (TSPA-LA) model assessing the impacts of a hypothetical igneous intrusion event on the repository total system performance. This submodel is carried out in accordance with Technical Work Plan for Waste Form Degradation Modeling, Testing, and Analyses in Support of SR and LA (BSC 2003a) and Total System Performance Assessment-License Application Methods and Approaches (BSC 2002a). The technical work plan is governed by the procedures of AP-SIII.10Q, Models. Any deviations from the technical work plan are documented in the TSPA-LA approach to implementing the models for waste package and waste form response during igneous intrusion is based on identification of damage zones. Zone 1 includes all emplacement drifts intruded by the basalt dike, and Zone 2 includes all other emplacement drifts in the repository that are not in Zone 1. This model report will document the following model: (1) Impacts of magma intrusion on the components of engineered barrier system (e.g., drip shields and cladding) of emplacement drifts in Zone 1, and the fate of waste forms. (2) Impacts of conducting magma heat and diffusing magma gases on the drip shields, waste packages, and cladding in the Zone 2 emplacement drifts adjacent to the intruded drifts. (3) Impacts of intrusion on Zone 1 in-drift thermal and geochemical environments, including seepage hydrochemistry. The scope of this model only includes impacts to the components stated above, and does not include impacts to other engineered barrier system (EBS) components such as the invert and

Igneous Intrusion Impacts on Waste Packages and Waste Forms

Energy Technology Data Exchange (ETDEWEB)

P. Bernot

2004-08-16

The purpose of this model report is to assess the potential impacts of igneous intrusion on waste packages and waste forms in the emplacement drifts at the Yucca Mountain Repository. The model is based on conceptual models and includes an assessment of deleterious dynamic, thermal, hydrologic, and chemical impacts. This constitutes the waste package and waste form impacts submodel of the Total System Performance Assessment for the License Application (TSPA-LA) model assessing the impacts of a hypothetical igneous intrusion event on the repository total system performance. This submodel is carried out in accordance with Technical Work Plan for Waste Form Degradation Modeling, Testing, and Analyses in Support of SR and LA (BSC 2003a) and Total System Performance Assessment-License Application Methods and Approaches (BSC 2002a). The technical work plan is governed by the procedures of AP-SIII.10Q, Models. Any deviations from the technical work plan are documented in the TSPA-LA approach to implementing the models for waste package and waste form response during igneous intrusion is based on identification of damage zones. Zone 1 includes all emplacement drifts intruded by the basalt dike, and Zone 2 includes all other emplacement drifts in the repository that are not in Zone 1. This model report will document the following model: (1) Impacts of magma intrusion on the components of engineered barrier system (e.g., drip shields and cladding) of emplacement drifts in Zone 1, and the fate of waste forms. (2) Impacts of conducting magma heat and diffusing magma gases on the drip shields, waste packages, and cladding in the Zone 2 emplacement drifts adjacent to the intruded drifts. (3) Impacts of intrusion on Zone 1 in-drift thermal and geochemical environments, including seepage hydrochemistry. The scope of this model only includes impacts to the components stated above, and does not include impacts to other engineered barrier system (EBS) components such as the invert and

SUBSURFACE EMPLACEMENT TRANSPORTATION SYSTEM

International Nuclear Information System (INIS)

Wilson, T.; Novotny, R.

1999-01-01

The objective of this analysis is to identify issues and criteria that apply to the design of the Subsurface Emplacement Transportation System (SET). The SET consists of the track used by the waste package handling equipment, the conductors and related equipment used to supply electrical power to that equipment, and the instrumentation and controls used to monitor and operate those track and power supply systems. Major considerations of this analysis include: (1) Operational life of the SET; (2) Geometric constraints on the track layout; (3) Operating loads on the track; (4) Environmentally induced loads on the track; (5) Power supply (electrification) requirements; and (6) Instrumentation and control requirements. This analysis will provide the basis for development of the system description document (SDD) for the SET. This analysis also defines the interfaces that need to be considered in the design of the SET. These interfaces include, but are not limited to, the following: (1) Waste handling building; (2) Monitored Geologic Repository (MGR) surface site layout; (3) Waste Emplacement System (WES); (4) Waste Retrieval System (WRS); (5) Ground Control System (GCS); (6) Ex-Container System (XCS); (7) Subsurface Electrical Distribution System (SED); (8) MGR Operations Monitoring and Control System (OMC); (9) Subsurface Facility System (SFS); (10) Subsurface Fire Protection System (SFR); (11) Performance Confirmation Emplacement Drift Monitoring System (PCM); and (12) Backfill Emplacement System (BES)

Waste Handling and Emplacement Options for Disposal of Radioactive Waste in Deep Boreholes.

Energy Technology Data Exchange (ETDEWEB)

Cochran, John R.; Hardin, Ernest

2015-11-01

Traditional methods cannot be used to handle and emplace radioactive wastes in boreholes up to 16,400 feet (5 km) deep for disposal. This paper describes three systems that can be used for handling and emplacing waste packages in deep borehole: (1) a 2011 reference design that is based on a previous study by Woodward–Clyde in 1983 in which waste packages are assembled into “strings” and lowered using drill pipe; (2) an updated version of the 2011 reference design; and (3) a new concept in which individual waste packages would be lowered to depth using a wireline. Emplacement on coiled tubing was also considered, but not developed in detail. The systems described here are currently designed for U.S. Department of Energy-owned high-level waste (HLW) including the Cesium- 137/Strontium-90 capsules from the Hanford Facility and bulk granular HLW from fuel processing in Idaho.

Automated emplacement and retrieval of hazardous waste

International Nuclear Information System (INIS)

Slocum, A.H.; Hou, W.M.

1987-01-01

The design of several dedicated machines to perform simple tasks often results in higher system reliability and efficiency than the design of a single, multifunctional machine. Similarly, a reliable system for emplacement and retrieval of nuclear waste can be realized if emplacement/retrieval operations are decomposed into a well-defined series of independent tasks. The basic methodology is to design a system that eliminates contact between the waste package and the vehicle in the event of machine failure. The disabled vehicle can then be withdrawn to a safe location, repaired, and set back to resume normal operation

Waste Handling and Emplacement Options for Disposal of Radioactive Waste in Deep Boreholes

International Nuclear Information System (INIS)

Cochran, John R.; Hardin, Ernest

2015-01-01

Traditional methods cannot be used to handle and emplace radioactive wastes in boreholes up to 16,400 feet (5 km) deep for disposal. This paper describes three systems that can be used for handling and emplacing waste packages in deep borehole: (1) a 2011 reference design that is based on a previous study by Woodward-Clyde in 1983 in which waste packages are assembled into ''strings'' and lowered using drill pipe; (2) an updated version of the 2011 reference design; and (3) a new concept in which individual waste packages would be lowered to depth using a wireline. Emplacement on coiled tubing was also considered, but not developed in detail. The systems described here are currently designed for U.S. Department of Energy-owned high-level waste (HLW) including the Cesium- 137/Strontium-90 capsules from the Hanford Facility and bulk granular HLW from fuel processing in Idaho.

Seepage Model for PA Including Drift Collapse

International Nuclear Information System (INIS)

Li, G.; Tsang, C.

2000-01-01

The purpose of this Analysis/Model Report (AMR) is to document the predictions and analysis performed using the Seepage Model for Performance Assessment (PA) and the Disturbed Drift Seepage Submodel for both the Topopah Spring middle nonlithophysal and lower lithophysal lithostratigraphic units at Yucca Mountain. These results will be used by PA to develop the probability distribution of water seepage into waste-emplacement drifts at Yucca Mountain, Nevada, as part of the evaluation of the long term performance of the potential repository. This AMR is in accordance with the ''Technical Work Plan for Unsaturated Zone (UZ) Flow and Transport Process Model Report'' (CRWMS M andO 2000 [153447]). This purpose is accomplished by performing numerical simulations with stochastic representations of hydrological properties, using the Seepage Model for PA, and evaluating the effects of an alternative drift geometry representing a partially collapsed drift using the Disturbed Drift Seepage Submodel. Seepage of water into waste-emplacement drifts is considered one of the principal factors having the greatest impact of long-term safety of the repository system (CRWMS M andO 2000 [153225], Table 4-1). This AMR supports the analysis and simulation that are used by PA to develop the probability distribution of water seepage into drift, and is therefore a model of primary (Level 1) importance (AP-3.15Q, ''Managing Technical Product Inputs''). The intended purpose of the Seepage Model for PA is to support: (1) PA; (2) Abstraction of Drift-Scale Seepage; and (3) Unsaturated Zone (UZ) Flow and Transport Process Model Report (PMR). Seepage into drifts is evaluated by applying numerical models with stochastic representations of hydrological properties and performing flow simulations with multiple realizations of the permeability field around the drift. The Seepage Model for PA uses the distribution of permeabilities derived from air injection testing in niches and in the cross drift to

Salt Repository Project waste emplacement mode decision paper: Revison 1

International Nuclear Information System (INIS)

1987-08-01

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

IGNEOUS INTRUSION IMPACTS ON WASTE PACKAGES AND WASTE FORMS

International Nuclear Information System (INIS)

Bernot, P.

2004-01-01

The purpose of this model report is to assess the potential impacts of igneous intrusion on waste packages and waste forms in the emplacement drifts at the Yucca Mountain Repository. The models are based on conceptual models and includes an assessment of deleterious dynamic, thermal, hydrologic, and chemical impacts. The models described in this report constitute the waste package and waste form impacts submodel of the Total System Performance Assessment for the License Application (TSPA-LA) model assessing the impacts of a hypothetical igneous intrusion event on the repository total system performance. This submodel is carried out in accordance with Technical Work Plan for Waste Form Degradation Modeling, Testing, and Analyses in Support of LA (BSC 2004 [DIRS:167796]) and Total System Performance Assessment-License Application Methods and Approaches (BSC 2003 [DIRS: 166296]). The technical work plan was prepared in accordance with AP-2.27Q, Planning for Science Activities. Any deviations from the technical work plan are documented in the following sections as they occur. The TSPA-LA approach to implementing the models for waste package and waste form response during igneous intrusion is based on identification of damage zones. Zone 1 includes all emplacement drifts intruded by the basalt dike, and Zone 2 includes all other emplacement drifts in the repository that are not in Zone 1. This model report will document the following model assessments: (1) Mechanical and thermal impacts of basalt magma intrusion on the invert, waste packages and waste forms of the intersected emplacement drifts of Zone 1. (2) Temperature and pressure trends of basaltic magma intrusion intersecting Zone 1 and their potential effects on waste packages and waste forms in Zone 2 emplacement drifts. (3) Deleterious volatile gases, exsolving from the intruded basalt magma and their potential effects on waste packages of Zone 2 emplacement drifts. (4) Post-intrusive physical

Heat transfer effects in vertically emplaced high level nuclear waste container

International Nuclear Information System (INIS)

Moujaes, S.F.; Lei, Y.M.

1994-01-01

Modeling free convection heat transfer in an cylindrical annular enclosure is still an active area of research and an important problem to be addressed in the high level nuclear waste repository. For the vertically emplaced waste container, the air gap which is between the container shell and the rock borehole, have an important role of dissipating heat to surrounding rack. These waste containers are vertically emplaced in the borehole 300 meters below ground, and in a horizontal grid of 30 x 8 meters apart. The borehole will be capped after the container emplacement. The expected initial heat generated is between 3--4.74 kW per container depending on the type of waste. The goal of this study is to use a computer simulation model to find the borehole wall, air-gap and the container outer wall temperature distributions

Emplacement feasibility of a multi-tier, expanded capacity repository at Yucca Mountain, Nevada USA

International Nuclear Information System (INIS)

Apted, Michael; Kessler, John; Fairhurst, Charles

2008-01-01

A geological repository at Yucca Mountain has been proposed for the disposal of spent fuel from the US commercial reactors and other radioactive waste. A legislative capacity of 70,000 MTHM has been set by the Nuclear Waste Policy Act of 1982, including 63,000 MTHM of commercial spent nuclear fuel (CSNF), the projected amount of CSNF that will be produced by about 2014. Policy issues remain as to how to handle waste that is generated beyond 2014 from a growing nuclear industry in the US. The Electric Power Research Institute (EPRI) is independently evaluating the technical, rather than legislative, limit of CSNF that could be safely disposed at Yucca Mountain. Geological, thermal management, safety and cost factors have been recently evaluated by EPRI (2006; 2007) for grouped emplacement drifts and/or a multi-tier repository. EPRI's evaluation of emplacement feasibility for a multi-tier concept is described here. Expanded capacity concepts as envisioned for Yucca Mountain (EPRI, 2006; 2007) assume excavation of one or two additional levels of drifts parallel to or above and/or below the original drift excavations. For the latter multi-tier concept each 'tier' or 'level' would essentially replicate the original layer with a 30-m separation between tiers. This arrangement essentially doubles or triples the capacity of the repository for a two- or three-tier design, respectively. The main issues that affect the feasibility of expanded capacity design are; (i) ventilation requirements; (ii) radiation hazards; (iii) thermal and thermo-mechanical constraints. (i)Ventilation: The repository design involves waste packages mounted in close proximity to each other in 600-m long drifts that remain open and actively ventilated for at least 50-100 years. Analyses,conservatively assuming that all three repository levels operate simultaneously, indicate no technological obstacles in meeting ventilation requirements for sustained simultaneous operation ba sed on current industrial

Conceptual designs of automated systems for underground emplacement and retrieval of nuclear waste

International Nuclear Information System (INIS)

Slocum, A.H.; Hou, W.M.; Park, K.; Hochmuth, C.; Thurston, D.C.

1987-01-01

Current designs of underground nuclear waste repositories have not adequately addressed the possibility of automated, unmanned emplacement and retrieval. This report will present design methodologies for development of an automated system for underground emplacement of nuclear waste. By scaling generic issues to different repositories, it is shown that a two vehicle automated waste emplacement/retrieval system can be designed to operate in a fail-safe mode. Evaluation of cost at this time is not possible. Significant gains in worker safety, however, can be realized by minimizing the possibility of human exposure

Monte-Carlo based comparison of the personal dose for emplacement scenarios of spent nuclear fuel casks in generic deep geological repositories

Energy Technology Data Exchange (ETDEWEB)

Suarez, Hector Sauri; Becker, Franz; Metz, Volker [Karlsruhe Institute of Technology (KIT), Eggenstein-Leopoldshafen (Germany). Inst. for Nuclear Waste Disposal (INE); Pang, Bo [Karlsruhe Institute of Technology (KIT), Eggenstein-Leopoldshafen (Germany). Inst. for Nuclear Waste Disposal (INE); Shenzhen Univ. (China). College of Physics and Energy

2017-06-15

In the operational phase of a deep geological disposal facility for high-level nuclear waste, the radiation field in the vicinity of a waste cask is influenced by the backscattered radiation of the surrounding walls of the emplacement drift. For a comparison of disposal of spent nuclear fuel in various host rocks, it is of interest to investigate the influence of the surrounding materials on the radiation field and the personal radiation exposure. In this generic study individual dosimetry of personnel involved in emplacement of casks with spent nuclear fuel in drifts in rock salt and in a clay formation was modelled.

Near-drift thermal analysis including combined modes of conduction, convection, and radiation

International Nuclear Information System (INIS)

Ho, C.K.; Francis, N.D.

1995-01-01

The performance of waste packages containing high-level nuclear wastes at underground repositories such as the potential repository at Yucca Mountain, Nevada, depends, in part, on the thermodynamic environment immediately surrounding the buried waste packages. For example, degradation of the waste packages can be caused by corrosive and microbial processes, which are influenced by both the relative humidity and temperature within the emplacement drifts. In this paper, the effects of conduction, convection, and radiation are investigated for a heat-generating waste package in an empty-drift. Simulations explicitly modeling radiation from the waste package to the drift wall are compared simulations using only conduction. Temperatures, relative humidities, and vapor mass fractions are compared at various locations within the drift. In addition, the effects of convection on relative humidity and moisture distribution within the drift are presented

Heat transfer effects in vertically emplaced high level nuclear waste container

International Nuclear Information System (INIS)

Moujaes, S.F.; Lei, Y.M.

1994-01-01

Modeling free convection heat transfer in a cylindrical annular enclosure is still an active area of research and an important problem to be addressed in the high level nuclear waste repository. For the vertically emplaced waste container, the air gap which is between the container shell and the rock borehole, have an important role of dissipating heat to surrounding rock. These waste containers are vertically emplaced in the borehole 300 meters just below ground, and in a horizontal grid of 30 x 8 meters apart. The borehole will be capped after the container emplacement. The expected initial heat generated is between 3-4.74 kW per container depending on the type of waste. The goal of this study is to use a computer simulation model to find the borehole wall, air-gap and the container outer wall temperature distributions. The borehole wall temperature history has been found in the previous study, and was estimated to reach a maximum temperature of about 218 degrees C after 18 years from the emplacement. The temperature history of the rock surface is then used for the air-gap simulation. The problem includes convection and radiation heat transfer in a vertical enclosure. This paper will present the results of the convection in the air-gap over one thousand years after the containers' emplacement. During this long simulation period it was also observed that a multi-cellular air flow pattern can be generated in the air gap

Defense High Level Waste Disposal Container System Description Document

International Nuclear Information System (INIS)

2000-01-01

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. U.S. 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 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 which to identify the disposal container and its contents. Different materials

Simulation of heat transfer around a canister placed horizontally in a drift

International Nuclear Information System (INIS)

Moujaes, S.; Bhargava, A.

1994-01-01

The Yucca Mountain Site Characterization Project is investigating the feasibility of locating a high level radioactive nuclear waste repository at Yucca Mountain, Nevada. The bore hole and the in-drift waste emplacement schemes are under evaluation as potential repository drift geometries. This paper presents a two-dimensional finite element thermal analysis of the nuclear waste canister placed horizontally in a drift. Simulation has been carried out for 1000 years and the peak temperatures at the walls of the drift and at the center of the canister have been determined. The effect of the three modes of heat transfer, conduction, natural convection and radiation, is also discussed

EVALUATION OF WASTE PACKAGE EXTERNAL ENVIRONMENTAL CONDITION STUDY

International Nuclear Information System (INIS)

E. N. Lindner and E. F. Dembowski

1998-01-01

The U. S. Department of Energy (DOE) is studying Yucca Mountain as the possible site for a permanent underground repository for disposal of spent nuclear fuel (SNF) and other high-level waste (HLW). The emplacement of high-level radioactive waste in Yucca Mountain will release a large amount of heat into the rock above and below the repository. Due to this heat, the rock temperature will rise, and then decrease when the production of decay heat falls below the rate at which heat escapes from the hot zone. In addition to raising the rock temperature, the heat will vaporize water, which will condense in cooler regions. The condensate water may drain back toward the emplacement drifts or it may ''shed'' through the pillars between emplacement drifts. Other effects, such as coupled chemical and mechanical processes, may influence the movement of water above, within, and below the emplacement drifts. This study examined near field environmental parameters that could have an effect on the waste package, including temperature, humidity, seepage rate, pH of seepage, chemistry (dissolved salts/minerals) of seepage, composition of drift atmosphere, colloids, and biota. This report is a Type I analysis performed in support of the development of System Description Documents (SDDs). A Type I analysis is a quantitative or qualitative analysis that may fulfill any of a variety of purposes associated with the Monitored Geologic Repository (MGR), other than providing direct analytical support for design output documents. A Type I analysis may establish design input, as defined in the ''Quality Assurance Requirements and Description'' (QARD) (DOE 1998). This study establishes a technical basis for emplacement drift (i.e. at the waste package surface) environment criteria to be considered in the development of the waste package design. The information will support development of several SDDs and resolve emplacement drift external environment questions in the criteria of those

SEEPAGE INTO DRIFTS IN UNSATRUATED FRACTURED ROCK AT YUCCA MOUNTAIN

International Nuclear Information System (INIS)

JENS BIRHOLZER; GUOMIN LI; CHIN-FU TSANG; YVONNE TSANG

1998-01-01

An important issue for the long-term performance of underground nuclear waste repositories is the rate of seepage into the waste emplacement drifts. A prediction of the future seepage rate is particularly complicated for the potential repository site at Yucca Mountain, Nevada, as it is located in thick, partially saturated, fractured tuff formations. The long-term situation in the drifts several thousand years after waste emplacement will be characterized by a relative humidity level close to or equal to 100%. as the drifts will be sealed and unventilated, and the waste packages will have cooled. The underground tunnels will then act as capillary barriers for the unsaturated flow, ideally diverting water around them, if the capillary forces are stronger than gravity and viscous forces. Seepage into the drifts will only be possible if the hydraulic pressure in the rock close to the drift walls increases to positive values; i.e., the flow field becomes locally saturated. In the present work, we have developed and applied a methodology to study the potential rate of seepage into underground cavities embedded in a variably saturated, heterogeneous fractured rock formation. The fractured rock mass is represented as a stochastic continuum where the fracture permeabilities vary by several orders of magnitude. Three different realizations of random fracture permeability fields are generated, with the random permeability structure based on extensive fracture mapping, borehole video analysis, and in-situ air permeability testing. A 3-D numerical model is used to simulate the heterogeneous steady-state flow field around the drift, with the drift geometry explicitly represented within the numerical discretization grid. A variety of flow scenarios are considered assuming present-day and future climate conditions at Yucca Mountain. The numerical study is complemented by theoretical evaluations of the drift seepage problem, using stochastic perturbation theory to develop a better

Drift-Scale THC Seepage Model

International Nuclear Information System (INIS)

C.R. Bryan

2005-01-01

The purpose of this report (REV04) is to document the thermal-hydrologic-chemical (THC) seepage model, which simulates the composition of waters that could potentially seep into emplacement drifts, and the composition of the gas phase. The THC seepage model is processed and abstracted for use in the total system performance assessment (TSPA) for the license application (LA). This report has been developed in accordance with ''Technical Work Plan for: Near-Field Environment and Transport: Coupled Processes (Mountain-Scale TH/THC/THM, Drift-Scale THC Seepage, and Post-Processing Analysis for THC Seepage) Report Integration'' (BSC 2005 [DIRS 172761]). The technical work plan (TWP) describes planning information pertaining to the technical scope, content, and management of this report. The plan for validation of the models documented in this report is given in Section 2.2.2, ''Model Validation for the DS THC Seepage Model,'' of the TWP. The TWP (Section 3.2.2) identifies Acceptance Criteria 1 to 4 for ''Quantity and Chemistry of Water Contacting Engineered Barriers and Waste Forms'' (NRC 2003 [DIRS 163274]) as being applicable to this report; however, in variance to the TWP, Acceptance Criterion 5 has also been determined to be applicable, and is addressed, along with the other Acceptance Criteria, in Section 4.2 of this report. Also, three FEPS not listed in the TWP (2.2.10.01.0A, 2.2.10.06.0A, and 2.2.11.02.0A) are partially addressed in this report, and have been added to the list of excluded FEPS in Table 6.1-2. This report has been developed in accordance with LP-SIII.10Q-BSC, ''Models''. This report documents the THC seepage model and a derivative used for validation, the Drift Scale Test (DST) THC submodel. The THC seepage model is a drift-scale process model for predicting the composition of gas and water that could enter waste emplacement drifts and the effects of mineral alteration on flow in rocks surrounding drifts. The DST THC submodel uses a drift

Drift Degradation Analysis

Energy Technology Data Exchange (ETDEWEB)

Dwayne C. Kicker

2001-09-28

A statistical description of the probable block sizes formed by fractures around the emplacement drifts has been developed for each of the lithologic units of the repository host horizon. A range of drift orientations with the drift azimuth varied in 15{sup o} increments has been considered in the static analysis. For the quasi-static seismic analysis, and the time-dependent and thermal effects analysis, two drift orientations have been considered: a drift azimuth of 105{sup o} and the current emplacement drift azimuth of 75{sup o}. The change in drift profile resulting from progressive deterioration of the emplacement drifts has been assessed both with and without backfill. Drift profiles have been determined for four different time increments, including static (i.e., upon excavation), 200 years, 2,000 years, and 10,000 years. The effect of seismic events on rock fall has been analyzed. Block size distributions and drift profiles have been determined for three seismic levels, including a 1,000-year event, a 5,000-year event, and a 10,000-year event. Data developed in this modeling and analysis activity have been entered into the TDMS (DTN: MO0109RDDAAMRR.003). The following conclusions have resulted from this drift degradation analysis: (1) The available fracture data are suitable for supporting a detailed key block analysis of the repository host horizon rock mass. The available data from the north-south Main Drift and the east-west Cross Drift provide a sufficient representative fracture sample of the repository emplacement drift horizon. However, the Tptpln fracture data are only available from a relatively small section of the Cross Drift, resulting in a smaller fracture sample size compared to the other lithologic units. This results in a lower degree of confidence that the key block data based on the Tptpln data set is actually representative of the overall Tptpln key block population. (2) The seismic effect on the rock fall size distribution for all events

Sediment mechanical response due to emplacement of a waste canister

International Nuclear Information System (INIS)

Karnes, C.H.; Dawson, P.R.; Silva, A.J.; Brown, W.T.

1980-01-01

Preliminary studies have been conducted to determine the interaction between a waste canister and seabed sediment during and after emplacement. Empirical and approximate methods for determining the depth reached by a freefall penetrator indicate that a boosted penetrator emplacement method may be necessary. Hole closure is necessary, but has not been verified because calculations and laboratory experiments show sensitivity to boundary conditions which control the degree of dynamic hole closure. Laboratory studies show that closure will take place by creep deformation but closure times in seabed environments are uncertain. For assumed thermomechanical properties of sediments, it is shown that a heat generating waste canister will probably not move a significant distancce during the heat generation period

Seepage into drifts with mechanical degradation

International Nuclear Information System (INIS)

Li, Guomin; Tsang, Chin-Fu

2002-01-01

Seepage into drifts in unsaturated tuff is an important issue for the long-term performance of the potential nuclear waste repository at Yucca Mountain, Nevada. Drifts in which waste packages will potentially be emplaced are subject to degradation in the form of rockfall from the drift ceiling induced by stress relief, seismic, or thermal effects. The objective of this study is to calculate seepage rates for various drift-degradation scenarios and for different values of percolation flux for the Topopah Spring middle nonlithophysal (Tptpmn) and the Topopah Spring lower lithophysal (Tptpll) units. Seepage calculations are conducted by (1) defining a heterogeneous permeability model on the drift scale that is consistent with field data, (2) selecting calibrated parameters associated with the Tptpmn and Tptpll units, and (3) simulating seepage on detailed degraded-drift profiles, which were obtained from a separate rock mechanics engineering analysis. The simulation results indicate (1) that the seepage threshold (i.e., the percolation flux at which seepage first occurs) is not significantly changed by drift degradation, and (2) the degradation-induced increase in seepage above the threshold is influenced more by the shape of the cavity created by rockfall than the rockfall volume

Thermal impact of waste emplacement and surface cooling associated with geologic disposal of nuclear waste

International Nuclear Information System (INIS)

Wang, J.S.Y.; Mangold, D.C.; Spencer, R.K.; Tsang, C.F.

1982-01-01

The age of nuclear waste - the length of time between its removal from the reactor cores and its emplacement in a repository - is a significant factor in determining the thermal loading of a repository. The surface cooling period as well as the density and sequence of waste emplacement affects both the near-field repository structure and the far-field geologic environment. To investigate these issues, a comprehensive review was made of the available literature pertaining to thermal effects and thermal properties of mined geologic repositories. This included a careful evaluation of the effects of different surface cooling periods of the wastes, which is important for understanding the optimal thermal loading of a repository. The results led to a clearer understanding of the importance of surface cooling in evaluating the overall thermal effects of a radioactive waste repository. The principal findings from these investigations are summarized in this paper

Thermal simulation of drift emplacement (TSS): In-situ instrumentation and numerical modeling of stress measurement methods

International Nuclear Information System (INIS)

Heusermann, S.

1988-01-01

In the course of the planned demonstration test Thermal Simulation of Drift Emplacement (TSS) BGR is carrying out in-situ-measurements of rock stresses, rock deformability and permeability of salt rock and backfill material. The following techniques developed and proved by BGR during the last years are planned to be used in the TSS project: overcoring technique, dilatometer technique, hard inclusion technique, slot-cutting techniques, large-flatjack technique, compensation tests in laboratory, vacuum tests, injection tests, and tracer tests. The purpose of measurements is to determine: the initial stress state; stress gradients around test drifts; stress change caused by mining activities, by creep and stress relaxation and by temperature; the in-situ load-deformation behavior of rock salt; the permeability of rock salt around test drifts; the compaction behavior of backfill material; and the load-deformation behavior of rock salt and borehole grout in laboratory tests

Evolution of repository and waste package designs for Yucca Mountain disposal system for spent nuclear fuel and high-level radioactive waste

International Nuclear Information System (INIS)

Rechard, Rob P.; Voegele, Michael D.

2014-01-01

This paper summarizes the evolution of the engineered barrier design for the proposed Yucca Mountain disposal system. Initially, the underground facility used a fairly standard panel and drift layout excavated mostly by drilling and blasting. By 1993, the layout of the underground facility was changed to accommodate construction by a tunnel boring machine. Placement of the repository in unsaturated zone permitted an extended period without backfilling; placement of the waste package in an open drift permitted use of much larger, and thus hotter packages. Hence in 1994, the underground facility design switched from floor emplacement of waste in small, single walled stainless steel or nickel alloy containers to in-drift emplacement of waste in large, double-walled containers. By 2000, the outer layer was a high nickel alloy for corrosion resistance and the inner layer was stainless steel for structural strength. Use of large packages facilitated receipt and disposal of high volumes of spent nuclear fuel. In addition, in-drift package placement saved excavation costs. Options considered for in-drift emplacement included different heat loads and use of backfill. To avoid dripping on the package during the thermal period and the possibility of localized corrosion, titanium drip shields were added for the disposal drifts by 2000. In addition, a handling canister, sealed at the reactor to eliminate further handling of bare fuel assemblies, was evaluated and eventually adopted in 2006. Finally, staged development of the underground layout was adopted to more readily adjust to changes in waste forms and Congressional funding. - Highlights: • Progression of events associated with repository design to accommodate tunnel boring machine and in-drift waste package emplacement are discussed. • Change in container design from small, single-layered stainless steel vessel to large, two-layered nickel alloy vessel is discussed. • The addition of drip shield to limit the

Drift-Scale THC Seepage Model

Energy Technology Data Exchange (ETDEWEB)

C.R. Bryan

2005-02-17

The purpose of this report (REV04) is to document the thermal-hydrologic-chemical (THC) seepage model, which simulates the composition of waters that could potentially seep into emplacement drifts, and the composition of the gas phase. The THC seepage model is processed and abstracted for use in the total system performance assessment (TSPA) for the license application (LA). This report has been developed in accordance with ''Technical Work Plan for: Near-Field Environment and Transport: Coupled Processes (Mountain-Scale TH/THC/THM, Drift-Scale THC Seepage, and Post-Processing Analysis for THC Seepage) Report Integration'' (BSC 2005 [DIRS 172761]). The technical work plan (TWP) describes planning information pertaining to the technical scope, content, and management of this report. The plan for validation of the models documented in this report is given in Section 2.2.2, ''Model Validation for the DS THC Seepage Model,'' of the TWP. The TWP (Section 3.2.2) identifies Acceptance Criteria 1 to 4 for ''Quantity and Chemistry of Water Contacting Engineered Barriers and Waste Forms'' (NRC 2003 [DIRS 163274]) as being applicable to this report; however, in variance to the TWP, Acceptance Criterion 5 has also been determined to be applicable, and is addressed, along with the other Acceptance Criteria, in Section 4.2 of this report. Also, three FEPS not listed in the TWP (2.2.10.01.0A, 2.2.10.06.0A, and 2.2.11.02.0A) are partially addressed in this report, and have been added to the list of excluded FEPS in Table 6.1-2. This report has been developed in accordance with LP-SIII.10Q-BSC, ''Models''. This report documents the THC seepage model and a derivative used for validation, the Drift Scale Test (DST) THC submodel. The THC seepage model is a drift-scale process model for predicting the composition of gas and water that could enter waste emplacement drifts and the effects of mineral

STRUCTURAL CALCULATION OF AN EMPLACEMENT PALLET STATICALLY LOADED BY A WASTE PACKAGE

International Nuclear Information System (INIS)

S. Mastilovic

2000-01-01

The purpose of this calculation is to determine the structural response of the emplacement pallet (EP) subjected to static load from the mounted waste package (WP). The scope of this document is limited to reporting the calculation results in terms of stress intensity magnitudes. This calculation is associated with the waste emplacement systems design; calculations are performed by the Waste Package Design group. AP-3.12Q, Revision 0, ICN 0, Calculations, is used to perform the calculation and develop the document. The finite element solutions are performed by using the commercially available ANSYS Version (V) 5.4 finite element code. The results of these calculations are provided in terms of maximum stress intensity magnitudes

Quarter-scale modeling of room convergence effects on CH [contact-handled] TRU drum waste emplacements using WIPP [Waste Isolation Pilot Plant] reference design geometries

International Nuclear Information System (INIS)

VandeKraats, J.

1987-11-01

This study investigates the effect of horizontal room convergence on CH waste packages emplaced in the WIPP Reference Design geometry (rooms 13 feet high by 33 feet wide, with minus 3/8 inch screened backfill emplaced over and around the waste packages) as a function of time. Based on two tests, predictions were made with regard to full-scale 6-packs emplaced in the Reference Design geometry. These are that load will be transmitted completely through the stack within the first five years after waste emplacement and all drums in all 6-packs will be affected; that virtually all drums will show some deformation eight years after emplacement; that some drums may breach before the eighth year after emplacement has elapsed; and that based on criteria developed during testing, it is predicted that 1% of the drums emplaced will be breached after 8 years and, after 15 years, approximately 12% of the drums are predicted to be breached. 8 refs., 41 figs., 3 tabs

Emplacement technology for the direct disposal of spent fuel into deep vertical boreholes

International Nuclear Information System (INIS)

Bollingerfehr, W.; Filbert, W.; Wehrmann, J.

2008-01-01

In the early sixties it was decided to investigate salt formations on its suitability to host heat generating radioactive waste in Germany. In the reference repository concept consequently the emplacement of vitrified waste canisters in deep vertical boreholes inside a salt mine was considered whereas spent fuel should be disposed of in self shielding casks (type POLLUX) in horizontal drifts. The POLLUX casks, 65 t heavy carbon steel casks, will be laid down on the floor of a horizontal drift in one of the disposal zones to be constructed in the salt dome at the 870 m level. The space between casks and drift walls will be backfilled with crushed salt. The transport, the handling und the emplacement of POLLUX casks were subject of successfully performed demonstration and in situ tests in the nineties and resulted in an adjustment of the atomic law. The borehole disposal concept comprises the emplacement of unshielded canisters with vitrified HLW in boreholes with a diameter of 60 cm and a depth of up to 300 m. In order to facilitate the fast encapsulation of the waste canister by the host rock (rock salt), no lining of the boreholes is planned. With regard to harmonize and optimize the emplacement technology for both categories of packages (vitrified waste and spent fuel) alternatives were developed. In this context the borehole emplacement technique for consolidated spent fuel as already foreseen for high-level reprocessing waste was reconsidered. This review resulted in the design of a new disposal package, a fuel rod canister (type 'BSK 3'), and an appropriate modified transport and emplacement technology. This concept (called BSK 3-concept) provides the following optimization possibilities: (i) A new steel canister of the same diameter (43 cm) as the standardized HLW canisters applied for high-level waste and compacted technological waste from reprocessing abroad can be filled with fuel rods of 3 PWR or 9 BWR fuel assemblies. (II) The standardized canister

Uncertainty and sensitivity results for pre-waste-emplacement groundwater travel time

International Nuclear Information System (INIS)

Kaplan, P.G.

1992-01-01

In this paper uncertainty and sensitivity analyses for pre-waste-emplacement groundwater travel time conducted. Although preliminary, a numbed of interesting results were obtained. Uncertainty in the ground water travel time statistics, as measured by the coefficient of variation, increases and then decrease as the modeled system transitions from matrix-dominated to fracture-dominated flow. The uncertainty analysis also suggests that the median, as opposed to the mean, may be a better indicator of performance with respect to the regulatory criterion. The sensitivity analysis shows a strong correlation between an effective fracture property, fracture porosity, and failure to meet the regulatory pre-waste-emplacement groundwater travel time criterion of 1,000 years

Drift scale thermomechanical analysis for thermal loading and retrievability studies

International Nuclear Information System (INIS)

Tsai, F.C.

1995-01-01

The repository portion of the Mined Geologic Disposal System for the disposal of spent nuclear fuel and high-level radioactive waste is currently in the advanced conceptual design stage. In support of systems studies, a numerical method was used to estimate the stability of emplacement drifts. Thermomechanical analyses, using the Discontinuous Deformation Analysis code, were performed using input data from Yucca Mountain documents. The analysis found that the stresses produced in the rock at thermal loads of 27.4 kilograms uranium per m2 (KgU/m2) would exceed stability criteria and could result in tunnel instability. At thermal loads between 20.5 KgU/m2, the drift is predicted to be stable and its structural integrity remains after thermal loading. In this case, the smaller diameter drift emplacement appears to have better stability. However, local rock spalling may occur. According to the numerical prediction, more rock fall may occur during the retrieval period due to the stress relaxation caused by the rapid cooling in the immediate drift area

Dike/Drift Interactions

Energy Technology Data Exchange (ETDEWEB)

E. Gaffiney

2004-11-23

This report presents and documents the model components and analyses that represent potential processes associated with propagation of a magma-filled crack (dike) migrating upward toward the surface, intersection of the dike with repository drifts, flow of magma in the drifts, and post-magma emplacement effects on repository performance. The processes that describe upward migration of a dike and magma flow down the drift are referred to as the dike intrusion submodel. The post-magma emplacement processes are referred to as the post-intrusion submodel. Collectively, these submodels are referred to as a conceptual model for dike/drift interaction. The model components and analyses of the dike/drift interaction conceptual model provide the technical basis for assessing the potential impacts of an igneous intrusion on repository performance, including those features, events, and processes (FEPs) related to dike/drift interaction (Section 6.1).

Dike/Drift Interactions

International Nuclear Information System (INIS)

Gaffiney, E.

2004-01-01

This report presents and documents the model components and analyses that represent potential processes associated with propagation of a magma-filled crack (dike) migrating upward toward the surface, intersection of the dike with repository drifts, flow of magma in the drifts, and post-magma emplacement effects on repository performance. The processes that describe upward migration of a dike and magma flow down the drift are referred to as the dike intrusion submodel. The post-magma emplacement processes are referred to as the post-intrusion submodel. Collectively, these submodels are referred to as a conceptual model for dike/drift interaction. The model components and analyses of the dike/drift interaction conceptual model provide the technical basis for assessing the potential impacts of an igneous intrusion on repository performance, including those features, events, and processes (FEPs) related to dike/drift interaction (Section 6.1)

REPOSITORY LAYOUT SUPPORTING DESIGN FEATURE NO.13 - WASTE PACKAGE SELF SHIELDING

International Nuclear Information System (INIS)

Owen, J.

1999-01-01

The objective of this analysis is to develop a repository layout, for Feature No. 13, that will accommodate self-shielding waste packages (WP) with an areal mass loading of 25 metric tons of uranium per acre (MTU/acre). The scope of this analysis includes determination of the number of emplacement drifts, amount of emplacement drift excavation required, and a preliminary layout for illustrative purposes

Investigative study of the underground excavations for a nuclear waste repository in tuff: Nevada Nuclear Waste Storage Investigations Project

International Nuclear Information System (INIS)

St John, C.M.

1987-07-01

Numerical studies were conducted on the behavior of a tuff rock mass within which emplacement drifts for a nuclear waste repository are excavated. The first study evaluated the effects of rockbolting and excavation-induced damage on the behavior of the rock mass round typical drifts. The second study provided a simple means of assessing the significance of drift shape, drift size, and in-situ state of stress on the deformation and stress in the vicinity of drifts for vertical and horizontal emplacement of waste. Neither study considered the effect of heating of the rock mass after emplacement of the waste so the conclusions pertain only to the conditions immediately after excavation of the underground openings. The results of analyses of the rockbolted excavations indicated that rockbolts do not have a significant influence on the states of deformation or stress within the rock mass, and that the rockbolts are subjected to acceptable levels of stress even if installed as close to the face of the excavation as possible. Accordingly, rockbolts were not considered in the study of drift shape, drift size, and the in-situ state of stress. That study indicated that stable openings of the dimensions investigated can be constructed within a tuff rock mass with the properties assumed. Of the parameters investigated, the in-situ state of stress appeared to be most important. Potentially adverse conditions were predicted if the in-situ horizontal stress is very low, but current indications are that it lies within a range which is consistent with good conditions and a stable roof. 28 refs., 49 figs., 11 tabs

EVALUATION OF RISKS AND WASTE CHARACTERIZATION REQUIREMENTS FOR THE TRANSURANIC WASTE EMPLACED IN WIPP DURING 1999

International Nuclear Information System (INIS)

Channell, J.K.; Walker, B.A.

2000-01-01

Specifically this report: 1. Compares requirements of the WAP that are pertinent from a technical viewpoint with the WIPP pre-Permit waste characterization program, 2. Presents the results of a risk analysis of the currently emplaced wastes. Expected and bounding risks from routine operations and possible accidents are evaluated; and 3. Provides conclusions and recommendations

Performance implications of waste package emplacement orientation

International Nuclear Information System (INIS)

Wilder, D.G.

1991-05-01

Emplacement borehole orientation directly impacts many aspects of the Engineered Barrier System (EBS) and interactions with the near field environment. This paper considers the impacts of orientation on the hydrologic portion of the environment and its interactions with the EBS. The hydrologic environment is considered from a conceptual standpoint, the numerical analyses are left for subsequent work. As reported in this paper, several aspects of the hydrological environment are more favorable for long term performance of vertically oriented rather than horizontally oriented Waste Packages. 19 refs., 15 figs

Elements of transport and emplacement system

International Nuclear Information System (INIS)

1981-02-01

This report, undertaken to review proposals for transport, handling and emplacement of high-level radioactive wastes in an underground repository, appropriate to the U.K. context, falls under the headings: basic design concepts; waste block size and configuration; self-shielded or partially shielded blocks; concept of emplacement in long boreholes; concept of emplacement in short boreholes; concept of emplacement in tunnels; methods of emplacement; stages of disposal; repository access by adit, incline or shaft; handling techniques within repository; conventional and radiological safety; costs; areas for further research and development. (U.K.)

Hazard and consequence analysis for waste emplacement at the Waste Isolation Pilot Plant

International Nuclear Information System (INIS)

Gerstner, D.M.; Clayton, S.G.; Farrell, R.F.; McCormick, J.A.; Ortiz, C.; Standiford, D.L.

1996-01-01

The Carlsbad Area Office established and analyzed the safety bases for the design and operations as documented in the WIPP Safety Analysis Report (SAR). Additional independent efforts are currently underway to assess the hazards associated with the long-term (10,000 year) isolation period as required by 40 CFR 191. The structure of the WIPP SAR is unique due to the hazards involved, and the agreement between the State of New Mexico and the DOE regarding SAR content and format. However, the hazards and accident analysis philosophy as contained in DOE-STD-3009-94 was followed as closely as possible, while adhering to state agreements. Hazards associated with WIPP waste receipt, emplacement, and disposal operations were systematically identified using a modified Hazard and Operability Study (HAZOP) technique. The WIPP HAZOP assessed the potential internal, external, and natural phenomena events that can cause the identified hazards to develop into accidents. The hazard assessment identified deviations from the intended design and operation of the waste handling system, analyzed potential accident consequences to the public and workers, estimated likelihood of occurrence, and evaluated associated preventative and mitigative features. It was concluded from the assessment that the proposed WIPP waste emplacement operations and design are sufficient to ensure safety of the public, workers, and environment, over the 35 year disposal phase

Fluid flow and reactive transport around potential nuclear waste emplacement tunnels at Yucca Mountain, Nevada

International Nuclear Information System (INIS)

Spycher, N.F.; Sonnenthal, E.L.; Apps, J.A.

2002-01-01

The evolution of fluid chemistry and mineral alteration around a potential waste emplacement tunnel (drift) is evaluated using numerical modeling. The model considers the flow of water, gas, and heat, plus reactions between minerals, CO 2 gas, and aqueous species, and porosity permeability-capillary pressure coupling for a dual permeability (fractures and matrix) medium. Two possible operating temperature modes are investigated: a ''high-temperature'' case with temperatures exceeding the boiling point of water for several hundred years, and a ''loW--temperature'' case with temperatures remaining below boiling for the entire life of the repository. In both cases, possible seepage waters are characterized by dilute to moderate salinities and mildly alkaline pH values. These trends in fluid composition and mineral alteration are controlled by various coupled mechanisms. For example, upon heating and boiling, CO 2 exsolution from pore waters raises pH and causes calcite precipitation. In condensation zones, this CO 2 redissolves, resulting in a decrease in pH that causes calcite dissolution and enhances feldspar alteration to clays. Heat also enhances dissolution of wallrock minerals leading to elevated silica concentrations. Amorphous silica precipitates through evaporative concentration caused by boiling in the high-temperature case, but does not precipitate in the loW--temperature case. Some alteration of feldspars to clays and zeolites is predicted in the high-temperature case. In both cases, calcite precipitates when percolating waters are heated near the drift. The predicted porosity decrease around drifts in the high-temperature case (several percent of the fracture volume) is larger by at least one order of magnitude than in the low temperature case. Although there are important differences between the two investigated temperature modes in the predicted evolution of fluid compositions and mineral alteration around drifts, these differences are small relative to

High level radioactive waste repositories. Task 3. Review of underground handling and emplacement. 1. Executive summary

Energy Technology Data Exchange (ETDEWEB)

1981-01-01

A review is presented of proposals for transport, handling and emplacement of high-level radioactive waste in an underground repository appropriate to the U.K. context, with particular reference to waste block size and configuration; self-shielded or partially-shielded block; stages of disposal; transport by road/rail to repository site; handling techniques within repository; emplacement in vertical holes or horizontal tunnels; repository access by adit, incline or shaft; conventional and radiological safety; costs; and major areas of uncertainty requiring research or development.

Drift-Scale Coupled Processes (DST and THC Seepage) Models

International Nuclear Information System (INIS)

Dixon, P.

2004-01-01

The purpose of this Model Report (REV02) is to document the unsaturated zone (UZ) models used to evaluate the potential effects of coupled thermal-hydrological-chemical (THC) processes on UZ flow and transport. This Model Report has been developed in accordance with the ''Technical Work Plan for: Performance Assessment Unsaturated Zone'' (Bechtel SAIC Company, LLC (BSC) 2002 [160819]). The technical work plan (TWP) describes planning information pertaining to the technical scope, content, and management of this Model Report in Section 1.12, Work Package AUZM08, ''Coupled Effects on Flow and Seepage''. The plan for validation of the models documented in this Model Report is given in Attachment I, Model Validation Plans, Section I-3-4, of the TWP. Except for variations in acceptance criteria (Section 4.2), there were no deviations from this TWP. This report was developed in accordance with AP-SIII.10Q, ''Models''. This Model Report documents the THC Seepage Model and the Drift Scale Test (DST) THC Model. The THC Seepage Model is a drift-scale process model for predicting the composition of gas and water that could enter waste emplacement drifts and the effects of mineral alteration on flow in rocks surrounding drifts. The DST THC model is a drift-scale process model relying on the same conceptual model and much of the same input data (i.e., physical, hydrological, thermodynamic, and kinetic) as the THC Seepage Model. The DST THC Model is the primary method for validating the THC Seepage Model. The DST THC Model compares predicted water and gas compositions, as well as mineral alteration patterns, with observed data from the DST. These models provide the framework to evaluate THC coupled processes at the drift scale, predict flow and transport behavior for specified thermal-loading conditions, and predict the evolution of mineral alteration and fluid chemistry around potential waste emplacement drifts. The DST THC Model is used solely for the validation of the THC

Effects of the deviation characteristics of nuclear waste emplacement boreholes on borehole liner stresses

International Nuclear Information System (INIS)

Glowka, D.A.

1990-09-01

This report investigates the effects of borehole deviation on the useability of lined boreholes for the disposal of high-level nuclear waste at the proposed Yucca Mountain Repository in Nevada. Items that lead to constraints on borehole deviation include excessive stresses that could cause liner failure and possible binding of a waste container inside the liner during waste emplacement and retrieval operations. Liner stress models are developed for two general borehole configurations, one for boreholes drilled with a steerable bit and one for boreholes drilled with a non-steerable bit. Procedures are developed for calculating liner stresses that arise both during insertion of the liner into a borehole and during the thermal expansion process that follows waste emplacement. The effects of borehole curvature on the ability of the waste container to pass freely inside the liner without binding are also examined. Based on the results, specifications on borehole deviation allowances are developed for specific vertical and horizontal borehole configurations of current interest. 11 refs., 22 figs., 4 tabs

TRANSPORT AND EMPLACEMENT EQUIPMENT DESCRIPTIONS

International Nuclear Information System (INIS)

1997-01-01

The objective and the scope of this document are to list and briefly describe the major mobile equipment necessary for waste package (WP) Transport and Emplacement in the proposed subsurface nuclear waste repository at Yucca Mountain. Primary performance characteristics and some specialized design features of the equipment are explained and summarized in the individual subsections of this document. The Transport and Emplacement equipment described in this document consists of the following: (1) WP Transporter; (2) Reusable Rail Car; (3) Emplacement Gantry; (4) Gantry Carrier; and (5) Transport Locomotive

Subsurface geology of a potential waste emplacement site, Salt Valley Anticline, Grand County, Utah

Science.gov (United States)

Hite, R.J.

1977-01-01

The Salt Valley anticline, which is located about 32 km northeast of Moab, Utah, is perhaps one of the most favorable waste emplacement sites in the Paradox basin. The site, which includes about 7.8 km 2, is highly accessible and is adjacent to a railroad. The anticline is one of a series of northwest-trending salt anticlines lying along the northeast edge of the Paradox basin. These anticlines are cored by evaporites of the Paradox Member of the Hermosa Formation of Middle Pennsylvanian age. The central core of the Salt Valley anticline forms a ridgelike mass of evaporites that has an estimated amplitude of 3,600 m. The evaporite core consists of about 87 percent halite rock, which includes some potash deposits; the remainder is black shale, silty dolomite, and anhydrite. The latter three lithologies are referred to as 'marker beds.' Using geophysical logs from drill holes on the anticline, it is possible to demonstrate that the marker beds are complexly folded and faulted. Available data concerning the geothermal gradient and heatflow at the site indicate that heat from emplaced wastes should be rapidly dissipated. Potentially exploitable resources of potash and petroleum are present at Salt Valley. Development of these resources may conflict with use of the site for waste emplacement.

Subsurface geology of a potential waste emplacement site, Salt Valley Anticline, Grand County, Utah

International Nuclear Information System (INIS)

Hite, R.J.

1977-01-01

The Salt Valley anticline, which is located about 32 km northeast of Moab, Utah, is perhaps one of the most favorable waste emplacement sites in the Paradox basin. The site, which includes about 7.8 km 2 , is highly accessible and is adjacent to a railroad. The anticline is one of a series of northwest-trending salt antilcines lying along the northeast edge of the Paradox basin. These anticlines are cored by evaporites of the Paradox Member of the Hermosa Formation of Middle Pennsylvanian age. The central core of the Salt Valley anticline forms a ridgelike mass of evaporites that has an estimated amplitude of 3,600 m. The evaporite core consists of about 87 percent halite rock, which includes some potash deposits; the remainder is black shale, silty dolomite, and anhydrite. The latter three lithologies are referred to as ''marker beds.'' Using geophysical logs from drill holes on the anticline, it is possible to demonstrate that the marker beds are complexly folded and faulted. Available data concerning the geothermal gradient and heatflow at the site indicate that heat from emplaced wastes should be rapidly dissipated. Potentially exploitable resources of potash and petroleum are present at Salt Valley. Development of these resources may conflict with use of the site for waste emplacement

Waste package performance allocation system study report

International Nuclear Information System (INIS)

Memory, R.D.

1994-01-01

The Waste Package Performance Allocation system study was performed in order to provide a technical basis for the selection of the waste package period of substantially complete containment and its resultant contribution to the overall total system performance. This study began with a reference case based on the current Mined Geologic Disposal System (MGDS) baseline design and added a number of alternative designs. The waste package designs were selected from the designs being considered in detail during Advanced Conceptual Design (ACD). The waste packages considered were multi-barrier packages with a 0.95 cm Alloy 825 inner barrier and a 10, 20, or 45 cm thick carbon steel outer barrier. The waste package capacities varied from 6 to 12 to 21 Pressurized Water Reactor (PWR) fuel assemblies. The vertical borehole and in-drift emplacement modes were also considered, as were thermal loadings of 25, 57, and 114 kW/acre. The repository cost analysis indicated that the 21 PWR in-drift emplacement mode option with the 10 cm and 20 cm outer barrier thicknesses are the least expensive and that the 12 PWR in-drift case has approximately the same cost as the 6 PWR vertical borehole. It was also found that the cost increase from the 10 cm outer barrier waste package to the 20 cm waste package was less per centimeter than the increase from the 20 cm outer barrier waste package to the 45 cm outer barrier waste package. However, the repository cost was nearly linear with the outer barrier thickness for the 21 PWR in-drift case. Finally, corrosion rate estimates are provided and the relationship of repository cost versus waste package lifetime is discussed as is cumulative radionuclide release from the waste package and to the accessible environment for time periods of 10,000 years and 100,000 years

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

International Nuclear Information System (INIS)

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

1980-08-01

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

Analysis of near-field thermal and psychometric waste package environment using ventilation

International Nuclear Information System (INIS)

Danko, G.

1995-03-01

The ultimate objective of the Civilian Radioactive Waste Management System (CRWMS) Program is to safely emplace and isolate the nations' spent nuclear fuel (SNF) and radioactive wastes in a geologic repository. Radioactive waste emplaced in a geologic repository will generate heat, increasing the temperature in the repository. The magnitude of this temperature increase depends upon (1) the heat source, i.e. the thermal loading of the repository, and (2) the geologic and engineered heat transport characteristics of the repository. Thermal management techniques currently under investigation include ventilation of the emplacement drifts during the preclosure period which could last as long as 100 years. Understanding the amount of heat and moisture removed from the emplacement drifts and near-field rock by ventilation, are important in determining performance of the engineered barrier system (EBS), as well as the corrosive environment of the waste packages, and the interaction of the EBS with the near-field host rock. Since radionuclide releases and repository system performance are significantly affected by the corrosion rate related to the psychometric environment, it is necessary to predict the amount of heat and moisture that are removed from the repository horizon using a realistic model for a wide range of thermal loading. This can be realized by coupling the hydrothermal model of the rock mass to a ventilation/climate model which includes the heat and moisture transport on the rock-air interface and the dilution of water vapor in the drift. This paper deals with the development of the coupled model concept, and determination of the boundary conditions for the calculations

Seepage into drifts in unsaturated fractured rock at Yucca Mountain

International Nuclear Information System (INIS)

Birkholzer, Jens; Li, Guomin; Tsang, Chin-Fu; Tsang, Yvonne

1998-01-01

An important issue for the long-term performance of underground nuclear waste repository is the rate of seepage into the waste emplacement drifts. A prediction of the future seepage rate is particularly complicated for the potential repository site at Yucca Mountain, Nevada, as it is located in thick, partially saturated, fractured tuff formations. The long-term situation in the drifts several thousand years after waste emplacement will be characterized by a relative humidity level close to or equal to 100%, as the drifts will be sealed and unventilated, and the waste packages will have cooled. The underground tunnels will then act as capillary barriers for the unsaturated flow, ideally diverting water around them, if the capillary forces are stronger than gravity and viscous forces. Seepage into the drifts will only be possible if the hydraulic pressure in the rock close to the drift walls increases to positive values; i.e., the flow field becomes locally saturated. In the present work, they have developed and applied a methodology to study the potential rate of seepage into underground cavities embedded in a variably saturated, heterogeneous fractured rock formation. The fractured rock mass is represented as a stochastic continuum where the fracture permeabilities vary by several orders of magnitude. Three different realizations of random fracture permeability fields are generated, with the random permeability structure based on extensive fracture mapping, borehole video analysis, and in-situ air permeability testing. A 3-D numerical model is used to simulate the heterogeneous steady-state flow field around the drift, with the drift geometry explicitly represented within the numerical discretization grid. A variety of flow scenarios are considered assuming present-day and future climate conditions at Yucca Mountain. The numerical study is complemented by theoretical evaluations of the drift seepage problem, using stochastic perturbation theory to develop a better

Field investigation of the drift shadow

International Nuclear Information System (INIS)

Su, Grace W.; Kneafsey, Timothy J.; Ghezzehei, Teamrat A.; Marshall, Brian D.; Cook, Paul J.

2005-01-01

A drift shadow is an area immediately beneath an underground void that, in theory, will be relatively drier than the surrounding rockmass. Numerical and analytical models of water flow through unsaturated rock predict the existence of a drift shadow, but field tests confirming its existence have yet to be performed. Proving the existence of drift shadows and understanding their hydrologic and transport characteristics could provide a better understanding of how contaminants move in the subsurface if released from waste emplacement drifts such as the proposed nuclear waste repository at Yucca Mountain, Nevada. We describe the field program that will be used to investigate the existence of a drift shadow and the corresponding hydrological process at the Hazel-Atlas silica-sandmine located at the Black Diamond Mines Regional Preserve in Antioch, California. The location and configuration of this mine makes it an excellent site to observe and measure drift shadow characteristics. The mine is located in a porous sandstone unit of the Domengine Formation, an approximately 230 meter thick series of interbedded Eocene-age shales, coals, and massive-bedded sandstones. The mining method used at the mine required the development of two parallel drifts, one above the other, driven along the strike of the mined sandstone stratum. This configuration provides the opportunity to introduce water into the rockmass in the upper drift and to observe and measure its flow around the underlying drift. The passive and active hydrologic tests to be performed are described. In the passive method, cores will be obtained in a radial pattern around a drift and will be sectioned and analyzed for in-situ water content and chemical constituents. With the active hydrologic test, water will be introduced into the upper drift of the two parallel drifts and the flow of the water will be tracked as it passes near the bottom drift. Tensiometers, electrical resistance probes, neutron probes, and ground

Preliminary uncertainty analysis of pre-waste-emplacement groundwater travel times for a proposed repository in basalt

International Nuclear Information System (INIS)

Clifton, P.M.; Arnett, R.C.

1984-01-01

Preliminary uncertainty analyses of pre-waste-emplacement groundwater travel times are presented for a potential high-level nuclear waste repository in the deep basalts beneath the Hanford Site, Washington State. The uncertainty analyses are carried out by means of a Monte Carlo technique, which requires the uncertain inputs to be described as either random variables or spatial stochastic processes. Pre-waste-emplacement groundwater travel times are modeled in a continuous, flat-lying basalt flow top that is assumed to overlie the repository horizon. Two-dimensional, steady state groundwater flow is assumed, and transmissivity, effective thickness, and regional hydraulic gradient are considered as uncertain inputs. Groundwater travel time distributions corresponding to three groundwater models are presented and compared. Limitations of these preliminary simulation results are discussed in detail

Concepts for operational period panel seal design at the Waste Isolation Pilot Plant

International Nuclear Information System (INIS)

Hansen, F.D.; Lin, M.S.; Van Sambeek, L.L.

1993-07-01

Concepts for underground panel or drift seals at the Waste Isolation Pilot Plant are developed to satisfy sealing requirements of the operational period. The concepts are divided into two groups. In the ''NOW'' group, design concepts are considered in which a sleeve structure is installed in the panel access immediately after excavation and before waste is emplaced. In the ''LATER'' group, no special measures are taken during excavation or before waste emplacement; the seal is installed at a later date, perhaps up to 35 years after the drift is excavated. Three concepts are presented in both the NOW and LATER groups. A rigid sleeve, a yielding sleeve, and steel rings with inflatable tubes are proposed as NOW concepts. One steel ring concept and two concrete monoliths are proposed for seals emplaced in older drifts. Advantages and disadvantages are listed for each concept. Based on the available information, it appears most feasible to recommend a LATER concept using a concrete monolith as a preferred seal for the operational period. Each concept includes the potential of remedial grout and/or construction of a chamber that could be used for monitoring leakage from a closed panel during the operational period. Supporting in situ demonstrations of elements of the concepts are recommended

Evaluation of the post-emplacement environment of high level radioactive waste packages at Yucca Mountain, Nevada

International Nuclear Information System (INIS)

Glassley, W.

1989-01-01

Evaluation of the post-emplacement environment around high-level radioactive waste containers is required by federal regulations. The information derived from this evaluation will be used to determine the service performance of the waste containers, the chemical and hydrological conditions that may influence radionuclide release and transport if containers are breached, and retrievability of the waste containers prior to closure of the repository. Laboratory studies, numerical simulations, and field experiments and tests are used to provide data necessary for this evaluation. Results obtained to date demonstrate that the post-emplacement environment in the welded tuff at Yucca Mountain, Nevada maintains relatively benign chemical features (i.e., near neutral pH, low concentrations of dissolved species) for most scenarios. The hydrological environment appears to be one of low flow volume and rates for the expected condition of an unsaturated medium. Emplacement borehole stability will be a function of fracture density and orientation, which may be influenced by microcrack development. Field studies and numerical simulations are in progress that will extend the results of laboratory studies to long time periods. The extent to which chemical, hydrological and mechanical processes can be adequately coupled through numerical simulations remains a matter of concern

Conceptual designs for waste packages for horizontal or vertical emplacement in a repository in salt for reference in the site characterization plan

International Nuclear Information System (INIS)

1987-06-01

This report includes the options of horizontal and vertical emplacement, the addition of a phased repository, an additional waste form (intact spent fuel), revised geotechnical data appropriate for the Deaf Smith County site, new corrosion data for the container, and new repository design data. The waste package consists of waste form and canister within a thick-walled, low-carbon steel container surrounded by packing. The container is a hollow cylinder with a flat head welded to each end. The design concepts for the waste container or vertical and horizontal emplacement are identical. This report discusses the results of analyses of aspects of the reference waste package concept needing changes because of new data and information believed applicable to the Deaf Smith County site. Included are waste package conceptual designs or (1) the reference defense high-level waste form from the Savannah River Plant; (2) intact spent fuel with our pressurized-water-reactor or nine boiling-water-reactor assemblies per package for emplacement during Phase 1 of repository operation; and (3) spent fuel which has been disassembled and consolidated into a segmented cylindrical canister with rods from either 12 pressurized-water-reactor or 30 boiling-water-reactor assemblies per package for emplacement during Phase 2. 30 refs., 61 figs., 30 tabs

Thermomechanical analysis of underground excavations in the vicinity of a nuclear waste isolation panel

International Nuclear Information System (INIS)

St John, C.M.

1987-06-01

This report summarizes the results of a series of analyses of excavations in the vicinity of waste emplacement panels. Specific consideration is given to the access drifts running between adjacent emplacement panels, the drift intersection at the entrance to the emplacement panels, and the waste emplacement excavations. Both horizontal and vertical emplacement models are considered, but greater emphasis is placed on the former. Three numerical modeling procedures were used in this study: a finite-element model was used for three-dimensional stress analysis of the tunnel intersection, a model based on the closed-form solution for point heat sources was used to predict temperatures and stresses in the vicinity of the emplacement panel, and simple two-dimensional boundary-element models were used to predict temperatures and stresses around excavations of various shapes. The results of two-dimensional stress analyses were postprocessed to determine the extent to which the strength of a rock mass, containing a set of vertical joints, was exceeded. The results presented in this report do not indicate that there will be any particular stability problems at the tunnel intersection investigated. Further, the effect of waste emplacement within the adjacent panels is to decrease the vertical rock stresses and increase the horizontal rock stresses at the intersection. These stress changes will tend to enhance the stability of larger-span excavations, including the tunnel intersection and the alcoves necessary for horizontal emplacement of waste canisters. The relatively high horizontal stresses experienced by the access were identified as a potential concern. However, evaluation of recent data on the thermomechanical properties of the rock mass modeled here has indicated that the stress changes will not be as severe as stated herein

Coupled Thermal-Hydrologic-Chemical Coupled Model for In-Drift Disposal Test

Energy Technology Data Exchange (ETDEWEB)

Jordan, Amy B. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Zyvoloski, George Anthony [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Weaver, Douglas James [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Otto, Shawn [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Stauffer, Philip H. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2016-09-06

The simulation work presented in this report supports DOE-NE Used Fuel Disposition Campaign (UFDC) goals related to the development of drift scale in-situ field testing of heat-generating nuclear waste (HGNW) in salt formations. Numerical code verification and validation is an important part of the lead-up to field testing, allowing exploration of potential heater emplacement designs, monitoring locations, and perhaps most importantly the ability to predict heat and mass transfer around an evolving test. Such predictions are crucial for the design and location of sampling and monitoring that can be used to validate our understanding of a drift scale test that is likely to span several years.

Coupled Thermal-Hydrologic-Chemical Coupled Model for In-Drift Disposal Test

International Nuclear Information System (INIS)

Jordan, Amy B.; Zyvoloski, George Anthony; Weaver, Douglas James; Otto, Shawn; Stauffer, Philip H.

2016-01-01

The simulation work presented in this report supports DOE-NE Used Fuel Disposition Campaign (UFDC) goals related to the development of drift scale in-situ field testing of heat-generating nuclear waste (HGNW) in salt formations. Numerical code verification and validation is an important part of the lead-up to field testing, allowing exploration of potential heater emplacement designs, monitoring locations, and perhaps most importantly the ability to predict heat and mass transfer around an evolving test. Such predictions are crucial for the design and location of sampling and monitoring that can be used to validate our understanding of a drift scale test that is likely to span several years.

Thermal impact of waste emplacement and surface cooling associated with geologic disposal of nuclear waste

Energy Technology Data Exchange (ETDEWEB)

Wang, J.S.Y.; Mangold, D.C.; Spencer, R.K.; Tsang, C.F.

1982-08-01

The thermal effects associated with the emplacement of aged radioactive wastes in a geologic repository were studied, with emphasis on the following subjects: the waste characteristics, repository structure, and rock properties controlling the thermally induced effects; the current knowledge of the thermal, thermomechanical, and thermohydrologic impacts, determined mainly on the basis of previous studies that assume 10-year-old wastes; the thermal criteria used to determine the repository waste loading densities; and the technical advantages and disadvantages of surface cooling of the wastes prior to disposal as a means of mitigating the thermal impacts. The waste loading densities determined by repository designs for 10-year-old wastes are extended to older wastes using the near-field thermomechanical criteria based on room stability considerations. Also discussed are the effects of long surface cooling periods determined on the basis of far-field thermomechanical and thermohydrologic considerations. The extension of the surface cooling period from 10 years to longer periods can lower the near-field thermal impact but have only modest long-term effects for spent fuel. More significant long-term effects can be achieved by surface cooling of reprocessed high-level waste.

Thermal impact of waste emplacement and surface cooling associated with geologic disposal of nuclear waste

International Nuclear Information System (INIS)

Wang, J.S.Y.; Mangold, D.C.; Spencer, R.K.; Tsang, C.F.

1982-08-01

The thermal effects associated with the emplacement of aged radioactive wastes in a geologic repository were studied, with emphasis on the following subjects: the waste characteristics, repository structure, and rock properties controlling the thermally induced effects; the current knowledge of the thermal, thermomechanical, and thermohydrologic impacts, determined mainly on the basis of previous studies that assume 10-year-old wastes; the thermal criteria used to determine the repository waste loading densities; and the technical advantages and disadvantages of surface cooling of the wastes prior to disposal as a means of mitigating the thermal impacts. The waste loading densities determined by repository designs for 10-year-old wastes are extended to older wastes using the near-field thermomechanical criteria based on room stability considerations. Also discussed are the effects of long surface cooling periods determined on the basis of far-field thermomechanical and thermohydrologic considerations. The extension of the surface cooling period from 10 years to longer periods can lower the near-field thermal impact but have only modest long-term effects for spent fuel. More significant long-term effects can be achieved by surface cooling of reprocessed high-level waste

Handling and Emplacement Options for Deep Borehole Disposal Conceptual Design.

Energy Technology Data Exchange (ETDEWEB)

Cochran, John R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Hardin, Ernest [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2015-07-01

This report presents conceptual design information for a system to handle and emplace packages containing radioactive waste, in boreholes 16,400 ft deep or possibly deeper. Its intended use is for a design selection study that compares the costs and risks associated with two emplacement methods: drill-string and wireline emplacement. The deep borehole disposal (DBD) concept calls for siting a borehole (or array of boreholes) that penetrate crystalline basement rock to a depth below surface of about 16,400 ft (5 km). Waste packages would be emplaced in the lower 6,560 ft (2 km) of the borehole, with sealing of appropriate portions of the upper 9,840 ft (3 km). A deep borehole field test (DBFT) is planned to test and refine the DBD concept. The DBFT is a scientific and engineering experiment, conducted at full-scale, in-situ, without radioactive waste. Waste handling operations are conceptualized to begin with the onsite receipt of a purpose-built Type B shipping cask, that contains a waste package. Emplacement operations begin when the cask is upended over the borehole, locked to a receiving flange or collar. The scope of emplacement includes activities to lower waste packages to total depth, and to retrieve them back to the surface when necessary for any reason. This report describes three concepts for the handling and emplacement of the waste packages: 1) a concept proposed by Woodward-Clyde Consultants in 1983; 2) an updated version of the 1983 concept developed for the DBFT; and 3) a new concept in which individual waste packages would be lowered to depth using a wireline. The systems described here could be adapted to different waste forms, but for design of waste packaging, handling, and emplacement systems the reference waste forms are DOE-owned high- level waste including Cs/Sr capsules and bulk granular HLW from fuel processing. Handling and Emplacement Options for Deep Borehole Disposal Conceptual Design July 23, 2015 iv ACKNOWLEDGEMENTS This report has

FIELD INVESTIGATION OF THE DRIFT SHADOW

International Nuclear Information System (INIS)

G.W. Su; T.J. Kneafsey

2006-01-01

A drift shadow is an area immediately beneath an underground void that, in theory, will be relatively drier than the surrounding rock mass. Numerical and analytical models of water flow through unsaturated rock predict the existence of a drift shadow, but field tests confirming the existence of the drift shadow have yet to be performed. Proving the existence of drift shadows and understanding their hydrologic and transport characteristics could provide a better understanding of how contaminants move in the subsurface if released from waste emplacement drifts such as the proposed nuclear waste repository at Yucca Mountain, Nevada. We describe the field program that will be used to investigate the existence of a drift shadow--and the corresponding hydrological process at the Hazel-Atlas silica-sand mine located at the Black Diamond Mines Regional Preserve in Antioch, California. The location and configuration of this mine makes it an excellent site to observe and measure drift shadow characteristics. The mine is located in a porous sandstone unit of the Domengine formation, an approximately 230 meter thick series of interbedded Eocene-age shales, coals, and massive-bedded sandstones. The mining method used at the mine required the development of two parallel drifts, one above the other, driven along the strike of the mined sandstone stratum. This configuration provides the opportunity to introduce water into the rock mass in the upper drift and to observe and measure its flow around the underlying drift. The passive and active hydrologic tests to be performed are described. In the passive method, cores will be obtained in a radial pattern around a drift and will be sectioned and analyzed for in-situ water content using a gravimetric technique, as well as analyzed for chemistry. With the active hydrologic test, water will be introduced into the upper drift of the two parallel drifts and the flow of the water will be tracked as it passes near the bottom drift

The Swedish Concept for Disposal of Spent Nuclear Fuel: Differences Between Vertical and Horizontal Waste Canister Emplacement

International Nuclear Information System (INIS)

Bennett, D.G.; Hicks, T.W.

2005-10-01

The Swedish Nuclear Power Inspectorate (SKI) is preparing for the review of licence applications related to the disposal of spent nuclear fuel. The Swedish Nuclear Fuel and Waste Management Company (SKB) refers to its proposals for the disposal of spent nuclear fuel as the KBS-3 concept. In the KBS-3 concept, SKB plans that, after 30 to 40 years of interim storage, spent fuel will be disposed of at a depth of about 500 m in crystalline bedrock, surrounded by a system of engineered barriers. The principle barrier to radionuclide release is a cylindrical copper canister. Within the copper canister, the spent fuel is supported by a cast iron insert. Outside the copper canister is a layer of bentonite clay, known as the buffer, which is designed to provide mechanical protection for the canisters and to limit the access of groundwater and corrosive substances to their surfaces. The bentonite buffer is also designed to sorb radionuclides released from the canisters, and to filter any colloids that may form within the waste. SKB is expected to base its forthcoming licence applications on a repository design in which the waste canisters are emplaced in vertical boreholes (KBS-3V). However, SKB has also indicated that it might be possible and, in some respects, beneficial to dispose of the waste canisters in horizontal tunnels (KBS-3H). There are many similarities between the KBS-3V and KBS-3H designs. There are, however, uncertainties associated with both of the designs and, when compared, both possess relative advantages and disadvantages. SKB has identified many of the key factors that will determine the evolution of a KBS-3H repository and has plans for research and development work in many of the areas where the differences between the KBS-3V and KBS-3H designs mean that they could be significant in terms of repository performance. With respect to the KBS-3H design, key technical issues are associated with: 1. The accuracy of deposition drift construction. 2. Water

Progress in waste package and engineered barrier system performance assessment and design

International Nuclear Information System (INIS)

Van Luik, A.; Stahl, D.; Harrison, D.

1993-01-01

As part of the U.S. Department of Energy's evaluation of site suitability for a potential high-level radioactive waste repository, long-term interactions between the engineered barrier system and the site must be determined. This requires a waste-package/engineered-system design, a description of the environment around the emplacement zone, and models that simulate operative processes describing these engineered/natural systems interactions. Candidate designs are being evaluated, including a more robust, multi-barrier waste package, and a drift emplacement mode. Tools for evaluating designs, and emplacement mode are the currently available waste-package/engineered-system performance assessment codes development for the project. For assessments that support site suitability, environmental impact, or licensing decisions, more capable codes are needed. Code capability requirements are being written, and existing codes are to be evaluated against those requirements. Recommendations are being made to focus waste-packaging/engineered-system code-development

Double Diffusive Natural Convection in a Nuclear Waste Repository

International Nuclear Information System (INIS)

Y. Hao; J. Nitao; T.A. Buscheck; Y. Sun

2006-01-01

In this study, we conduct a two-dimensional numerical analysis of double diffusive natural convection in an emplacement drift for a nuclear waste repository. In-drift heat and moisture transport is driven by combined thermal- and compositional-induced buoyancy forces. Numerical results demonstrate buoyancy-driven convective flow patterns and configurations during both repository heat-up and cool-down phases. It is also shown that boundary conditions, particularly on the drip-shield surface, have strong impacts on the in-drift convective flow and transport

Effective Thermal Conductivity For Drift-Scale Models Used In TSPA-SR

Energy Technology Data Exchange (ETDEWEB)

N.D. Francis

2001-01-25

The objective of this calculation is to develop a time dependent in-drift effective thermal conductivity parameter that will approximate heat conduction, thermal radiation, and natural convection heat transfer using a single mode of heat transfer (heat conduction). In order to reduce the physical and numerical complexity of the heat transfer processes that occur (and must be modeled) as a result of the emplacement of heat generating wastes, a single parameter will be developed that approximates all forms of heat transfer from the waste package surface to the drift wall (or from one surface exchanging heat with another). Subsequently, with this single parameter, one heat transfer mechanism (e.g., conduction heat transfer) can be used in the models. The resulting parameter is to be used as input in the drift-scale process-level models applied in total system performance assessments for the site recommendation (TSPA-SR). The format of this parameter will be a time-dependent table for direct input into the thermal-hydrologic (TH) and the thermal-hydrologic-chemical (THC) models.

Effective Thermal Conductivity For Drift-Scale Models Used In TSPA-SR

International Nuclear Information System (INIS)

N.D. Francis

2001-01-01

The objective of this calculation is to develop a time dependent in-drift effective thermal conductivity parameter that will approximate heat conduction, thermal radiation, and natural convection heat transfer using a single mode of heat transfer (heat conduction). In order to reduce the physical and numerical complexity of the heat transfer processes that occur (and must be modeled) as a result of the emplacement of heat generating wastes, a single parameter will be developed that approximates all forms of heat transfer from the waste package surface to the drift wall (or from one surface exchanging heat with another). Subsequently, with this single parameter, one heat transfer mechanism (e.g., conduction heat transfer) can be used in the models. The resulting parameter is to be used as input in the drift-scale process-level models applied in total system performance assessments for the site recommendation (TSPA-SR). The format of this parameter will be a time-dependent table for direct input into the thermal-hydrologic (TH) and the thermal-hydrologic-chemical (THC) models

Comparison of CFD Natural Convection and Conduction-only Models for Heat Transfer in the Yucca Mountain Project Drifts

International Nuclear Information System (INIS)

Hadgu, T.; Webb, S.; Itamura, M.

2004-01-01

Yucca Mountain, Nevada has been designated as the nation's high-level radioactive waste repository and the U.S. Department of Energy has been approved to apply to the U.S. Nuclear Regulatory Commission for a license to construct a repository. Heat transfer in the Yucca Mountain Project (YMP) drift enclosures is an important aspect of repository waste emplacement. Canisters containing radioactive waste are to be emplaced in tunnels drilled 500 m below the ground surface. After repository closure, decaying heat is transferred from waste packages to the host rock by a combination of thermal radiation, natural convection and conduction heat transfer mechanism?. Current YMP mountain-scale and drift-scale numerical models often use a simplified porous medium code to model fluid and heat flow in the drift openings. To account for natural convection heat transfer, the thermal conductivity of the air was increased in the porous medium model. The equivalent thermal conductivity, defined as the ratio of total heat flow to conductive heat flow, used in the porous media models was based on horizontal concentric cylinders. Such modeling does not effectively capture turbulent natural convection in the open spaces as discussed by Webb et al. (2003) yet the approach is still widely used on the YMP project. In order to mechanistically model natural convection conditions in YMP drifts, the computational fluid dynamics (CFD) code FLUENT (Fluent, Incorporated, 2001) has been used to model natural convection heat transfer in the YMP emplacement drifts. A two-dimensional (2D) model representative of YMP geometry (e.g., includes waste package, drip shield, invert and drift wall) has been developed and numerical simulations made (Francis et al., 2003). Using CFD simulation results for both natural convection and conduction-only heat transfer in a single phase, single component fluid, equivalent thermal conductivities have been calculated for different Rayleigh numbers. Correlation

Effects of the deviation characteristics of nuclear waste emplacement boreholes on borehole liner stresses; Yucca Mountain Project

Energy Technology Data Exchange (ETDEWEB)

Glowka, D.A.

1990-09-01

This report investigates the effects of borehole deviation on the useability of lined boreholes for the disposal of high-level nuclear waste at the proposed Yucca Mountain Repository in Nevada. Items that lead to constraints on borehole deviation include excessive stresses that could cause liner failure and possible binding of a waste container inside the liner during waste emplacement and retrieval operations. Liner stress models are developed for two general borehole configurations, one for boreholes drilled with a steerable bit and one for boreholes drilled with a non-steerable bit. Procedures are developed for calculating liner stresses that arise both during insertion of the liner into a borehole and during the thermal expansion process that follows waste emplacement. The effects of borehole curvature on the ability of the waste container to pass freely inside the liner without binding are also examined. Based on the results, specifications on borehole deviation allowances are developed for specific vertical and horizontal borehole configurations of current interest. 11 refs., 22 figs., 4 tabs.

The offshore disposal of radioactive waste by drilled emplacement: A feasibility study

International Nuclear Information System (INIS)

Bury, M.R.C.

1985-01-01

This book is a report, based on a study by Taylor Woodrow Construction Limited, on the overall feasibility of the disposal of high-level radioactive waste in boreholes drilled deep into the ocean bed. The work comprises an engineering appraisal of the disposal process with a view to establishing technical and operational feasibility and providing overall cost information to enable an economic assessment to be made. Contents: Summary report; Reference criteria; Drilling operation; Transfer of radioactive waste, personnel and other supplies; Handling of radioactive waste on board; Lowering strings of canisters; Emplacement and backfilling of canisters; Preliminary design of marine platform; Retrieval of flasks or canisters lost or misplaced; Variations to the features of the lowering system; Logistics of the operation; Construction cost estimate; Operational costs; Appendix

Natural convection in tunnels at Yucca Mountain and impact on drift seepage

Energy Technology Data Exchange (ETDEWEB)

Halecky, N.; Birkholzer, J.T.; Peterson, P.

2010-04-15

The decay heat from radioactive waste that is to be disposed in the once proposed geologic repository at Yucca Mountain (YM) will significantly influence the moisture conditions in the fractured rock near emplacement tunnels (drifts). Additionally, large-scale convective cells will form in the open-air drifts and will serve as an important mechanism for the transport of vaporized pore water from the fractured rock in the drift center to the drift end. Such convective processes would also impact drift seepage, as evaporation could reduce the build up of liquid water at the tunnel wall. Characterizing and understanding these liquid water and vapor transport processes is critical for evaluating the performance of the repository, in terms of water-induced canister corrosion and subsequent radionuclide containment. To study such processes, we previously developed and applied an enhanced version of TOUGH2 that solves for natural convection in the drift. We then used the results from this previous study as a time-dependent boundary condition in a high-resolution seepage model, allowing for a computationally efficient means for simulating these processes. The results from the seepage model show that cases with strong natural convection effects are expected to improve the performance of the repository, since smaller relative humidity values, with reduced local seepage, form a more desirable waste package environment.

Evaluation of Potential Impacts of Microbial Activity on Drift Chemistry

Energy Technology Data Exchange (ETDEWEB)

Y. Wang

2004-11-18

''Evaluation of Potential Impacts of Microbial Activity on Drift Chemistry'' focuses on the potential for microbial communities that could be active in repository emplacement drifts to influence the in-drift bulk chemical environment. This report feeds analyses to support the inclusion or exclusion of features, events, and processes (FEPs) in the total system performance assessment (TSPA) for the license application (LA), but this work is not expected to generate direct feeds to the TSPA-LA. The purpose was specified by, and the evaluation was performed and is documented in accordance with, ''Technical Work Plan For: Near-Field Environment and Transport In-Drift Geochemistry Analyses'' (BSC 2004 [DIRS 172402], Section 2.1). This report addresses all of the FEPs assigned by the technical work plan (TWP), including the development of exclusion arguments for FEPs that are not carried forward to the TSPA-LA. Except for an editorial correction noted in Section 6.2, there were no other deviations from the TWP. This report documents the completion of all assigned tasks, as follows (BSC 2004 DIRS 172402, Section 1.2.1): (1) Perform analyses to evaluate the potential for microbial activity in the waste emplacement drift under the constraints of anticipated physical and chemical conditions. (2) Evaluate uncertainties associated with these analyses. (3) Determine whether the potential for microbes warrants a feed to TSPA-LA to account for predicted effects on repository performance. (4) Provide information to address the ''Yucca Mountain Review Plan, Final Report'' (NUREG-1804) (NRC 2003 [DIRS 163274]) and Key Technical Issues and agreements, as appropriate. (5) Develop information for inclusion or exclusion of FEPs.

Evaluation of Potential Impacts of Microbial Activity on Drift Chemistry

International Nuclear Information System (INIS)

Y. Wang

2004-01-01

''Evaluation of Potential Impacts of Microbial Activity on Drift Chemistry'' focuses on the potential for microbial communities that could be active in repository emplacement drifts to influence the in-drift bulk chemical environment. This report feeds analyses to support the inclusion or exclusion of features, events, and processes (FEPs) in the total system performance assessment (TSPA) for the license application (LA), but this work is not expected to generate direct feeds to the TSPA-LA. The purpose was specified by, and the evaluation was performed and is documented in accordance with, ''Technical Work Plan For: Near-Field Environment and Transport In-Drift Geochemistry Analyses'' (BSC 2004 [DIRS 172402], Section 2.1). This report addresses all of the FEPs assigned by the technical work plan (TWP), including the development of exclusion arguments for FEPs that are not carried forward to the TSPA-LA. Except for an editorial correction noted in Section 6.2, there were no other deviations from the TWP. This report documents the completion of all assigned tasks, as follows (BSC 2004 DIRS 172402, Section 1.2.1): (1) Perform analyses to evaluate the potential for microbial activity in the waste emplacement drift under the constraints of anticipated physical and chemical conditions. (2) Evaluate uncertainties associated with these analyses. (3) Determine whether the potential for microbes warrants a feed to TSPA-LA to account for predicted effects on repository performance. (4) Provide information to address the ''Yucca Mountain Review Plan, Final Report'' (NUREG-1804) (NRC 2003 [DIRS 163274]) and Key Technical Issues and agreements, as appropriate. (5) Develop information for inclusion or exclusion of FEPs

Concepts and Technologies for Radioactive Waste Disposal in Rock Salt

Directory of Open Access Journals (Sweden)

Wernt Brewitz

2007-01-01

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

Comparative simulation study of coupled THM processes near back-filled and open-drift nuclear waste repositories in Task D of the International DECOVALEX Project

International Nuclear Information System (INIS)

Rutqvist, J.; Birkholzer, J.T.; Chijimatsu, M.; Kolditz, O.; Liu, Quan-Sheng; Oda, Y.; Wang, Wenqing; Zhang, Cheng-Yuan

2006-01-01

As part of the ongoing international DECOVALEX project, four research teams used five different models to simulate coupled thermal, hydrological, and mechanical (THM) processes near underground waste emplacement drifts. The simulations were conducted for two generic repository types, one with open and the other with back-filled repository drifts, under higher and lower post-closure temperature, respectively. In the completed first model inception phase of the project, a good agreement was achieved between the research teams in calculating THM responses for both repository types, although some disagreement in hydrological responses are currently being resolved. Good agreement in the basic thermal-mechanical responses was also achieved for both repository types, even though some teams used relatively simplified thermal-elastic heat-conduction models that neglect complex near-field thermal-hydrological processes. The good agreement between the complex and simplified process models indicates that the basic thermal-mechanical responses can be predicted with a relatively high confidence level

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

International Nuclear Information System (INIS)

1987-12-01

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

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

Energy Technology Data Exchange (ETDEWEB)

1987-12-01

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

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

International Nuclear Information System (INIS)

1987-12-01

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

Operational procedures for receiving, packaging, emplacing, and retrieving high-level and transuranic waste in a geologic repository in TUFF

International Nuclear Information System (INIS)

Dennis, A.W.; Mulkin, R.

1984-01-01

The Nevada Nuclear Waste Storage Investigations Project, directed by the Nevada Operations Office of the Department of Energy, is currently developing conceptual designs for a commercial nuclear waste repository. In this paper, the preliminary repository operating plans are identified and the proposed repository waste inventory is discussed. The receipt rates for truck and rail car shipments of waste are determined as are the required repository waste emplacement rates

CLASSIFICATION OF THE MGR WASTE EMPLACEMENT/RETRIEVAL SYSTEM

International Nuclear Information System (INIS)

J.A. Ziegler

2000-01-01

The purpose of this analysis is to document the Quality Assurance (QA) classification of the Monitored Geologic Repository (MGR) waste emplacement/retrieved system structures, systems and components (SSCs) performed by the MGR Preclosure Safety and Systems Engineering Section. This analysis also provides the basis for revision of YMP/90-55Q, Q-List (YMP 2000). The Q-List identifies those MGR SSCs subject to the requirements of DOE/RW-0333P, Quality Assurance Requirements and Description (QARD) (DOE 2000). This QA classification incorporates the current MGR design and the results of the ''Design Basis Event Frequency and Dose Calculation for Site Recommendation'' (CRWMS M andO 2000a). The content and technical approach of this analysis is in accordance with the development plan ''QA Classification of MGR Structures, Systems, and Components'' (CRWMS M andO 1999b)

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

Drift-scale thermo-hydrologic analyses for the proposed repository at Yucca Mountain, NV

Energy Technology Data Exchange (ETDEWEB)

Reeves, M.; Lingineni, S. [M& O/Intera, Inc., Las Vegas, NV (United States)

1995-03-01

Focusing on the near field of a waste package, this paper presents results obtained from a thermo-hydrologic model. Vertically, the conceptual model extends from ground surface to a depth of 1000 m below the water table. Horizontally, it is bounded by the centerlines of two pillars. Inside an assumed circular drift, an invert (backfill) surrounds the waste package. Here, a coarse-grained crushed tuff is assumed to provide a capillary barrier against infiltrating ground water. For comparison, the invert is replaced by intact tuff, corresponding to borehole emplacement. In further examining the near-field thermo-hydrology of a waste package, the study considers several values of the areal power density together with both nonenhanced and enhanced levels of vapor diffusivity. In order to characterize drift resaturation effects deriving from the latter, the calculations extend to 100,000 years. The clarity of the coding present in LBL`s integrated finite-difference model TOUGH2 has permitted us to make several code changes. For the most part, they expedite convergence on difficult applications. The paper briefly discusses these changes.

Radioactive waste isolation in salt: Peer review of the Fluor Technology, Inc., report and position paper concerning waste emplacement mode and its effect on repository conceptual design

International Nuclear Information System (INIS)

Hambley, D.F.; Russell, J.E.; Whitfield, R.G.

1987-02-01

Recommendations for revising the Fluor Technology, Inc., draft position paper entitled Evaluation of Waste Emplacement Mode and the final report entitled Waste Package/Repository Impact Study include: reevaluate the relative rankings for the various emplacement modes; delete the following want objectives: maximize ability to locate the package horizon because sufficient flexibility exists to locate rooms in the relatively clean San Andres Unit 4 Salt and maximize far-field geologic integrity during retrieval because by definition the far field will be unaffected by thermal and stress perturbations caused by remining; give greater emphasis to want objectives regarding cost and use of present technology; delete the following statements from pages 1-1 and 1-2 of the draft position paper: ''No thought or study was given to the impacts of this configuration [vertical emplacement] on repository construction or short and long-term performance of the site'' and ''Subsequent salt repository designs adopted the vertical emplacement configuration as the accepted method without further evaluation.''; delete App. E and lines 8-17 of page 1-4 of the draft position paper because they are inappropriate; adopt a formal decision-analysis procedure for the 17 identified emplacement modes; revise App. F of the impact study to more accurately reflect current technology; consider designing the underground layout to take advantage of stress-relief techniques; consider eliminating reference to fuel assemblies <10 yr ''out-of-reactor''; model the temperature distribution, assuming that the repository is constructed in an infinitely large salt body; state that the results of creep analyses must be considered tentative until they can be validated by in situ measurements; and reevaluate the peak radial stresses on the waste package so that the calculated stress conditions more closely approximate expected in situ conditions

A feasibility study of the disposal of radioactive waste in deep ocean sediments by drilled emplacement

International Nuclear Information System (INIS)

Bury, M.R.C.

1983-08-01

This report describes the second phase of a study of the feasibility of disposal and isolation of high level radioactive waste in holes drilled deep into the sediments of the ocean. In this phase, work has concentrated on establishing the state of the art of the various operations and developing the design, in particular the drilling operation, the loading of flasks containing waste canisters from supply vessels onto the platform, the handling of radioactive waste on board, and its emplacement into predrilled holes. In addition, an outline design of the offshore platform has been prepared. (author)

License Application Design Selection Feature Report: Waste Package Self Shielding Design Feature 13

International Nuclear Information System (INIS)

Tang, J.S.

2000-01-01

In the Viability Assessment (VA) reference design, handling of waste packages (WPs) in the emplacement drifts is performed remotely, and human access to the drifts is precluded when WPs are present. This report will investigate the feasibility of using a self-shielded WP design to reduce the radiation levels in the emplacement drifts to a point that, when coupled with ventilation, will create an acceptable environment for human access. This provides the benefit of allowing human entry to emplacement drifts to perform maintenance on ground support and instrumentation, and carry out performance confirmation activities. More direct human control of WP handling and emplacement operations would also be possible. However, these potential benefits must be weighed against the cost of implementation, and potential impacts on pre- and post-closure performance of the repository and WPs. The first section of this report will provide background information on previous investigations of the self-shielded WP design feature, summarize the objective and scope of this document, and provide quality assurance and software information. A shielding performance and cost study that includes several candidate shield materials will then be performed in the subsequent section to allow selection of two self-shielded WP design options for further evaluation. Finally, the remaining sections will evaluate the impacts of the two WP self-shielding options on the repository design, operations, safety, cost, and long-term performance of the WPs with respect to the VA reference design

Drift Degradation Analysis

International Nuclear Information System (INIS)

D. Kicker

2004-01-01

Degradation of underground openings as a function of time is a natural and expected occurrence for any subsurface excavation. Over time, changes occur to both the stress condition and the strength of the rock mass due to several interacting factors. Once the factors contributing to degradation are characterized, the effects of drift degradation can typically be mitigated through appropriate design and maintenance of the ground support system. However, for the emplacement drifts of the geologic repository at Yucca Mountain, it is necessary to characterize drift degradation over a 10,000-year period, which is well beyond the functional period of the ground support system. This document provides an analysis of the amount of drift degradation anticipated in repository emplacement drifts for discrete events and time increments extending throughout the 10,000-year regulatory period for postclosure performance. This revision of the drift degradation analysis was developed to support the license application and fulfill specific agreement items between the U.S. Nuclear Regulatory Commission (NRC) and the U.S. Department of Energy (DOE). The earlier versions of ''Drift Degradation Analysis'' (BSC 2001 [DIRS 156304]) relied primarily on the DRKBA numerical code, which provides for a probabilistic key-block assessment based on realistic fracture patterns determined from field mapping in the Exploratory Studies Facility (ESF) at Yucca Mountain. A key block is defined as a critical block in the surrounding rock mass of an excavation, which is removable and oriented in an unsafe manner such that it is likely to move into an opening unless support is provided. However, the use of the DRKBA code to determine potential rockfall data at the repository horizon during the postclosure period has several limitations: (1) The DRKBA code cannot explicitly apply dynamic loads due to seismic ground motion. (2) The DRKBA code cannot explicitly apply loads due to thermal stress. (3) The DRKBA

Drift Degradation Analysis

Energy Technology Data Exchange (ETDEWEB)

D. Kicker

2004-09-16

Degradation of underground openings as a function of time is a natural and expected occurrence for any subsurface excavation. Over time, changes occur to both the stress condition and the strength of the rock mass due to several interacting factors. Once the factors contributing to degradation are characterized, the effects of drift degradation can typically be mitigated through appropriate design and maintenance of the ground support system. However, for the emplacement drifts of the geologic repository at Yucca Mountain, it is necessary to characterize drift degradation over a 10,000-year period, which is well beyond the functional period of the ground support system. This document provides an analysis of the amount of drift degradation anticipated in repository emplacement drifts for discrete events and time increments extending throughout the 10,000-year regulatory period for postclosure performance. This revision of the drift degradation analysis was developed to support the license application and fulfill specific agreement items between the U.S. Nuclear Regulatory Commission (NRC) and the U.S. Department of Energy (DOE). The earlier versions of ''Drift Degradation Analysis'' (BSC 2001 [DIRS 156304]) relied primarily on the DRKBA numerical code, which provides for a probabilistic key-block assessment based on realistic fracture patterns determined from field mapping in the Exploratory Studies Facility (ESF) at Yucca Mountain. A key block is defined as a critical block in the surrounding rock mass of an excavation, which is removable and oriented in an unsafe manner such that it is likely to move into an opening unless support is provided. However, the use of the DRKBA code to determine potential rockfall data at the repository horizon during the postclosure period has several limitations: (1) The DRKBA code cannot explicitly apply dynamic loads due to seismic ground motion. (2) The DRKBA code cannot explicitly apply loads due to thermal

Emplacement and retrieval equipment design considerations for a repository in salt

International Nuclear Information System (INIS)

Nair, B.R.; Bahorich, R.J.

1987-01-01

The current design concept for the disposal of nuclear high level waste packages in a repository in salt is based on the emplacement of individual packages in vertical boreholes in the underground mine floor. A key requirement is that the waste packages be capable of being retrieved during the last 26 years of the 76-year repository operating period. The unique design considerations relating to the retrieval of waste packages emplaced in bedded salt are presented in this paper. The information is based on the experience developed during the design of vertical emplacement and retrieval equipment in support of the Sandia Defense High Level Waste experiments at the Waste Isolation Pilot Plant. Also included are the impact of retrievability on the design of the equipment, the special salt cutting technology that was developed for this application, and a description of the equipment

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

Pre-waste-emplacement ground-water travel time sensitivity and uncertainty analyses for Yucca Mountain, Nevada

International Nuclear Information System (INIS)

Kaplan, P.G.

1993-01-01

Yucca Mountain, Nevada is a potential site for a high-level radioactive-waste repository. Uncertainty and sensitivity analyses were performed to estimate critical factors in the performance of the site with respect to a criterion in terms of pre-waste-emplacement ground-water travel time. The degree of failure in the analytical model to meet the criterion is sensitive to the estimate of fracture porosity in the upper welded unit of the problem domain. Fracture porosity is derived from a number of more fundamental measurements including fracture frequency, fracture orientation, and the moisture-retention characteristic inferred for the fracture domain

New three-dimensional far-field potential repository thermomechanical calculations

International Nuclear Information System (INIS)

Hardy, M.P.; Bai, M.; Goodrich, R.R.; Lin, M.; Carlisle, S.; Bauer, S.J.

1993-03-01

The thermomechanical effect on the exploratory ramps, drifts, and shafts as a result of high-level nuclear waste disposal is examined using a three-dimensional thermo-elastic model. The repository layout modeled is based on the use of mechanical mining of all excavations with equivalent waste emplacement areal power densities of 57 and 80 kW/acre. Predicted temperatures and stress changes for the north and south access drifts, east main drift, east-west exploratory drift, the north and south Calico Hills access ramps, the Calico Hills north-south exploratory drift, and the optional exploratory studies facility and man and materials shafts are presented for times 10, 35, 50, 100, 300, 500, 1000, 2000, 5000, and 10,000 years after the start of waste emplacement. The study indicates that the east-west exploratory drift at the repository horizon is subject to the highest thermomechanical impact because it is located closest the buried waste canisters. For most exploratory openings, the thermally induced temperatures and stresses tend to reach the maximum magnitudes at approximately 1000 years after waste emplacement

Drift-Scale Coupled Processes (DST and THC Seepage) Models

Energy Technology Data Exchange (ETDEWEB)

E. Gonnenthal; N. Spyoher

2001-02-05

The purpose of this Analysis/Model Report (AMR) is to document the Near-Field Environment (NFE) and Unsaturated Zone (UZ) models used to evaluate the potential effects of coupled thermal-hydrologic-chemical (THC) processes on unsaturated zone flow and transport. This is in accordance with the ''Technical Work Plan (TWP) for Unsaturated Zone Flow and Transport Process Model Report'', Addendum D, Attachment D-4 (Civilian Radioactive Waste Management System (CRWMS) Management and Operating Contractor (M and O) 2000 [153447]) and ''Technical Work Plan for Nearfield Environment Thermal Analyses and Testing'' (CRWMS M and O 2000 [153309]). These models include the Drift Scale Test (DST) THC Model and several THC seepage models. These models provide the framework to evaluate THC coupled processes at the drift scale, predict flow and transport behavior for specified thermal loading conditions, and predict the chemistry of waters and gases entering potential waste-emplacement drifts. The intended use of this AMR is to provide input for the following: (1) Performance Assessment (PA); (2) Abstraction of Drift-Scale Coupled Processes AMR (ANL-NBS-HS-000029); (3) UZ Flow and Transport Process Model Report (PMR); and (4) Near-Field Environment (NFE) PMR. The work scope for this activity is presented in the TWPs cited above, and summarized as follows: continue development of the repository drift-scale THC seepage model used in support of the TSPA in-drift geochemical model; incorporate heterogeneous fracture property realizations; study sensitivity of results to changes in input data and mineral assemblage; validate the DST model by comparison with field data; perform simulations to predict mineral dissolution and precipitation and their effects on fracture properties and chemistry of water (but not flow rates) that may seep into drifts; submit modeling results to the TDMS and document the models. The model development, input data

Drift-Scale Coupled Processes (DST and THC Seepage) Models

International Nuclear Information System (INIS)

Sonnenthale, E.

2001-01-01

The purpose of this Analysis/Model Report (AMR) is to document the Near-Field Environment (NFE) and Unsaturated Zone (UZ) models used to evaluate the potential effects of coupled thermal-hydrologic-chemical (THC) processes on unsaturated zone flow and transport. This is in accordance with the ''Technical Work Plan (TWP) for Unsaturated Zone Flow and Transport Process Model Report'', Addendum D, Attachment D-4 (Civilian Radioactive Waste Management System (CRWMS) Management and Operating Contractor (M and O) 2000 [1534471]) and ''Technical Work Plan for Nearfield Environment Thermal Analyses and Testing'' (CRWMS M and O 2000 [153309]). These models include the Drift Scale Test (DST) THC Model and several THC seepage models. These models provide the framework to evaluate THC coupled processes at the drift scale, predict flow and transport behavior for specified thermal loading conditions, and predict the chemistry of waters and gases entering potential waste-emplacement drifts. The intended use of this AMR is to provide input for the following: Performance Assessment (PA); Near-Field Environment (NFE) PMR; Abstraction of Drift-Scale Coupled Processes AMR (ANL-NBS-HS-000029); and UZ Flow and Transport Process Model Report (PMR). The work scope for this activity is presented in the TWPs cited above, and summarized as follows: Continue development of the repository drift-scale THC seepage model used in support of the TSPA in-drift geochemical model; incorporate heterogeneous fracture property realizations; study sensitivity of results to changes in input data and mineral assemblage; validate the DST model by comparison with field data; perform simulations to predict mineral dissolution and precipitation and their effects on fracture properties and chemistry of water (but not flow rates) that may seep into drifts; submit modeling results to the TDMS and document the models. The model development, input data, sensitivity and validation studies described in this AMR are

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

International Nuclear Information System (INIS)

1984-01-01

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

Dike/Drift Interactions

Energy Technology Data Exchange (ETDEWEB)

E.S. Gaffney

2003-10-08

This report documents the model of events associated with a potential intrusion of magma from a volcanic dike into a drift or drifts in the Yucca Mountain Nuclear Waste Repository. The following topics are included in this report: (1) A discussion of dike propagation, which provides the basis for describing the path that a representative dike, or swarm of dikes, would follow during an event. (2) A discussion of magma flow, which evaluates the interaction at the junction of the propagating dike with the drift and the movement of magmatic products into and down drifts and, potentially, through a drift to the surface by way of access drift or a secondary dike opened up along the drift. (3) A discussion of gas flow and conductive cooling of a magma-filled drift, describing how an adjacent drift that has not been intersected by a dike could be affected by post-intrusion phenomena. Note that a gas flow analysis is also addressed in ''Igneous Intrusion Impacts on Waste Form and Waste Packages'' (BSC 2003 [DIRS 161810]), and those results are consistent with the results presented in this report.

Dike/Drift Interactions

International Nuclear Information System (INIS)

E.S. Gaffney

2003-01-01

This report documents the model of events associated with a potential intrusion of magma from a volcanic dike into a drift or drifts in the Yucca Mountain Nuclear Waste Repository. The following topics are included in this report: (1) A discussion of dike propagation, which provides the basis for describing the path that a representative dike, or swarm of dikes, would follow during an event. (2) A discussion of magma flow, which evaluates the interaction at the junction of the propagating dike with the drift and the movement of magmatic products into and down drifts and, potentially, through a drift to the surface by way of access drift or a secondary dike opened up along the drift. (3) A discussion of gas flow and conductive cooling of a magma-filled drift, describing how an adjacent drift that has not been intersected by a dike could be affected by post-intrusion phenomena. Note that a gas flow analysis is also addressed in ''Igneous Intrusion Impacts on Waste Form and Waste Packages'' (BSC 2003 [DIRS 161810]), and those results are consistent with the results presented in this report

Review of potential subsurface permeable barrier emplacement and monitoring technologies

International Nuclear Information System (INIS)

Riggsbee, W.H.; Treat, R.L.; Stansfield, H.J.; Schwarz, R.M.; Cantrell, K.J.; Phillips, S.J.

1994-02-01

This report focuses on subsurface permeable barrier technologies potentially applicable to existing waste disposal sites. This report describes candidate subsurface permeable barriers, methods for emplacing these barriers, and methods used to monitor the barrier performance. Two types of subsurface barrier systems are described: those that apply to contamination.in the unsaturated zone, and those that apply to groundwater and to mobile contamination near the groundwater table. These barriers may be emplaced either horizontally or vertically depending on waste and site characteristics. Materials for creating permeable subsurface barriers are emplaced using one of three basic methods: injection, in situ mechanical mixing, or excavation-insertion. Injection is the emplacement of dissolved reagents or colloidal suspensions into the soil at elevated pressures. In situ mechanical mixing is the physical blending of the soil and the barrier material underground. Excavation-insertion is the removal of a soil volume and adding barrier materials to the space created. Major vertical barrier emplacement technologies include trenching-backfilling; slurry trenching; and vertical drilling and injection, including boring (earth augering), cable tool drilling, rotary drilling, sonic drilling, jetting methods, injection-mixing in drilled holes, and deep soil mixing. Major horizontal barrier emplacement technologies include horizontal drilling, microtunneling, compaction boring, horizontal emplacement, longwall mining, hydraulic fracturing, and jetting methods

Multiscale Model Simulations of Temperature and Relative Humidity for the License Application of the Proposed Yucca Mountain Repository

Science.gov (United States)

Buscheck, T.; Glascoe, L.; Sun, Y.; Gansemer, J.; Lee, K.

2003-12-01

For the proposed Yucca Mountain geologic repository for high-level nuclear waste, the planned method of disposal involves the emplacement of cylindrical packages containing the waste inside horizontal tunnels, called emplacement drifts, bored several hundred meters below the ground surface. The emplacement drifts reside in highly fractured, partially saturated volcanic tuff. An important phenomenological consideration for the licensing of the proposed repository at Yucca Mountain is the generation of decay heat by the emplaced waste and the consequences of this decay heat. Changes in temperature will affect the hydrologic and chemical environment at Yucca Mountain. A thermohydrologic-modeling tool is necessary to support the performance assessment of the Engineered Barrier System (EBS) of the proposed repository. This modeling tool must simultaneously account for processes occurring at a scale of a few tens of centimeters around individual waste packages, for processes occurring around the emplacement drifts themselves, and for processes occurring at the multi-kilometer scale of the mountain. Additionally, many other features must be considered including non-isothermal, multiphase-flow in fractured porous rock of variable liquid-phase saturation and thermal radiation and convection in open cavities. The Multiscale Thermohydrologic Model (MSTHM) calculates the following thermohydrologic (TH) variables: temperature, relative humidity, liquid-phase saturation, evaporation rate, air-mass fraction, gas-phase pressure, capillary pressure, and liquid- and gas-phase fluxes. The TH variables are determined as a function of position along each of the emplacement drifts in the repository and as a function of waste-package (WP) type. These variables are determined at various generic locations within the emplacement drifts, including the waste package and drip-shield surfaces and in the invert; they are also determined at various generic locations in the adjoining host rock

Engineering studies: high-level radioactive waste repositories task 3 - review of underground handling and emplacement. 1. Executive summary

Energy Technology Data Exchange (ETDEWEB)

1981-02-01

The report reviews proposals for transport, handling and emplacement of high-level radioactive waste in an underground repository with particular reference to: waste block size and configuration; self-shielded or partially-shielded block; stages of disposal; transport by road/rail to repository site; handling techniques within repository; emplacement in vertical holes or horizontal tunnels; repository access by adit, incline or shaft; conventional and radiological safety; costs; and major areas of uncertainty requiring research or development. In carrying out this programme due attention was given to work already carried out both in the U.K. and overseas and where appropriate comparisons with this work have been made to substantiate and explain the observations made in this report. The examination and use of this previous work however has been confined to those proposals which were considered capable of meeting the basic design criterion for a U.K. based repository, that the maximum temperature achieved by the host rock should not exceed 100/sup 0/C.

Drift design methodology and preliminary application for the Yucca Mountain Site Characterization Project

International Nuclear Information System (INIS)

Hardy, M.P.; Bauer, S.J.

1991-12-01

Excavation stability in an underground nuclear waste repository is required during construction, emplacement, retrieval (if required), and closure phases to ensure worker health and safety, and to prevent development of potential pathways for radionuclide migration in the post-closure period. Stable excavations are developed by appropriate excavation procedures, design of the room shape, design and installation of rock support reinforcement systems, and implementation of appropriate monitoring and maintenance programs. In addition to the loads imposed by the in situ stress field, the repository drifts will be impacted by thermal loads developed after waste emplacement and, periodically, by seismic loads from naturally occurring earthquakes and underground nuclear events. A priori evaluation of stability is required for design of the ground support system, to confirm that the thermal loads are reasonable, and to support the license application process. In this report, a design methodology for assessing drift stability is presented. This is based on site conditions, together with empirical and analytical methods. Analytical numerical methods are emphasized at this time because empirical data are unavailable for excavations in welded tuff either at elevated temperatures or under seismic loads. The analytical methodology incorporates analysis of rock masses that are systematically jointed, randomly jointed, and sparsely jointed. In situ thermal and seismic loads are considered. Methods of evaluating the analytical results and estimating ground support requirements for all the full range of expected ground conditions are outlines. The results of a preliminary application of the methodology using the limited available data are presented. 26 figs., 55 tabs

ERG [Engineering Review Group] and GRG [Geologic Review Group] review of the horizontal versus vertical modes of waste emplacement at the Deaf Smith County site, Texas

International Nuclear Information System (INIS)

Chytrowski, B.R.

1988-01-01

The Engineering Review Group (ERG) and Geologic Review Group (GRG) were established by the Office of Nuclear Waste Isolation (ONWI) to help evaluate specific issues in the US Department of Energy's nuclear waste repository program. The December 1985 meeting and the February 1986 meeting dealt with the evaluation of the Fluor Technology, Inc., architect-engineer recommendation of the horizontal mode of waste package emplacement for the Site Characterization Plan Conceptual Design Report (SCP-CDR). The ONWI recommendation regarding horizontal and vertical modes of waste package emplacement and associated studies was reviewed. This report documents the ERG and GRG's comments and recommendations on this subject and ONWI responses to the specific points raised by these groups. The ERG and GRG joint review groups concurred with ONWI recommendations that additional studies are required in order to reach a decision on the method of emplacement to be used. In the opinion of these groups, both methods can be implemented; however, should the decision be reached today the vertical mode would be preferred

Landsat investigations of the northern Paradox basin, Utah and Colorado: implications for radioactive waste emplacement

Science.gov (United States)

Friedman, Jules D.; Simpson, Shirley L.

1978-01-01

The first stages of a remote-sensing project on the Paradox basin, part of the USGS (U.S. Geological Survey) radioactive waste-emplacement program, consisted of a review and selection of the best available satellite scanner images to use in geomorphologic and tectonic investigations of the region. High-quality Landsat images in several spectral bands (E-2260-17124 and E-5165-17030), taken under low sun angle October 9 and 10, 1975, were processed via computer for planimetric rectification, histogram analysis, linear transformation of radiance values, and edge enhancement. A lineament map of the northern Paradox basin was subsequently compiled at 1:400,000 using the enhanced Landsat base. Numerous previously unmapped northeast-trending lineaments between the Green River and Yellowcat dome; confirmatory detail on the structural control of major segments of the Colorado, Gunnison, and Dolores Rivers; and new evidence for late Phanerozoic reactivation of Precambrian basement structures are among the new contributions to the tectonics of the region. Lineament trends appear to be compatible with the postulated Colorado lineament zone, with geophysical potential-field anomalies, and with a northeast-trending basement fault pattern. Combined Landsat, geologic, and geophysical field evidence for this interpretation includes the sinuousity of the composite Salt Valley anticline, the transection of the Moab-Spanish Valley anticline on its southeastern end by northeast-striking faults, and possible transection (?) of the Moab diapir. Similarly, northeast-trending lineaments in Cottonwood Canyon and elsewhere are interpreted as manifestations of structures associated with northeasterly trends in the magnetic and gravity fields of the La Sal Mountains region. Other long northwesterly lineaments near the western termination of the Ryan Creek fault zone. may be associated with the fault zone separating the Uncompahgre horst uplift from the Paradox basin. Implications of the

The Effect of Unheated Sections on Moisture Transport in the Emplacement Drift

International Nuclear Information System (INIS)

G. Danko; J. Birkholzer; D. Bahrami

2006-01-01

The goals of this study are: (1) to configure a thermal-hydrological, natural-ventilation model for simulating temperature, humidity, and condensate distributions in the coupled domains of in-drift airspace and near-field rockmass. Rokmass model: TOUGH2, in-drift model: MULTIFLUX (MF); (2) obtain meaningful results from the model for a practical application in which the beneficial effects of unheated drift sections are analyzed; and (3) study the sensitivity to the axial dispersion coefficient with the model

Enhanced Design Alternative IV

International Nuclear Information System (INIS)

Kramer, N.E.

1999-01-01

This report evaluates Enhanced Design Alternative (EDA) IV as part of the second phase of the License Application Design Selection (LADS) effort. The EDA IV concept was compared to the VA reference design using criteria from the Design Input Request for LADS Phase II EDA Evaluations (CRWMS M and O 1999b) and (CRWMS M and O 1999f). Briefly, the EDA IV concept arranges the waste packages close together in an emplacement configuration known as line load. Continuous pre-closure ventilation keeps the waste packages from exceeding their 350 C cladding and 200 C (4.3.6) drift wall temperature limits. This EDA concept keeps relatively high, uniform emplacement drift temperatures (post-closure) to drive water away from the repository and thus dry out the pillars between emplacement drifts. The waste package is shielded to permit human access to emplacement drifts and includes an integral filler inside the package to reduce the amount of water that can contact the waste form. Closure of the repository is desired 50 years after first waste is emplaced. Both backfill and drip shields will be emplaced at closure to improve post-closure performance. The EDA IV concept includes more defense-in-depth layers than the VA reference design because of its backfill, drip shield, waste package shielding, and integral filler features. These features contribute to the low dose-rate to the public achieved during the first 10,000 years of repository life as shown in Figure 3. Investigation of the EDA IV concept has led to the following general conclusions: (1) The total life cycle cost for EDA IV is about $21.7 billion which equates to a $11.3 billion net present value (both figures rounded up). (2) The incidence of design basis events for EDA IV is similar to the VA reference design. (3) The emplacement of the waste packages in drifts will be similar to the VA reference design. However, heavier equipment may be required because the shielded waste package will be heavier. (4) The heavier

Waste package/engineered barrier system design concepts for the direct disposal of spent fuel in the potential United States' repository at Yucca Mountain, Nevada

International Nuclear Information System (INIS)

Stahl, D.; Harrison, D.J.

1993-01-01

The goal of the US Department of Energy's (DOE) Yucca Mountain Site Characterization Project (YMP) waste package development program is to design a waste package and associated engineered barrier system (EBS) that meets the applicable regulatory requirements for safe disposal of spent nuclear fuel and solidified high-level waste (HLW) in a geologic repository. Attainment of this goal relies on a multi-barrier approach, the unsaturated nature of the Yucca Mountain site, consideration of technical alternatives, and sufficient resolution of technical and regulatory uncertainties. To accomplish this, an iterative system engineering approach will be used. The NWPA of 1982 limits the content of the first US repository to 70,000 metric tons of heavy metal (MTHM). The DOE Mission Plan describes the implementation of the provisions of the NWPA for the waste management system. The Draft 1988 approach will involve selecting candidate designs, evaluating them against performance requirements, and then selecting one or two preferred designs for further detailed evaluation and final design. The reference design of the waste package described in the YMP Site Characterization Plan is a thin-walled, vertical borehole-emplaced waste package with an air gap between the package and the rock wall. The reference design appeared to meet the design requirement. However, the degree of uncertainty was large. This uncertainty led to considering several more-robust design concepts during the Advanced Conceptual Design phase of the program that include small, drift-emplaced packages and higher capacity, drift-emplaced packages, both partially and totally self-shielded. Metallic as well as ceramic materials are being considered

Construction, emplacement, and retrievability (preclosure)

International Nuclear Information System (INIS)

McClain, W.

1985-01-01

Each of the three preclosure subgroups of the Construction, Emplacement, and Retrievability Working Group adopted a six-step approach to identify and assess current needs in geotechnical modeling and characterization. This approach may be summarized as follows: identify phenomena related to emplacement of high-level nuclear wastes, identify types of models which are required to calculate the phenomena, establish the input data needs for the models, assess the current availability of the models, assess the current status of documentation, verification, and validation of the models, and determine the adequacy of instrumentation and measurement techniques to (a) validate the models, where necessary, and (b) obtain input data for design. Systematic application of these six steps leads to the establishment of the research requirements for geotechnical modeling and characterization. A summary of modeling techniques which apply to the three subsequent sections on construction, emplacement, and retrievability is presented. Research needs, which apply to all preclosure activities, are summarized

Criticality Potential of Waste Packages Containing DOE SNF Affected by Igneous Intrusion

International Nuclear Information System (INIS)

D.S. Kimball; C.E. Sanders

2006-01-01

The Department of Energy (DOE) is currently preparing an application to submit to the U.S. Nuclear Regulatory Commission for a construction authorization for a monitored geologic repository. The repository will contain spent nuclear fuel (SNF) and defense high-level waste (DHLW) in waste packages placed in underground tunnels, or drifts. The primary objective of this paper is to perform a criticality analysis for waste packages containing DOE SNF affected by a disruptive igneous intrusion event in the emplacement drifts. The waste packages feature one DOE SNF canister placed in the center and surrounded by five High-Level Waste (HLW) glass canisters. The effective neutron multiplication factor (k eff ) is determined for potential configurations of the waste package during and after an intrusive igneous event. Due to the complexity of the potential scenarios following an igneous intrusion, finding conservative and bounding configurations with respect to criticality requires some additional considerations. In particular, the geometry of a slumped and damaged waste package must be examined, drift conditions must be modeled over a range of parameters, and the chemical degradation of DOE SNF and waste package materials must be considered for the expected high temperatures. The secondary intent of this calculation is to present a method for selecting conservative and bounding configurations for a wide range of end conditions

Conceptual design of retrieval systems for emplaced transuranic waste containers in a salt bed depository. Final report

International Nuclear Information System (INIS)

Fogleman, S.F.

1980-04-01

The US Department of Energy and the Nuclear Regulatory Commission have jurisdiction over the nuclear waste management program. Design studies were previously made of proposed repository site configurations for the receiving, processing, and storage of nuclear wastes. However, these studies did not provide operational designs that were suitable for highly reliable TRU retrieval in the deep geologic salt environment for the required 60-year period. The purpose of this report is to develop a conceptual design of a baseline retrieval system for emplaced transuranic waste containers in a salt bed depository. The conceptual design is to serve as a working model for the analysis of the performance available from the current state-of-the-art equipment and systems. Suggested regulations would be based upon the results of the performance analyses

Forced cooling of a nuclear waste repository mine drift - a scoping analysis

Energy Technology Data Exchange (ETDEWEB)

Boyd, R D [Sandia National Labs., Albuquerque, NM (USA)

1982-12-01

Nuclear waste repositories, with decay heat generation beneath the mine drift floors, are force-cooled with air so that re-entrance is possible many years after the waste has been buried. A numerical model has been developed which uses heat transfer coefficients as input. It has been demonstrated that mixed (forced and free) convective and surface roughness effects are significant and must be included in future experiments if reliable predictions are to be made of the time required to cool the repository. For example, when repository mine drifts in volcanic tuff are force-cooled, with forced convection being the only energy transport mechanism, it takes approx.= 0.1 year to cool the mine surface to a safe temperature. However, when mixed convection is the primary transport mechanism it takes approx.= 1.0 year to cool the mine. In addition to mixed convection, other effects are delineated.

Thermal Management and Analysis for a Potential Yucca Mountain Repository

International Nuclear Information System (INIS)

Dr. A. Van Luik

2004-01-01

In the current Yucca Mountain repository design concept, heat from the emplaced waste (mostly from spent nuclear fuel) would keep the temperature of the rock around the waste packages higher than the boiling point of water for hundreds to thousands of years after the repository is closed. The design concept allows below-boiling portions of the pillars between drifts to serve as pathways for the drainage of thermally mobilized water and percolating groundwater by limiting the distance that boiling temperatures extend into the surrounding rock. This design concept takes advantage of host rock dry out, which would create a dry environment within the emplacement drifts and reduce the amount of water that might otherwise be available to enter the drifts and contact the waste packages during this thermal pulse. Table 1 provides an overview of design constraints related to thermal management after repository closure. The Yucca Mountain repository design concept also provides flexibility to allow for operation over a range of lower thermal operating conditions. The thermal conditions within the emplacement drifts can be varied, along with the relative humidity, by modifying operational parameters such as the thermal output of the waste packages, the spacing of the waste packages in the emplacement drifts, and the duration and rate of active and passive ventilation. A lower range has been examined to quantify lower-temperature thermal conditions (temperatures and associated humidity conditions) in the emplacement drifts and to quantify impacts to the required emplacement area and excavated drift length. This information has been used to evaluate the potential long-term performance of a lower-temperature repository and to estimate the increase in costs associated with operating a lower-temperature repository. This presentation provides an overview of the thermal management evaluations that have been conducted to investigate a range of repository thermal conditions and

Thermal management and analysis for a potential yucca mountain repository

International Nuclear Information System (INIS)

Van Luik, A.

2005-01-01

In the current Yucca Mountain repository design concept, heat from the emplaced. waste (mostly from spent nuclear fuel.) would keep the temperature of the rock around the waste packages higher than the boiling point of water for hundreds to thousands of years after the repository is closed. The design concept allows below-boiling portions of the pillars between drifts to serve as pathways for the drainage of thermally mobilized water and percolating groundwater by limiting the distance that boiling temperatures extend into the surrounding rock. This design concept takes advantage of host rock dry out, which would create a dry environment within the emplacement drifts and reduce the amount of water that might otherwise be available to enter the drifts and contact the waste packages during this thermal pulse. The Yucca Mountain repository design concept also provides flexibility to allow for operation over a range of lower thermal operating conditions. The thermal conditions within the emplacement drifts can be varied, along with the relative humidity, by modifying operational parameters such as the thermal output of the waste packages, the spacing of the waste packages in the emplacement drifts, and. the duration and rate of active and passive ventilation. A lower range has been examined to quantify lower-temperature thermal conditions (temperatures and associated humidity conditions) in the emplacement drifts and to quantify impacts to the required emplacement area and excavated drift length. This information has been used to evaluate the potential long-term performance of a lower-temperature repository and to estimate the increase in costs associated with operating a lower-temperature repository. This presentation provides an overview of the thermal management evaluations that have been conducted to investigate a range of repository thermal conditions and includes a summary of the technical basis that supports these evaluations. The majority of the material

The Effects of Unheated Sections on Moisture Transport in the Emplacement Drift

International Nuclear Information System (INIS)

G. Danko; D. Bahrami; J.T. Birkholzer

2006-01-01

A thermal-hydrologic natural-ventilation model is configured for simulating temperature, humidity, and condensate distributions in the coupled domains of the in-drift airspace and the near-field rockmass. Meaningful results are obtained from the model for a practical application in which the beneficial effects of unheated drift sections are analyzed. Sensitivity to the axial dispersion coefficient is also studied with the model

Multiscale Thermohydrologic Model Analyses of Heterogeneity and Thermal-Loading Factors for the Proposed Repository at Yucca Mountain

International Nuclear Information System (INIS)

Glascoe, L.G.; Buscheck, T.A.; Gansemer, J.; Sun, Y.; Lee, K.

2002-01-01

The MultiScale ThermoHydrologic Model (MSTHM) predicts thermohydrologic (TH) conditions in emplacement drifts and the adjoining host rock throughout the proposed nuclear-waste repository at Yucca Mountain. The MSTHM is a computationally efficient approach that accounts for TH processes occurring at a scale of a few tens of centimeters around individual waste packages and emplacement drifts, and for heat flow at the multi-kilometer scale at Yucca Mountain. The modeling effort presented here is an early investigation of the repository and is simulated at a lower temperature mode and with a different panel loading than the repository currently being considered for license application. We present these recent lower temperature mode MSTHM simulations that address the influence of repository-scale thermal-conductivity heterogeneity and the influence of preclosure operational factors affecting thermal-loading conditions. We can now accommodate a complex repository layout with emplacement drifts lying in non-parallel planes using a superposition process that combines results from multiple mountain-scale submodels. This development, along with other improvements to the MSTHM, enables more rigorous analyses of preclosure operational factors. These improvements include the ability to (1) predict TH conditions on a drift-by-drift basis, (2) represent sequential emplacement of waste packages along the drifts, and (3) incorporate distance- and time-dependent heat-removal efficiency associated with drift ventilation. Alternative approaches to addressing repository-scale thermal-conductivity heterogeneity are investigated. We find that only one of the four MSTHM submodel types needs to incorporate thermal-conductivity heterogeneity. For a particular repository design, we find that the most influential parameters are (1) percolation-flux distribution, (2) thermal-conductivity heterogeneity within the host-rock units, (3) the sequencing of waste-package emplacement, and (4) the

Ground Control System Description Document

International Nuclear Information System (INIS)

Eric Loros

2001-01-01

The Ground Control System contributes to the safe construction and operation of the subsurface facility, including accesses and waste emplacement drifts, by maintaining the configuration and stability of the openings during construction, development, emplacement, and caretaker modes for the duration of preclosure repository life. The Ground Control System consists of ground support structures installed within the subsurface excavated openings, any reinforcement made to the rock surrounding the opening, and inverts if designed as an integral part of the system. The Ground Control System maintains stability for the range of geologic conditions expected at the repository and for all expected loading conditions, including in situ rock, construction, operation, thermal, and seismic loads. The system maintains the size and geometry of operating envelopes for all openings, including alcoves, accesses, and emplacement drifts. The system provides for the installation and operation of sensors and equipment for any required inspection and monitoring. In addition, the Ground Control System provides protection against rockfall for all subsurface personnel, equipment, and the engineered barrier system, including the waste package during the preclosure period. The Ground Control System uses materials that are sufficiently maintainable and that retain the necessary engineering properties for the anticipated conditions of the preclosure service life. These materials are also compatible with postclosure waste isolation performance requirements of the repository. The Ground Control System interfaces with the Subsurface Facility System for operating envelopes, drift orientation, and excavated opening dimensions, Emplacement Drift System for material compatibility, Monitored Geologic Repository Operations Monitoring and Control System for ground control instrument readings, Waste Emplacement/Retrieval System to support waste emplacement operations, and the Subsurface Excavation System

Uncanistered Spent Nuclear fuel Disposal Container System Description Document

International Nuclear Information System (INIS)

Pettit, N. E.

2001-01-01

The Uncanistered Spent Nuclear Fuel (SNF) 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 with intact uncanistered assemblies and/or individually canistered SNF assemblies and sealed in the surface waste handling facilities, transferred to the underground through the access drifts, and emplaced in emplacement drifts. The Uncanistered SNF Disposal Container provides long-term confinement of the commercial SNF placed inside, and withstands the loading, transfer, emplacement, and retrieval loads and environments. The Uncanistered SNF Disposal Container System provides containment of waste for a designated period of time, and limits radionuclide release. The disposal container maintains the waste in a designated configuration, withstands maximum handling and rockfall loads, limits the individual SNF assembly temperatures after emplacement, limits the introduction of moderator into the disposal container during the criticality control period, resists corrosion in the expected handling and repository environments, and provides containment of waste in the event of an accident

MULTISCALE THERMOHYDROLOGIC MODEL

International Nuclear Information System (INIS)

T. Buscheck

2005-01-01

The intended purpose of the multiscale thermohydrologic model (MSTHM) is to predict the possible range of thermal-hydrologic conditions, resulting from uncertainty and variability, in the repository emplacement drifts, including the invert, and in the adjoining host rock for the repository at Yucca Mountain. The goal of the MSTHM is to predict a reasonable range of possible thermal-hydrologic conditions within the emplacement drift. To be reasonable, this range includes the influence of waste-package-to-waste-package heat output variability relevant to the license application design, as well as the influence of uncertainty and variability in the geologic and hydrologic conditions relevant to predicting the thermal-hydrologic response in emplacement drifts. This goal is quite different from the goal of a model to predict a single expected thermal-hydrologic response. As a result, the development and validation of the MSTHM and the associated analyses using this model are focused on the goal of predicting a reasonable range of thermal-hydrologic conditions resulting from parametric uncertainty and waste-package-to-waste-package heat-output variability. Thermal-hydrologic conditions within emplacement drifts depend primarily on thermal-hydrologic conditions in the host rock at the drift wall and on the temperature difference between the drift wall and the drip-shield and waste-package surfaces. Thus, the ability to predict a reasonable range of relevant in-drift MSTHM output parameters (e.g., temperature and relative humidity) is based on valid predictions of thermal-hydrologic processes in the host rock, as well as valid predictions of heat-transfer processes between the drift wall and the drip-shield and waste-package surfaces. Because the invert contains crushed gravel derived from the host rock, the invert is, in effect, an extension of the host rock, with thermal and hydrologic properties that have been modified by virtue of the crushing (and the resulting

MULTISCALE THERMOHYDROLOGIC MODEL

Energy Technology Data Exchange (ETDEWEB)

T. Buscheck

2005-07-07

The intended purpose of the multiscale thermohydrologic model (MSTHM) is to predict the possible range of thermal-hydrologic conditions, resulting from uncertainty and variability, in the repository emplacement drifts, including the invert, and in the adjoining host rock for the repository at Yucca Mountain. The goal of the MSTHM is to predict a reasonable range of possible thermal-hydrologic conditions within the emplacement drift. To be reasonable, this range includes the influence of waste-package-to-waste-package heat output variability relevant to the license application design, as well as the influence of uncertainty and variability in the geologic and hydrologic conditions relevant to predicting the thermal-hydrologic response in emplacement drifts. This goal is quite different from the goal of a model to predict a single expected thermal-hydrologic response. As a result, the development and validation of the MSTHM and the associated analyses using this model are focused on the goal of predicting a reasonable range of thermal-hydrologic conditions resulting from parametric uncertainty and waste-package-to-waste-package heat-output variability. Thermal-hydrologic conditions within emplacement drifts depend primarily on thermal-hydrologic conditions in the host rock at the drift wall and on the temperature difference between the drift wall and the drip-shield and waste-package surfaces. Thus, the ability to predict a reasonable range of relevant in-drift MSTHM output parameters (e.g., temperature and relative humidity) is based on valid predictions of thermal-hydrologic processes in the host rock, as well as valid predictions of heat-transfer processes between the drift wall and the drip-shield and waste-package surfaces. Because the invert contains crushed gravel derived from the host rock, the invert is, in effect, an extension of the host rock, with thermal and hydrologic properties that have been modified by virtue of the crushing (and the resulting

In-Drift Accumulation of Fissile Material From Waste Packages Containing Plutonium Disposition Waste Forms

International Nuclear Information System (INIS)

H.W. Stockman; S. LeStrange

2000-01-01

The objective of this calculation is to provide estimates of the amount of fissile material flowing out of the waste package (source term) and the accumulation of fissile elements (U and Pu) in a crushed-tuff invert. These calculations provide input for the analysis of repository impacts of the Pu-ceramic waste forms. In particular, the source term results are used as input to the far-field accumulation calculation reported in Ref. 51, and the in-drift accumulation results are used as inputs for the criticality calculations reported in Ref. 2. The results are also summarized and interpreted in Ref. 52. The scope of this calculation is the waste package (WP) Viability Assessment (VA) design, which consists of an outer corrosion-allowance material (CAM) and an inner corrosion-resistant material (CRM). This design is used in this calculation in order to be consistent with earlier Pu-ceramic degradation calculations (Ref. 15). The impact of the new Enhanced Design Alternative-I1 (EDA-11) design on the results will be addressed in a subsequent report. The design of the invert (a leveling foundation, which creates a level surface of the drift floor and supports the WP mounting structure) is consistent with the EDA-I1 design. The invert will be composed of crushed stone and a steel support structure (Ref. 17). The scope of this calculation is also defined by the nominal degradation scenario, which involves the breach of the WP (Section 10.5.1.2, Ref. 48), followed by the influx of water. Water in the WP may, in time, gradually leach the fissile components and neutron absorbers out of the ceramic waste forms. Thus, the water in the WP may become laden with dissolved actinides (e.g., Pu and U), and may eventually overflow or leak from the WP. Once the water leaves the WP, it may encounter the invert, in which the actinides may reprecipitate. Several factors could induce reprecipitation; these factors include: the high surface area of the crushed stone, and the presence of

In-Drift Accumulation of Fissile Material From Waste Packages Containing Plutonium Disposition Waste Form

Energy Technology Data Exchange (ETDEWEB)

H.W> Stockman; S. LeStrange

2000-09-28

The objective of this calculation is to provide estimates of the amount of fissile material flowing out of the waste package (source term) and the accumulation of fissile elements (U and Pu) in a crushed-tuff invert. These calculations provide input for the analysis of repository impacts of the Pu-ceramic waste forms. In particular, the source term results are used as input to the far-field accumulation calculation reported in Ref. 51, and the in-drift accumulation results are used as inputs for the criticality calculations reported in Ref. 2. The results are also summarized and interpreted in Ref. 52. The scope of this calculation is the waste package (WP) Viability Assessment (VA) design, which consists of an outer corrosion-allowance material (CAM) and an inner corrosion-resistant material (CRM). This design is used in this calculation in order to be consistent with earlier Pu-ceramic degradation calculations (Ref. 15). The impact of the new Enhanced Design Alternative-I1 (EDA-11) design on the results will be addressed in a subsequent report. The design of the invert (a leveling foundation, which creates a level surface of the drift floor and supports the WP mounting structure) is consistent with the EDA-I1 design. The invert will be composed of crushed stone and a steel support structure (Ref. 17). The scope of this calculation is also defined by the nominal degradation scenario, which involves the breach of the WP (Section 10.5.1.2, Ref. 48), followed by the influx of water. Water in the WP may, in time, gradually leach the fissile components and neutron absorbers out of the ceramic waste forms. Thus, the water in the WP may become laden with dissolved actinides (e.g., Pu and U), and may eventually overflow or leak from the WP. Once the water leaves the WP, it may encounter the invert, in which the actinides may reprecipitate. Several factors could induce reprecipitation; these factors include: the high surface area of the crushed stone, and the presence of

Development and demonstration of prototype transportation equipment for emplacing HL vitrified waste canisters into small diameter bored horizontal disposal cells

International Nuclear Information System (INIS)

Seidler, Wolf K.; Bosgiraud, Jean-Michel; Londe, Louis

2008-01-01

Over a period of 4 and years the National Radioactive Waste Management Agency (Andra), working with a variety of Contractors mostly specializing in nuclear orientated mechanical applications, successfully designed, fabricated and demonstrated 2 very different prototype high level waste transport systems. The first system, based on air cushion technology, was developed primarily for very heavy loads (17 to 45 tonnes). The results of this work are described in a separate presentation (Paper 21) at this Conference. The second system, developed by Andra within the framework of the ESDRED Project, generally referred to as the 'Pushing Robot System' for vitrified waste canisters, is the subject of this paper. The 'Pushing Robot System' is a part of the French national disposal concept that is described in Andra's 'Dossier 2005'. The latter is a public document that can be viewed on Andra's web site (www.andra.fr). The 'Pushing Robot System' system is designed for the deep geological disposal (in clay formations) of 'C' type vitrified waste canisters. In its entirety the system provides for the transport, emplacement and, if necessary, the retrieval of those canisters. Nothing in the design of the Andra emplacement equipment would preclude its utilization in horizontal openings in other types of geological settings. Over a period of some 8 years Andra has developed the 'Pushing Robot System' in 3 phases. Initially there was only the 'Conceptual Design' (Phase 1) which was incorporated in the Dossier 2005. This was followed by Phase 2 i.e. the design and fabrication of a simplified full scale prototype system henceforth referred to a P1, which includes a Pushing Robot, a Dummy Canister and a Test Bench. P1 details were also incorporated in the Dossier 2005. Finally, during Phase 3, a second more comprehensive full scale prototype system P2 has been designed and is being assembled and tested this month. This system includes a Transport Shuttle, a Transfer Shielding Cask, a

Subsurface Ventilation System Description Document

Energy Technology Data Exchange (ETDEWEB)

Eric Loros

2001-07-25

The Subsurface Ventilation System supports the construction and operation of the subsurface repository by providing air for personnel and equipment and temperature control for the underground areas. Although the system is located underground, some equipment and features may be housed or located above ground. The system ventilates the underground by providing ambient air from the surface throughout the subsurface development and emplacement areas. The system provides fresh air for a safe work environment and supports potential retrieval operations by ventilating and cooling emplacement drifts. The system maintains compliance within the limits established for approved air quality standards. The system maintains separate ventilation between the development and waste emplacement areas. The system shall remove a portion of the heat generated by the waste packages during preclosure to support thermal goals. The system provides temperature control by reducing drift temperature to support potential retrieval operations. The ventilation system has the capability to ventilate selected drifts during emplacement and retrieval operations. The Subsurface Facility System is the main interface with the Subsurface Ventilation System. The location of the ducting, seals, filters, fans, emplacement doors, regulators, and electronic controls are within the envelope created by the Ground Control System in the Subsurface Facility System. The Subsurface Ventilation System also interfaces with the Subsurface Electrical System for power, the Monitored Geologic Repository Operations Monitoring and Control System to ensure proper and safe operation, the Safeguards and Security System for access to the emplacement drifts, the Subsurface Fire Protection System for fire safety, the Emplacement Drift System for repository performance, and the Backfill Emplacement and Subsurface Excavation Systems to support ventilation needs.

Subsurface Ventilation System Description Document

Energy Technology Data Exchange (ETDEWEB)

NONE

2000-10-12

The Subsurface Ventilation System supports the construction and operation of the subsurface repository by providing air for personnel and equipment and temperature control for the underground areas. Although the system is located underground, some equipment and features may be housed or located above ground. The system ventilates the underground by providing ambient air from the surface throughout the subsurface development and emplacement areas. The system provides fresh air for a safe work environment and supports potential retrieval operations by ventilating and cooling emplacement drifts. The system maintains compliance within the limits established for approved air quality standards. The system maintains separate ventilation between the development and waste emplacement areas. The system shall remove a portion of the heat generated by the waste packages during preclosure to support thermal goals. The system provides temperature control by reducing drift temperature to support potential retrieval operations. The ventilation system has the capability to ventilate selected drifts during emplacement and retrieval operations. The Subsurface Facility System is the main interface with the Subsurface Ventilation System. The location of the ducting, seals, filters, fans, emplacement doors, regulators, and electronic controls are within the envelope created by the Ground Control System in the Subsurface Facility System. The Subsurface Ventilation System also interfaces with the Subsurface Electrical System for power, the Monitored Geologic Repository Operations Monitoring and Control System to ensure proper and safe operation, the Safeguards and Security System for access to the emplacement drifts, the Subsurface Fire Protection System for fire safety, the Emplacement Drift System for repository performance, and the Backfill Emplacement and Subsurface Excavation Systems to support ventilation needs.

Abstraction of Drift-Scale Coupled Processes

International Nuclear Information System (INIS)

Francis, N.D.; Sassani, D.

2000-01-01

This Analysis/Model Report (AMR) describes an abstraction, for the performance assessment total system model, of the near-field host rock water chemistry and gas-phase composition. It also provides an abstracted process model analysis of potentially important differences in the thermal hydrologic (TH) variables used to describe the performance of a geologic repository obtained from models that include fully coupled reactive transport with thermal hydrology and those that include thermal hydrology alone. Specifically, the motivation of the process-level model comparison between fully coupled thermal-hydrologic-chemical (THC) and thermal-hydrologic-only (TH-only) is to provide the necessary justification as to why the in-drift thermodynamic environment and the near-field host rock percolation flux, the essential TH variables used to describe the performance of a geologic repository, can be obtained using a TH-only model and applied directly into a TSPA abstraction without recourse to a fully coupled reactive transport model. Abstraction as used in the context of this AMR refers to an extraction of essential data or information from the process-level model. The abstraction analysis reproduces and bounds the results of the underlying detailed process-level model. The primary purpose of this AMR is to abstract the results of the fully-coupled, THC model (CRWMS M andO 2000a) for effects on water and gas-phase composition adjacent to the drift wall (in the near-field host rock). It is assumed that drift wall fracture water and gas compositions may enter the emplacement drift before, during, and after the heating period. The heating period includes both the preclosure, in which the repository drifts are ventilated, and the postclosure periods, with backfill and drip shield emplacement at the time of repository closure. Although the preclosure period (50 years) is included in the process models, the postclosure performance assessment starts at the end of this initial period

Repository sealing concepts for the Nevada nuclear waste storage Investigations Project

International Nuclear Information System (INIS)

Fernandez, J.A.; Freshley, M.D.

1984-08-01

This report describes concepts for sealing a nuclear waste repository in an unsaturated tuff environment. The repository site under consideration is Yucca Mountain, which is on and adjacent to the Nevada Test Site. The hydrogeology of Yucca Mountain, preliminary repository concepts, functional requirements and performance criteria for sealing, federal and state regulations, and hydrological calculations are considered in developing the sealing concepts. Water flow through the unsaturated zone is expected to be small and generally vertically downward with some potential to occur through discrete fault and fracture zones. These assumptions are used in developing sealing concepts for shafts, ramps, and boreholes. Sealing of discrete, water-producing faults and fracture zones encountered in horizontal emplacement holes and in access and emplacement drifts is also described. 49 references, 21 figures, 6 tables

Verification of Drift Seal Systems at the Morsleben Repository, Germany - Proof of Technical Feasibility and Functionality

International Nuclear Information System (INIS)

Wollrath, Juergen; Mauke, R.; Siemann, M.

2014-01-01

The Morsleben repository (ERAM) for low- and intermediate-level mainly short-lived radioactive wastes is located in a former salt mine. The closure concept is based on extensive backfilling with salt concrete complemented by seals that should prevent the penetration of solution into the waste emplacement areas and the emission of radionuclides out of these areas. The drift seals located in rock salt are made up of one or more segments of salt concrete in lengths between 25 m and 30 m. The sealing structure consists of three components: the seal body made of salt concrete, the contact zone between the seal body and the surrounding rock salt and the rock salt excavation damaged zone (EDZ). To demonstrate the feasibility of constructing such a seal structure an in-situ experiment is performed and a test drift and an accompanying parallel drift have been excavated for the experiment. The trial construction also comprises geotechnical instrumentation for stress, strain, displacement, temperature and pore pressure measurements and a comprehensive site investigation programme has been carried out, notably with regard to the stress state and the convergence behaviour of the surrounding rock salt. In addition to in-situ measurements, test specimens from different areas of the construction have been drilled and tested (strength and permeability). The pressure chamber has also been filled with brine solution to ascertain the permeability of the whole sealing structure. Another in-situ experiment is planned for the seal to be built in the non-creeping anhydrite. Both in-situ experiments will aid to prove the technical feasibility and functionality of the drift seal systems

Subsurface Facility System Description Document

International Nuclear Information System (INIS)

Eric Loros

2001-01-01

The Subsurface Facility System encompasses the location, arrangement, size, and spacing of the underground openings. This subsurface system includes accesses, alcoves, and drifts. This system provides access to the underground, provides for the emplacement of waste packages, provides openings to allow safe and secure work conditions, and interfaces with the natural barrier. This system includes what is now the Exploratory Studies Facility. The Subsurface Facility System physical location and general arrangement help support the long-term waste isolation objectives of the repository. The Subsurface Facility System locates the repository openings away from main traces of major faults, away from exposure to erosion, above the probable maximum flood elevation, and above the water table. The general arrangement, size, and spacing of the emplacement drifts support disposal of the entire inventory of waste packages based on the emplacement strategy. The Subsurface Facility System provides access ramps to safely facilitate development and emplacement operations. The Subsurface Facility System supports the development and emplacement operations by providing subsurface space for such systems as ventilation, utilities, safety, monitoring, and transportation

Materials aspects of nuclear waste isolation

International Nuclear Information System (INIS)

Bennett, J.W.

1984-01-01

This paper is intended to provide an overview of the nuclear waste repository performance requirements and the roles which we expect materials to play in meeting these requirements. The objective of the U.S. Dept. of Energy's (DOE) program is to provide for the safe, permanent isolation of high-level radioactive wastes from the public. The Nuclear Waste Policy Act of 1982 (the Act) provides the mandate to accomplish this objective by establishing a program timetable, a schedule of procedures to be followed, and program funding (1 mil/kwhr for all nuclear generated electricity). The centerpiece of this plan is the design and operation of a mined geologic repository system for the permanent isolation of radioactive wastes. A nuclear waste repository contains several thousand acres of tunnels and drifts into which the nuclear waste will be emplaced, and several hundred acres for the facilities on the surface in which the waste is received, handled, and prepared for movement underground. With the exception of the nuclear material-related facilities, a repository is similar to a standard mining operation. The difference comes in what a repository is supposed to do - to contain an isolate nuclear waste from man and the environment

Results from an International Simulation Study on Couples Thermal, Hydrological, and Mechanical (THM) Processes Near Geological Nuclear Waste Repositories

International Nuclear Information System (INIS)

J. Rutqvist; J.T. Birkholzer; M. Chijimatsu; O. Kolditz; Q.S. Liu; Y. Oda; W. Wang; C.Y. Zhang

2006-01-01

As part of the ongoing international code comparison project DECOVALEX, four research teams used five different models to simulate coupled thermal, hydrological, and mechanical (THM) processes near underground waste emplacement drifts. The simulations were conducted for two generic repository types with open or back-filled repository drifts under higher and lower post-closure temperature, respectively. In the completed first model inception phase of the project, a good agreement was achieved between the research teams in calculating THM responses for both repository types, although some disagreement in hydrological responses are currently being resolved. Good agreement in the basic thermal-mechanical responses was achieved for both repository types, even with some teams using relatively simplified thermal-elastic heat-conduction models that neglect complex near-field thermal-hydrological processes. The good agreement between the complex and simplified (and well-known) process models indicates that the basic thermal-mechanical responses can be predicted with a relatively high confidence level. The research teams have now moved on to the second phase of the project, the analysis of THM-induced permanent (irreversible) changes and the impact of those changes on the fluid flow field near an emplacement drift

Investigation of excavation stability in a finite repository

International Nuclear Information System (INIS)

St John, C.M.; Mitchell, S.J.

1987-05-01

Two-dimensional thermal-mechanical analyses of a hypothetical repository in tuff were performed using a boundary-element computer code, with the objective of investigating the relative stability of several waste emplacement panel access drifts. The drifts were assumed to be located within a central shaft pillar, within a small pillar between adjacent panels, or at the repository perimeter. The results of analyses indicate that matrix failure is unlikely and that the extent of activation of pre-existing joints is influenced by the drift location and the initial state of stress. The results also show that several years after waste emplacement the influence of the initial State of Stress is small compared to the influence of thermally induced stresses. Regions of joint activation are seen to be most extensive around the drifts between adjacent panels. The predicted extent of such activation was small when it was assumed that the joints were vertical or near vertical. More extensive regions of activation of shallow dipping joints were noted. However, joints of such orientation are believed to occur relatively infrequently in the candidate horizon for waste emplacement and the activation of the shallow dipping joints does not necessarily lead to drift stability concerns. 3 refs., 24 figs., 7 tabs

Retrievability as proposed in the US repository concept

International Nuclear Information System (INIS)

Harrington, P.G.

2000-01-01

The Nuclear Waste Policy Act states that any repository shall be designed and constructed to permit retrieval. Reasons for retrieval include public health and safety, environmental concerns, and recovery of economically valuable contents of spent nuclear fuel. The Nuclear Regulatory Commission requires that waste must be retrievable at any time up to 50 years after start of emplacement. The US Department of Energy intends to maintain a retrieval capability throughout the preclosure period. Possible preclosure periods range from a minimum of 50 years to as much as 300 years. Repository closure includes sealing all accessible portions of the repository, including ventilation shafts, access ramps and boreholes. Drip shields will be installed over the waste packages. Access to the repository after closure is not intended. The proposed repository includes horizontal emplacement drifts located in the unsaturated zone. The emplacement drift centerline spacing is 81 meters to provide a subboiling region between drifts for water drainage. A drip shield covers the waste packages. All emplacement drifts remain open until closure of the repository, providing performance benefits such as removing heat and moisture during the preclosure period and lowering postclosure temperatures. This does not impede retrieval, permitting a reversal of the emplacement process to accomplish retrieval under normal conditions. The preclosure period is therefore not to enhance retrievability, but does improve performance, and the resultant extension of the retrievability capability is a secondary effect. Information must be provided from the performance confirmation program to support a regulatory decision to close. Closure would isolate the repository from the accessible environment, preclude preferential flowpaths for water into the mountain, and minimize the possibility of inadvertent intrusion. (author)

Drift Degradation Analysis

International Nuclear Information System (INIS)

G.H. Nieder-Westermann

2005-01-01

The outputs from the drift degradation analysis support scientific analyses, models, and design calculations, including the following: (1) Abstraction of Drift Seepage; (2) Seismic Consequence Abstraction; (3) Structural Stability of a Drip Shield Under Quasi-Static Pressure; and (4) Drip Shield Structural Response to Rock Fall. This report has been developed in accordance with ''Technical Work Plan for: Regulatory Integration Modeling of Drift Degradation, Waste Package and Drip Shield Vibratory Motion and Seismic Consequences'' (BSC 2004 [DIRS 171520]). The drift degradation analysis includes the development and validation of rockfall models that approximate phenomenon associated with various components of rock mass behavior anticipated within the repository horizon. Two drift degradation rockfall models have been developed: the rockfall model for nonlithophysal rock and the rockfall model for lithophysal rock. These models reflect the two distinct types of tuffaceous rock at Yucca Mountain. The output of this modeling and analysis activity documents the expected drift deterioration for drifts constructed in accordance with the repository layout configuration (BSC 2004 [DIRS 172801])

Preliminary assessment of the thermal effects of an annular air space surrounding an emplaced nuclear waste canister

International Nuclear Information System (INIS)

Davis, B.W.

1979-01-01

Modeling results have previously shown that the presence of a large air space (e.g., a repository room) within a nuclear waste repository is expected to cause a waste canister's temperature to remain cooler than it would otherwise be. Results presented herein show that an annular air space surrounding the waste canisters can have similar cooling effects under certain prescribable conditions; for a 16 ft x 1 ft diameter canister containing 650 PWR rods which initially generate a total of 4.61 kw, analysis will show that annular air spaces greater than 11 in will permit the canister surface to attain peak temperatures lower than that which would result from a zero-gap/perfect thermal contact. It was determined that the peak radial temperature gradient in the salt varies in proportion to the inverse of the drill hole radius. Thermal radiation is shown to be the dominant mode of heat transfer across an annular air space during the first two years after emplacement. Finally, a methodology is presented which will allow investigators to easily model radiation and convection heat transfer through air spaces by treating the space as a conduction element that possesses non-linear temperature dependent conductivity

Logistics Modeling of Emplacement Rate and Duration of Operations for Generic Geologic Repository Concepts

Energy Technology Data Exchange (ETDEWEB)

Kalinina, Elena Arkadievna; Hardin, Ernest

2015-11-01

This study identified potential geologic repository concepts for disposal of spent nuclear fuel (SNF) and (2) evaluated the achievable repository waste emplacement rate and the time required to complete the disposal for these concepts. Total repository capacity is assumed to be approximately 140,000 MT of spent fuel. The results of this study provide an important input for the rough-order-of-magnitude (ROM) disposal cost analysis. The disposal concepts cover three major categories of host geologic media: crystalline or hard rock, salt, and argillaceous rock. Four waste package sizes are considered: 4PWR/9BWR; 12PWR/21BWR; 21PWR/44BWR, and dual purpose canisters (DPCs). The DPC concepts assume that the existing canisters will be sealed into disposal overpacks for direct disposal. Each concept assumes one of the following emplacement power limits for either emplacement or repository closure: 1.7 kW; 2.2 kW; 5.5 kW; 10 kW; 11.5 kW, and 18 kW.

Logistics Modeling of Emplacement Rate and Duration of Operations for Generic Geologic Repository Concepts

International Nuclear Information System (INIS)

Kalinina, Elena Arkadievna; Hardin, Ernest

2015-01-01

This study identified potential geologic repository concepts for disposal of spent nuclear fuel (SNF) and (2) evaluated the achievable repository waste emplacement rate and the time required to complete the disposal for these concepts. Total repository capacity is assumed to be approximately 140,000 MT of spent fuel. The results of this study provide an important input for the rough-order-of-magnitude (ROM) disposal cost analysis. The disposal concepts cover three major categories of host geologic media: crystalline or hard rock, salt, and argillaceous rock. Four waste package sizes are considered: 4PWR/9BWR; 12PWR/21BWR; 21PWR/44BWR, and dual purpose canisters (DPCs). The DPC concepts assume that the existing canisters will be sealed into disposal overpacks for direct disposal. Each concept assumes one of the following emplacement power limits for either emplacement or repository closure: 1.7 kW; 2.2 kW; 5.5 kW; 10 kW; 11.5 kW, and 18 kW.

Thermal/thermomechanical analyses for the room region with horizontal and vertial modes of emplacement

International Nuclear Information System (INIS)

1988-01-01

Extensive thermal/thermomechanical analyses of the Site Characterization Plan-Conceptual Design at the Deaf Smith county Site, Texas, have been carried out for the room region with horizontal and vertical modes of emplacement. The main purpose of this study is to make a good comparison between these two modes of emplacement in this region. Homogeneous and nonhomogeneous strata under isothermal or transient temperature conditions cases were considered in the analyses. Furthermore, various pillar widths for the vertical mode emplacement were also taken into consideration. Only spent fuel (SF) waste was considered in this study. Finite element method was used throughout the analyses. The thermal responses were evaluated using SPECTROM-41 while the thermomechanical responses were calculated using SPECTROM-32. Thermal and thermomechanical comparisons between the two modes of emplacement for various cases were presented in this paper

Influence of heat flow on drift closure during Climax granite spent-fuel test: measurements and calculations

International Nuclear Information System (INIS)

Butkovich, T.R.; Yow, J.L. Jr.; Montan, D.N.

1982-01-01

Measurements of drift closure caused by the thermal load have been made routinely during the Spent Fuel Test in Climax granite since about six weeks after emplacement of the fuel. Horizontal and vertical closure was measured with a manually operated tape extensometer at various locations along the lengths of the drifts. Average closures ranged from 0 to 0.6 mm, horizontal and vertical, out to about 2.2 years since the start of the test. At the same time, displacements from the thermal loads were measured with rod extensometers emplaced to measure relative displacements between hole collars and anchor points in holes drilled from two parallel heater drifts. These data are compared with thermo-elastic finite element calculations which utilized measured properties of the Climax granite. The calculations show that more than half of the closures occur between fuel installation and the first closure measurement. The comparisons show that the results track each other, in that where closure followed by dilation is measured, the calculations also show this effect. The agreement is excellent, considering the averaged measured closures remain within 30% of the total calculated drift closures and the extremely small magnitude of the relative displacements (0.5 mm), measured or calculated

Modeled near-field environment porosity modifications due to coupled thermohydrologic and geochemical processes

International Nuclear Information System (INIS)

Glassley, W. E.; Nitao, J. J.

1998-01-01

Heat deposited by waste packages in nuclear waste repositories can modify rock properties by instigating mineral dissolution and precipitation along hydrothermal flow pathways. Modeling this reactive transport requires coupling fluid flow to permeability changes resulting from dissolution and precipitation. Modification of the NUFT thermohydrologic (TH) code package to account for this coupling in a simplified geochemical system has been used to model the time- dependent change in porosity, permeability, matrix and fracture saturation, and temperature in the vicinity of waste-emplacement drifts, using conditions anticipated for the potential Yucca Mountain repository. The results show, within a few hundred years, dramatic porosity reduction approximately 10 m above emplacement drifts. Most of this reduction is attributed to deposition of solute load at the boiling front, although some of it also results from decreasing temperature along the flow path. The actual distribution of the nearly sealed region is sensitive to the time- dependent characteristics of the thermal load imposed on the environment and suggests that the geometry of the sealed region can be engineered by managing the waste-emplacement strategy

Feasibility studies of air placed techniques as emplacement means of different backfilling materials in underground radioactive waste disposal

International Nuclear Information System (INIS)

Atabek, R.; Conche, P.; Lajudie, A.; Revertegat, E.

1992-01-01

Air placed techniques are likely to be used as emplacement means of different backfilling materials in underground waste repositories. A literature survey of the air placed techniques and equipments leads to the choice of the dry process taking into account the emplacement constraints (distance: 300 m, flow: 10 m 3 /h) and the large variety of materials to be placed. Tests performed in the case of cement-based materials (with and without addition of silica fumes), for different types of cement and as a function of the incidence of the jet, show that it is possible to put in place mortars of good quality. However heterogeneity in the material composition is found when the jet is stopped. This problem may be partly solved by a better automation of the process. Complementary tests, carried out with the preselected clay of Fourges Cahaignes, clearly demonstrate the ability of the air placed technique to put in place pure clay: a dry density of 1.50kg/m 3 is reached in the case of coarse material and for a final water content of 30% (in weight). Feasibility tests performed on clay-sand mixtures are not conclusive due to an unappropriate granulometry distribution of the sand. 11 figs., 9 tabs

Dike Propagation Near Drifts

International Nuclear Information System (INIS)

2002-01-01

The purpose of this Analysis and Model Report (AMR) supporting the Site Recommendation/License Application (SR/LA) for the Yucca Mountain Project is the development of elementary analyses of the interactions of a hypothetical dike with a repository drift (i.e., tunnel) and with the drift contents at the potential Yucca Mountain repository. This effort is intended to support the analysis of disruptive events for Total System Performance Assessment (TSPA). This AMR supports the Process Model Report (PMR) on disruptive events (CRWMS M and O 2000a). This purpose is documented in the development plan (DP) ''Coordinate Modeling of Dike Propagation Near Drifts Consequences for TSPA-SR/LA'' (CRWMS M and O 2000b). Evaluation of that Development Plan and the work to be conducted to prepare Interim Change Notice (ICN) 1 of this report, which now includes the design option of ''Open'' drifts, indicated that no revision to that DP was needed. These analyses are intended to provide reasonable bounds for a number of expected effects: (1) Temperature changes to the waste package from exposure to magma; (2) The gas flow available to degrade waste containers during the intrusion; (3) Movement of the waste package as it is displaced by the gas, pyroclasts and magma from the intruding dike (the number of packages damaged); (4) Movement of the backfill (Backfill is treated here as a design option); (5) The nature of the mechanics of the dike/drift interaction. These analyses serve two objectives: to provide preliminary analyses needed to support evaluation of the consequences of an intrusive event and to provide a basis for addressing some of the concerns of the Nuclear Regulatory Commission (NRC) expressed in the Igneous Activity Issue Resolution Status Report

W1045 environment surf drip shield and waste package outer barrier

International Nuclear Information System (INIS)

Gdowski, G.

1999-01-01

The environments on the drip shield and waste package outer barrier are controlled by the compositions of the waters that contact these components. the temperature (T) of these components, and the effective relative humidity (RH) at these components. Because the composition of the waters that are expected to enter the emplacement drifts (either by seepage flow or by episodic flow) have not been specified: well J13 water was chosen as the reference water (Harrar 1990). Section 6.2 discusses the accessible RH for the temperatures of interest at the repository horizon. Section 6.3 discusses the adsorption of water on metal alloys in the absence of hygroscopic salts. Because the temperatures of the DSs and the WPOBs are higher than those of the surrounding near-field environment, the relative humidity at the DSs and the WPOBs will be lower than that of the surrounding near-field environment. This difference is a result of the water partial pressure in the drift being constant and no higher than the equilibrium water vapor pressure at the temperature of the drift wall

Results From an International Simulation Study on Coupled Thermal, Hydrological, and Mechanical (THM) Processes Near Geological Nuclear Waste Repositories

International Nuclear Information System (INIS)

J. Rutqvist; D. Barr; J.T. Birkholzer; M. Chijimatsu; O. Kolditz; Q. Liu; Y. Oda; W. Wang; C. Zhang

2006-01-01

As part of the ongoing international DECOVALEX project, four research teams used five different models to simulate coupled thermal, hydrological, and mechanical (THM) processes near waste emplacement drifts of geological nuclear waste repositories. The simulations were conducted for two generic repository types, one with open and the other with back-filled repository drifts, under higher and lower postclosure temperatures, respectively. In the completed first model inception phase of the project, a good agreement was achieved between the research teams in calculating THM responses for both repository types, although some disagreement in hydrological responses is currently being resolved. In particular, good agreement in the basic thermal-mechanical responses was achieved for both repository types, even though some teams used relatively simplified thermal-elastic heat-conduction models that neglected complex near-field thermal-hydrological processes. The good agreement between the complex and simplified process models indicates that the basic thermal-mechanical responses can be predicted with a relatively high confidence level

WP EMPLACEMENT CONTROL AND COMMUNICATION EQUIPMENT DESCRIPTIONS

International Nuclear Information System (INIS)

Raczka, N.T.

1997-01-01

The objective and scope of this document are to list and briefly describe the major control and communication equipment necessary for waste package emplacement at the proposed nuclear waste repository at Yucca Mountain. Primary performance characteristics and some specialized design features of the required equipment are explained and summarized in the individual subsections of this document. This task was evaluated in accordance with QAP-2-0 and found not to be quality affecting. Therefore, this document was prepared in accordance with NAP-MG-012. The following control and communication equipment are addressed in this document: (1) Programmable Logic Controllers (PLC's); (2) Leaky Feeder Radio Frequency Communication Equipment; (3) Slotted Microwave guide Communication Equipment; (4) Vision Systems; (5) Radio Control Equipment; and (6) Enclosure Cooling Systems

The disposal of Canada's nuclear fuel waste: a study of postclosure safety of in-room emplacement of used CANDU fuel in copper containers in permeable plutonic rock volume 1: summary

International Nuclear Information System (INIS)

Wikjord, A.G.; Baumgartner, P.; Johnson, L.H.; Stanchell, F.W.; Zach, R.; Goodwin, B.W.

1996-06-01

The concept for disposal of Canada's nuclear fuel waste involves isolating the waste in corrosion-resistant containers emplaced and sealed within a vault at a depth of 500 to 1000 m in plutonic rock of the Canadian Shield. The case for the acceptability of the concept as a means of safely disposing of Canada's nuclear fuel waste is presented in an Environmental Impact Statement (EIS) The disposal concept permits a choice of methods, materials, site locations and designs. The EIS presents a case study of the long-term (i.e., postclosure) performance of a hypothetical implementation of the concept, referred to in this report as the reference disposal system. The reference disposal system is based on borehole emplacement of used CANDU fuel in Grade-2 titanium alloy containers in low-permeability, sparsely fractured plutonic rock of the Canadian Shield. We evaluate the long-term performance of another hypothetical implementation of the concept based on in-room emplacement of used CANDU fuel in copper containers in permeable plutonic rock. The geological characteristics of the geosphere assumed for this study result in short groundwater travel times from the disposal vault to the surface. In the present study, the principal barrier to the movement of contaminants is the long-lasting copper container. We show that the long-lasting container can effectively compensate for a permeable host rock which results in an unfavourable groundwater flow condition. These studies illustrate the flexibility of AECL's disposal concept to take advantage of the retention, delay, dispersion, dilution and radioactive decay of contaminants in a system of natural barriers provided by the geosphere and hydrosphere and of engineered barriers provided by the waste form, container, buffer, backfills, other vault seals and grouts. In an actual implementation, the engineered system would be designed for the geological conditions encountered at the host site. 34 refs., 2 tabs., 11 figs

Modelling magma-drift interaction at the proposed high-level radioactive waste repository at Yucca Mountain, Nevada, USA

NARCIS (Netherlands)

Woods, Andrew W.; Sparks, Steve; Bokhove, Onno; Lejeune, Anne-Marie; Connor, Charles B.; Hill, Britain E.

2002-01-01

We examine the possible ascent of alkali basalt magma containing 2 wt percent water through a dike and into a horizontal subsurface drift as part of a risk assessment for the proposed high-level radioactive waste repository beneath Yucca Mountain, Nevada, USA. On intersection of the dike with the

ABSTRACTION OF DRIFT SEEPAGE

International Nuclear Information System (INIS)

Wilson, Michael L.

2001-01-01

Drift seepage refers to flow of liquid water into repository emplacement drifts, where it can potentially contribute to degradation of the engineered systems and release and transport of radionuclides within the drifts. Because of these important effects, seepage into emplacement drifts is listed as a ''principal factor for the postclosure safety case'' in the screening criteria for grading of data in Attachment 1 of AP-3.15Q, Rev. 2, ''Managing Technical Product Inputs''. Abstraction refers to distillation of the essential components of a process model into a form suitable for use in total-system performance assessment (TSPA). Thus, the purpose of this analysis/model is to put the information generated by the seepage process modeling in a form appropriate for use in the TSPA for the Site Recommendation. This report also supports the Unsaturated-Zone Flow and Transport Process Model Report. The scope of the work is discussed below. This analysis/model is governed by the ''Technical Work Plan for Unsaturated Zone Flow and Transport Process Model Report'' (CRWMS MandO 2000a). Details of this activity are in Addendum A of the technical work plan. The original Work Direction and Planning Document is included as Attachment 7 of Addendum A. Note that the Work Direction and Planning Document contains tasks identified for both Performance Assessment Operations (PAO) and Natural Environment Program Operations (NEPO). Only the PAO tasks are documented here. The planning for the NEPO activities is now in Addendum D of the same technical work plan and the work is documented in a separate report (CRWMS MandO 2000b). The Project has been reorganized since the document was written. The responsible organizations in the new structure are the Performance Assessment Department and the Unsaturated Zone Department, respectively. The work plan for the seepage abstraction calls for determining an appropriate abstraction methodology, determining uncertainties in seepage, and providing

Seepage Model for PA Including Dift Collapse

Energy Technology Data Exchange (ETDEWEB)

G. Li; C. Tsang

2000-12-20

The purpose of this Analysis/Model Report (AMR) is to document the predictions and analysis performed using the Seepage Model for Performance Assessment (PA) and the Disturbed Drift Seepage Submodel for both the Topopah Spring middle nonlithophysal and lower lithophysal lithostratigraphic units at Yucca Mountain. These results will be used by PA to develop the probability distribution of water seepage into waste-emplacement drifts at Yucca Mountain, Nevada, as part of the evaluation of the long term performance of the potential repository. This AMR is in accordance with the ''Technical Work Plan for Unsaturated Zone (UZ) Flow and Transport Process Model Report'' (CRWMS M&O 2000 [153447]). This purpose is accomplished by performing numerical simulations with stochastic representations of hydrological properties, using the Seepage Model for PA, and evaluating the effects of an alternative drift geometry representing a partially collapsed drift using the Disturbed Drift Seepage Submodel. Seepage of water into waste-emplacement drifts is considered one of the principal factors having the greatest impact of long-term safety of the repository system (CRWMS M&O 2000 [153225], Table 4-1). This AMR supports the analysis and simulation that are used by PA to develop the probability distribution of water seepage into drift, and is therefore a model of primary (Level 1) importance (AP-3.15Q, ''Managing Technical Product Inputs''). The intended purpose of the Seepage Model for PA is to support: (1) PA; (2) Abstraction of Drift-Scale Seepage; and (3) Unsaturated Zone (UZ) Flow and Transport Process Model Report (PMR). Seepage into drifts is evaluated by applying numerical models with stochastic representations of hydrological properties and performing flow simulations with multiple realizations of the permeability field around the drift. The Seepage Model for PA uses the distribution of permeabilities derived from air injection testing in

Postclosure performance assessment of the SCP [Site Characterization Plan] conceptual design for horizontal emplacement: Revision 1

International Nuclear Information System (INIS)

1987-08-01

This report is a preliminary postclosure performance assessment of the repository design specified in the Site Characterization Plan Conceptual Design Report (SCP-CDR) for horizontal emplacement of high-level nuclear waste. At the time that these analyses were done, horizontal emplacement was the preferred orientation for the waste packages but vertical emplacement is now the reference design. This assessment consists of (1) a review of the regulatory requirements and strategy to demonstrate compliance with these requirements, (2) an analysis of the performance of the total repository system, (3) an analysis of the thermomechanical behavior of the repository, (4) an analysis of brine mobility in the repository, (5) an analysis of the waste package performance, (6) an analysis of the performance of seals, and (7) comments on the sensitivity of the various performance measures to uncertainties in the data and models. These are preliminary analyses and, in most cases, involve bounding calculations of the repository behavior. They have several purposes including (1) assessing how well this conceptual design ''measures up'' against requirements, (2) gaining experience in implementing the performance assessment strategy and tools and thereby learning where improvements are needed, (3) helping to identify needed data, and (4) helping to indicate required design modifications. 26 refs., 40 figs., 20 tabs

Nuclear waste package fabricated from concrete

International Nuclear Information System (INIS)

Pfeiffer, P.A.; Kennedy, J.M.

1987-03-01

After the United States enacted the Nuclear Waste Policy Act in 1983, the Department of Energy must design, site, build and operate permanent geologic repositories for high-level nuclear waste. The Department of Energy has recently selected three sites, one being the Hanford Site in the state of Washington. At this particular site, the repository will be located in basalt at a depth of approximately 3000 feet deep. The main concern of this site, is contamination of the groundwater by release of radionuclides from the waste package. The waste package basically has three components: the containment barrier (metal or concrete container, in this study concrete will be considered), the waste form, and other materials (such as packing material, emplacement hole liners, etc.). The containment barriers are the primary waste container structural materials and are intended to provide containment of the nuclear waste up to a thousand years after emplacement. After the containment barriers are breached by groundwater, the packing material (expanding sodium bentonite clay) is expected to provide the primary control of release of radionuclide into the immediate repository environment. The loading conditions on the concrete container (from emplacement to approximately 1000 years), will be twofold; (1) internal heat of the high-level waste which could be up to 400 0 C; (2) external hydrostatic pressure up to 1300 psi after the seepage of groundwater has occurred in the emplacement tunnel. A suggested container is a hollow plain concrete cylinder with both ends capped. 7 refs

Geochemical homogeneity of tuffs at the potential repository level, Yucca Mountain, Nevada

International Nuclear Information System (INIS)

Peterman, Zell E.; Cloke, Paul

2001-01-01

In a potential high-level radioactive waste repository at Yucca Mountain, Nevada, radioactive waste and canisters, drip shields protecting the waste from seepage and from rock falls, the backfill and invert material of crushed rock, the host rock, and water and gases contained within pores and fractures in the host rock together would form a complex system commonly referred to as the near-field geochemical environment. Materials introduced into the rock mass with the waste that are designed to prolong containment collectively are referred to as the Engineered Barrier System, and the host rock and its contained water and gases compose the natural system. The interaction of these component parts under highly perturbed conditions including temperatures well above natural ambient temperatures will need to be understood to assess the performance of the potential repository for long-term containment of nuclear waste. The geochemistry and mineralogy of the rock mass hosting the emplacement drifts must be known in order to assess the role of the natural system in the near-field environment. Emplacement drifts in a potential repository at Yucca Mountain would be constructed in the phenocryst-poor member of the Topopah Spring Tuff which is composed of both lithophysal and nonlithophysal zones. The chemical composition of the phenocryst-poor member has been characterized by numerous chemical analyses of outcrop samples and of core samples obtained by surface-based drilling. Those analyses have shown that the phenocryst-poor member of the Topopah Spring Tuff is remarkably uniform in composition both vertically and laterally. To verify this geochemical uniformity and to provide rock analyses of samples obtained directly from the potential repository block, major and trace elements were analyzed in core samples obtained from drill holes in the cross drift, which was driven to provide direct access to the rock mass where emplacement drifts would be constructed

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

International Nuclear Information System (INIS)

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

2009-01-01

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

Seismic response of rock joints and jointed rock mass

International Nuclear Information System (INIS)

Ghosh, A.; Hsiung, S.M.; Chowdhury, A.H.

1996-06-01

Long-term stability of emplacement drifts and potential near-field fluid flow resulting from coupled effects are among the concerns for safe disposal of high-level nuclear waste (HLW). A number of factors can induce drift instability or change the near-field flow patterns. Repetitive seismic loads from earthquakes and thermal loads generated by the decay of emplaced waste are two significant factors. One of two key technical uncertainties (KTU) that can potentially pose a high risk of noncompliance with the performance objectives of 10 CFR Part 60 is the prediction of thermal-mechanical (including repetitive seismic load) effects on stability of emplacement drifts and the engineered barrier system. The second KTU of concern is the prediction of thermal-mechanical-hydrological (including repetitive seismic load) effects on the host rock surrounding the engineered barrier system. The Rock Mechanics research project being conducted at the Center for Nuclear Waste Regulatory Analyses (CNWRA) is intended to address certain specific technical issues associated with these two KTUs. This research project has two major components: (i) seismic response of rock joints and a jointed rock mass and (ii) coupled thermal-mechanical-hydrological (TMH) response of a jointed rock mass surrounding the engineered barrier system (EBS). This final report summarizes the research activities concerned with the repetitive seismic load aspect of both these KTUs

The disposal of Canada`s nuclear fuel waste: a study of postclosure safety of in-room emplacement of used CANDU fuel in copper containers in permeable plutonic rock volume 1: summary

Energy Technology Data Exchange (ETDEWEB)

Wikjord, A G; Baumgartner, P; Johnson, L H; Stanchell, F W; Zach, R; Goodwin, B W

1996-06-01

The concept for disposal of Canada`s nuclear fuel waste involves isolating the waste in corrosion-resistant containers emplaced and sealed within a vault at a depth of 500 to 1000 m in plutonic rock of the Canadian Shield. The case for the acceptability of the concept as a means of safely disposing of Canada`s nuclear fuel waste is presented in an Environmental Impact Statement (EIS) The disposal concept permits a choice of methods, materials, site locations and designs. The EIS presents a case study of the long-term (i.e., postclosure) performance of a hypothetical implementation of the concept, referred to in this report as the reference disposal system. The reference disposal system is based on borehole emplacement of used CANDU fuel in Grade-2 titanium alloy containers in low-permeability, sparsely fractured plutonic rock of the Canadian Shield. We evaluate the long-term performance of another hypothetical implementation of the concept based on in-room emplacement of used CANDU fuel in copper containers in permeable plutonic rock. The geological characteristics of the geosphere assumed for this study result in short groundwater travel times from the disposal vault to the surface. In the present study, the principal barrier to the movement of contaminants is the long-lasting copper container. We show that the long-lasting container can effectively compensate for a permeable host rock which results in an unfavourable groundwater flow condition. These studies illustrate the flexibility of AECL`s disposal concept to take advantage of the retention, delay, dispersion, dilution and radioactive decay of contaminants in a system of natural barriers provided by the geosphere and hydrosphere and of engineered barriers provided by the waste form, container, buffer, backfills, other vault seals and grouts. In an actual implementation, the engineered system would be designed for the geological conditions encountered at the host site. 34 refs., 2 tabs., 11 figs.

Post emplacement environment of waste packages

International Nuclear Information System (INIS)

Knauss, K.G.; Oversby, V.M.; Wolery, T.J.

1983-01-01

Experiments have been conducted as part of the Nevada Nuclear Waste Storage Investigations Project to determine the changes in water chemistry due to reaction of the Topopah Spring tuff with natural groundwater at temperatures up to 150 0 C. The reaction extent has been investigated as a function of rock-to-water ratio, temperature, reaction time, physical state of the samples, and geographic location of the samples within the tuff unit. Results of these experiments will be used to provide information on the water chemistry to be expected if a high-level waste repository were to be constructed in the Topopah Spring tuff. 6 references, 5 figures, 1 table

Waste disposal developments within BNFL

International Nuclear Information System (INIS)

Johnson, L.F.

1989-01-01

British Nuclear Fuels plc has broad involvement in topics of radioactive waste generation, treatment, storage and disposal. The Company's site at Drigg has been used since 1959 for the disposal of low level waste and its facilities are now being upgraded and extended for that purpose. Since September 1987, BNFL on behalf of UK Nirex Limited has been managing an investigation of the Sellafield area to assess its suitability for deep underground emplacement of low and intermediate level radioactive wastes. An approach will be described to establish a partnership with the local community to work towards a concept of monitored, underground emplacement appropriate for each waste category. (author)

Mechanical Assessment of the Drip Shield Subject to Vibratory Motion and Dynamic and Static Rock Loading

International Nuclear Information System (INIS)

R.C. Quittmeyer

2005-01-01

The purpose of the drip shield (DS) is to divert water that may seep into emplacement drifts from contacting the waste packages, and to protect the waste packages from impact or static loading from rockfall. The objective of this document is to summarize, into one location, the results of a series of supporting engineering calculations that were developed to study the effect of static and dynamic loads on the mechanical performance of the DS. The potential DS loads are a result of: (1) Potential earthquake vibratory ground motion, and resulting interaction of the DS, waste package and pallet, and drift invert; (2) Dynamic impacts of rockfall resulting from emplacement drift damage as a result of earthquake vibratory motion; and (3) Static load of the caved rock rubble that may come to rest on the DS as a result of vibratory motion or from time-dependent yielding of the rock mass surrounding the emplacement drift. The potential mechanical failure mechanisms that may result from these loads include: (1) Overturning and/or separation of the interlocking DS segments; (2) Loss of structural integrity and stability of the DS, including excessive deformation or buckling; and (3) Localized damage to the top and side-wall plates of the DS. The scope of this document is limited to summarizing results presented in the supporting calculations in the areas of analysis of the potential for DS collapse, and determination of the damaged surface area of the DS plates. New calculations are presented to determine whether or not separation of DSs occur under vibratory motion

Performance Assessments of Generic Nuclear Waste Repositories in Shale

Science.gov (United States)

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

2017-12-01

Simulations of deep geologic disposal of nuclear waste in a generic shale formation showcase Geologic Disposal Safety Assessment (GDSA) Framework, a toolkit for repository performance assessment (PA) whose capabilities include domain discretization (Cubit), multiphysics simulations (PFLOTRAN), uncertainty and sensitivity analysis (Dakota), and visualization (Paraview). GDSA Framework is used to conduct PAs of two generic repositories in shale. The first considers the disposal of 22,000 metric tons heavy metal of commercial spent nuclear fuel. The second considers disposal of defense-related spent nuclear fuel and high level waste. Each PA accounts for the thermal load and radionuclide inventory of applicable waste types, components of the engineered barrier system, and components of the natural barrier system including the host rock shale and underlying and overlying stratigraphic units. Model domains are half-symmetry, gridded with Cubit, and contain between 7 and 22 million grid cells. Grid refinement captures the detail of individual waste packages, emplacement drifts, access drifts, and shafts. Simulations are run in a high performance computing environment on as many as 2048 processes. Equations describing coupled heat and fluid flow and reactive transport are solved with PFLOTRAN, an open-source, massively parallel multiphase flow and reactive transport code. Additional simulated processes include waste package degradation, waste form dissolution, radioactive decay and ingrowth, sorption, solubility, advection, dispersion, and diffusion. Simulations are run to 106 y, and radionuclide concentrations are observed within aquifers at a point approximately 5 km downgradient of the repository. Dakota is used to sample likely ranges of input parameters including waste form and waste package degradation rates and properties of engineered and natural materials to quantify uncertainty in predicted concentrations and sensitivity to input parameters. Sandia National

Geological and geotechnical limitations of radioactive waste retrievability in geologic disposals

Energy Technology Data Exchange (ETDEWEB)

Stahlmann, Joachim; Leon-Vargas, Rocio; Mintzlaff, Volker; Treidler, Ann-Kathrin [TU Braunschweig (Germany). Inst. for Soil Mechanics and Foundation Engineering

2015-07-01

The capability of retrieving radioactive waste emplaced in deep geological formations is nowadays in discussion in many countries. Based on the storage of high-level radioactive waste (HAW) in deep geological repositories there is a number of possible scenarios for their retrieval. Measurements for an improved retrieving capability may impact on the geotechnical and geological barriers, e.g. keeping open the access drifts for a long period of time can result in a bigger evacuation damage zone (EDZ) in the host rock which implies potential flow paths for ground water. Nevertheless, to limit the possible scenarios associated to the retrieval implementation, it is necessary to take in consideration which criteria will be used for an efficient monitoring program, while clearly determining the performance reliability of the geotechnical barriers. In addition, the integrity of the host rock as geological barrier has to be verified. Therefore, it is important to evaluate different design solutions and the most appropriate measurement methods to improve the retrievability process of wastes from a geological repository. A short presentation of the host rocks is given is this paper.

Emplacement engineering

Energy Technology Data Exchange (ETDEWEB)

Hill, Ernest E [Lawrence Radiation Laboratory, University of California, Livermore, CA (United States)

1970-05-01

Emplacement Engineering can be defined as that portion of a nuclear explosive project that is concerned with the emplacement of the explosive. This definition would then include virtually everything except the design and fabrication of the explosive and the post-shot-effects program. For future commercial application, the post-shot-effects program will essentially disappear. This emplacement portion of a nuclear explosive project constitutes a large fraction of the total project cost, but it has largely been overshadowed by the explosive and explosive-effects portions. As we move into commercial applications. Emplacement Engineering must receive more attention from both industry and government. To place emplacement costs in their proper relationship with total projects costs, we have performed a study of commercial underground nuclear explosive applications such as gas stimulation. Although there are many intangibles in such a study, we have been able to at least obtain some feel for the relative fractional costs of the non-explosive costs compared with the explosive costs. This study involved estimating the cost elements for applications using a single explosive at 5,000 ft, 10,000 ft, and 15,000 ft. For each depth, the cost estimates were made for a range of emplacement hole and explosive diameters. Results of these estimates for explosive-related costs, hole-related costs, and total costs are shown for the three depths. Note that the explosive package outside diameter is assumed as 2 inches less than the hole (or casing) inside diameter for all cases. For the 5,000-ft application the explosive-related costs dominate, and of particular importance is the indicated diameter for minimum total cost which occurs at approximately a 17.5-in. hole (15.5-in. explosive). The hole-related costs are in 'the same range as the explosive-related costs for the 10,000-ft application. For this case, the minimum total cost occurs at approximately a 14-in. hole (12-in. explosive

Performance confirmation data acquisition system

International Nuclear Information System (INIS)

McAffee, D.A.; Raczka, N.T.

1997-01-01

As part of the Viability Assessment (VA) work, this QAP-3-9 document presents and evaluates a comprehensive set of viable concepts for collecting Performance Confirmation (PC) related data. The concepts include: monitoring subsurface repository air temperatures, humidity levels and gaseous emissions via the subsurface ventilation systems, and monitoring the repository geo-technical parameters and rock mass from bore-holes located along the perimeter main drifts and throughout a series of human-rated Observation Drifts to be located in a plane 25 meters above the plane of the emplacement drifts. A key element of this document is the development and analysis of a purposed multi-purpose Remote Inspection Gantry that would provide direct, real-time visual, thermal, and radiological monitoring of conditions inside operational emplacement drifts and close-up observations of in-situ Waste Packages. Preliminary finite-element analyses are presented that indicate the technological feasibility of operating an inspection gantry inside the operational emplacement drifts for short inspection missions lasting 2--3 hours. Overall reliability, availability, and maintainability of the PC data collection concepts are discussed. Preliminary concepts for PC data collection network are also provided

Performance confirmation data acquisition system

Energy Technology Data Exchange (ETDEWEB)

McAffee, D.A.; Raczka, N.T. [Yucca Mountain Project, Las Vegas, NV (United States)

1997-12-31

As part of the Viability Assessment (VA) work, this QAP-3-9 document presents and evaluates a comprehensive set of viable concepts for collecting Performance Confirmation (PC) related data. The concepts include: monitoring subsurface repository air temperatures, humidity levels and gaseous emissions via the subsurface ventilation systems, and monitoring the repository geo-technical parameters and rock mass from bore-holes located along the perimeter main drifts and throughout a series of human-rated Observation Drifts to be located in a plane 25 meters above the plane of the emplacement drifts. A key element of this document is the development and analysis of a purposed multi-purpose Remote Inspection Gantry that would provide direct, real-time visual, thermal, and radiological monitoring of conditions inside operational emplacement drifts and close-up observations of in-situ Waste Packages. Preliminary finite-element analyses are presented that indicate the technological feasibility of operating an inspection gantry inside the operational emplacement drifts for short inspection missions lasting 2--3 hours. Overall reliability, availability, and maintainability of the PC data collection concepts are discussed. Preliminary concepts for PC data collection network are also provided.

Analysis of stresses on the 1st phase support of the monitoring drifts of the radioactive waste repository

Directory of Open Access Journals (Sweden)

Hatala Jozef

1999-09-01

Full Text Available In the paper, the stability analysis of the radioactive waste repository monitoring drifts’ support by means of the numerical modelling - finite element method is described. The aim of this analysis was to judge to what extent the designed 1st phase support’s parameters correspond with the geomechanical conditions determined by the engineering-geological survey.

Geological repository layout for radioactive high level long lived waste in argillite

International Nuclear Information System (INIS)

Gaussen, JL

2006-01-01

In the framework of the 1991 French radioactive waste act, ANDRA has studied the feasibility of a geological repository in the argillite layer of the Bure site for high level long lived waste. This presentation is focussed on the underground facilities which constitute the specific component of this project. The preliminary underground layout which has been elaborated is based on four categories of data: - the waste characteristics and inventory; - the geological properties of the host argillite; - the long term performance objectives of the repository; - the specifications in terms of operation and reversibility. The underground facilities consist of two types of works: the access works (shafts and drifts) and the disposal cells. The function of the access works is to permit the implementation of two concurrent activities: the nuclear operations (transfer and emplacement of the disposal packages into the disposal cells) and the construction of the next disposal cells. The design of the drifts network which matches up to this function is also influenced by two other specifications: the minimization of the drift dimensions in order to limit their influence on the integrity of the geological formation and the necessity of a safe ventilation in case of fire. The resulting layout is a network of 4 parallel drifts (2 of them being dedicated to the operation, the other two being dedicated to the construction activities). The average diameter of these access drifts is 7 meters. The link between the surface and the underground is ensured by 4 shafts. The most important function of the disposal cells is to contribute to the long term performance of the repository. In this regard, the thermal and geotechnical considerations play an important role. The B wastes (intermediate level wastes) are not (or not very) exothermic. Consequently, the design of their disposal cells result mainly from geotechnical considerations. The disposal packages (made of concrete) are piled up in

Remote excavation using the telerobotic small emplacement excavator

International Nuclear Information System (INIS)

Thompson, D.H.; Burks, B.L.; Killough, S.M.

1993-01-01

Oak Ridge National Laboratory is developing remote excavation technologies for the Office of Technology Development, Robotics Technology Development Program. This work is being done to meet the need for remote excavation and removal of radioactive and contaminated buried waste at several DOE sites. System requirements are based on the need to uncover and remove waste from burial sites in a way that does not cause unnecessary personnel exposure or additional environmental contamination. Goals for the current project are to demonstrate dexterous control of a backhoe with force feedback and to implement robotic operations that will improve productivity. The Telerobotic Small Emplacement Excavator is a prototype system that incorporates the needed robotic and telerobotic capabilities on a commercially available platform. The ability to add remote dexterous teleoperation and robotic operating modes is intended to be adaptable to other commercially available excavator systems

Thermal analysis of NNWSI conceptual waste package designs

International Nuclear Information System (INIS)

Stein, W.; Hockman, J.N.; O'Neal, W.C.

1984-04-01

Lawrence Livermore National Laboratory is involved in the design and testing of high-level nuclear waste packages. Many of the aspects of waste package design and testing (e.g., corrosion and leaching) depend in part on the temperature history of the emplaced packages. This report discusses thermal modeling and analysis of various emplaced waste package conceptual designs including the models used, the assumptions and approximations made, and the results obtained. 16 references

VENTILATION MODEL

International Nuclear Information System (INIS)

V. Chipman

2002-01-01

The purpose of the Ventilation Model is to simulate the heat transfer processes in and around waste emplacement drifts during periods of forced ventilation. The model evaluates the effects of emplacement drift ventilation on the thermal conditions in the emplacement drifts and surrounding rock mass, and calculates the heat removal by ventilation as a measure of the viability of ventilation to delay the onset of peak repository temperature and reduce its magnitude. The heat removal by ventilation is temporally and spatially dependent, and is expressed as the fraction of heat carried away by the ventilation air compared to the fraction of heat produced by radionuclide decay. One minus the heat removal is called the wall heat fraction, or the remaining amount of heat that is transferred via conduction to the surrounding rock mass. Downstream models, such as the ''Multiscale Thermohydrologic Model'' (BSC 2001), use the wall heat fractions as outputted from the Ventilation Model to initialize their postclosure analyses

Retrieval system for emplaced spent unreprocessed fuel (SURF) in salt bed depository. Baseline concept criteria specifications and mechanical failure probabilities

International Nuclear Information System (INIS)

Hudson, E.E.; McCleery, J.E.

1979-05-01

One of the integral elements of the Nuclear Waste Management Program is the material handling task of retrieving Canisters containing spent unreprocessed fuel from their emplacement in a deep geologic salt bed Depository. A study of the retrieval concept data base predicated this report. In this report, alternative concepts for the tasks are illustrated and critiqued, a baseline concept in scenario form is derived and basic retrieval subsystem specifications are presented with cyclic failure probabilities predicted. The report is based on the following assumptions: (a) during retrieval, a temporary radiation seal is placed over each Canister emplacement; (b) a sleeve, surrounding the Canister, was initially installed during the original emplacement; (c) the emplacement room's physical and environmental conditions established in this report are maintained while the task is performed

Emplacement of Columbia River flood basalt

Science.gov (United States)

Reidel, Stephen P.

1998-11-01

Evidence is examined for the emplacement of the Umatilla, Wilbur Creek, and the Asotin Members of Columbia River Basalt Group. These flows erupted in the eastern part of the Columbia Plateau during the waning phases of volcanism. The Umatilla Member consists of two flows in the Lewiston basin area and southwestern Columbia Plateau. These flows mixed to form one flow in the central Columbia Plateau. The composition of the younger flow is preserved in the center and the composition of the older flow is at the top and bottom. There is a complete gradation between the two. Flows of the Wilbur Creek and Asotin Members erupted individually in the eastern Columbia Plateau and also mixed together in the central Columbia Plateau. Comparison of the emplacement patterns to intraflow structures and textures of the flows suggests that very little time elapsed between eruptions. In addition, the amount of crust that formed on the earlier flows prior to mixing also suggests rapid emplacement. Calculations of volumetric flow rates through constrictions in channels suggest emplacement times of weeks to months under fast laminar flow for all three members. A new model for the emplacement of Columbia River Basalt Group flows is proposed that suggests rapid eruption and emplacement for the main part of the flow and slower emplacement along the margins as the of the flow margin expands.

The full-scale Emplacement (FE) Experiment at the Mont Terri URL

International Nuclear Information System (INIS)

Mueller, H.R.; Weber, H.P.; Koehler, S.; Vogt, T.; Vietor, T.

2012-01-01

Document available in extended abstract form only. The Full-Scale Emplacement (FE) Experiment at the Mont Terri underground research laboratory (URL) is a full-scale heater test in a clay-rich formation. It simulates the construction, waste emplacement and backfilling of a spent fuel (SF) / vitrified high-level waste (HLW) repository tunnel as realistically as possible. The entire experiment implementation as well as the post-closure THM(C) evolution will be monitored using several hundred sensors. These are distributed in the host rock in the near- and far-field, the tunnel lining, the engineered barrier system and on the heaters. The aim of this experiment is to investigate HLW repository-induced thermo-hydro-mechanical (THM) coupled effects on the host rock and the validation of existing coupled THM models. A further aim is the verification of the technical feasibility of constructing a 50 m repository section at full scale with all relevant components using standard industrial equipment. Finally, the experiment will demonstrate the canister and buffer emplacement procedures for underground conditions based on the Swiss disposal concept. Experimental layout The FE experiment is based on the Swiss disposal concept for SF / HLW. The 50 m long test gallery, at the end of the former MB test tunnel in the Mont Terri URL, will be realised with a diameter of approx. 3 m. In the experiment gallery, 3 heaters with dimensions similar to those of waste canisters will be emplaced on top of abutments built of bentonite blocks. The remaining space will be backfilled with compacted bentonite pellets. The experiment will be sealed off towards the start niche with a concrete plug holding the buffer in place and reducing air and water fluxes. The first scoping calculations and design modelling for the 'far-field' instrumentation have been completed; these works have been carried out using CodeBRIGHT and the multiphase flow simulator TOUGH2. With an initial heat output of 1500 W

PROBABILISTIC ANALYSES OF WASTE PACKAGE QUANTITIES IMPACTED BY POTENTIAL IGNEOUS DISRUPTION AT YUCCA MOUNTAIN

International Nuclear Information System (INIS)

M.G. Wallace

2005-01-01

A probabilistic analysis was conducted to estimate ranges for the numbers of waste packages that could be damaged in a potential future igneous event through a repository at Yucca Mountain. The analyses include disruption from an intrusive igneous event and from an extrusive volcanic event. This analysis supports the evaluation of the potential consequences of future igneous activity as part of the total system performance assessment for the license application for the Yucca Mountain Project (YMP). The first scenario, igneous intrusion, investigated the case where one or more igneous dikes intersect the repository. A swarm of dikes was characterized by distributions of length, width, azimuth, and number of dikes and the spacings between them. Through the use in part of a latin hypercube simulator and a modified video game engine, mathematical relationships were built between those parameters and the number of waste packages hit. Corresponding cumulative distribution function curves (CDFs) for the number of waste packages hit under several different scenarios were calculated. Variations in dike thickness ranges, as well as in repository magma bulkhead positions were examined through sensitivity studies. It was assumed that all waste packages in an emplacement drift would be impacted if that drift were intersected by a dike. Over 10,000 individual simulations were performed. Based on these calculations, out of a total of over 11,000 planned waste packages distributed over an area of approximately 5.5 km 2 , the median number of waste packages impacted was roughly 1/10 of the total. Individual cases ranged from 0 waste packages to the entire inventory being impacted. The igneous intrusion analysis involved an explicit characterization of dike-drift intersections, built upon various distributions that reflect the uncertainties associated with the inputs. The second igneous scenario, volcanic eruption (eruptive conduits), considered the effects of conduits formed in

Robotics and remote handling concepts for disposal of high-level nuclear waste

International Nuclear Information System (INIS)

McAffee, Douglas; Raczka, Norman; Schwartztrauber, Keith

1997-01-01

This paper summarizes preliminary remote handling and robotic concepts being developed as part of the US Department of Energy's (DOE) Yucca Mountain Project. The DOE is currently evaluating the Yucca Mountain Nevada site for suitability as a possible underground geologic repository for the disposal of high level nuclear waste. The current advanced conceptual design calls for the disposal of more than 12,000 high level nuclear waste packages within a 225 km underground network of tunnels and emplacement drifts. Many of the waste packages may weigh as much as 66 tonnes and measure 1.8 m in diameter and 5.6 m long. The waste packages will emit significant levels of radiation and heat. Therefore, remote handling is a cornerstone of the repository design and operating concepts. This paper discusses potential applications areas for robotics and remote handling technologies within the subsurface repository. It also summarizes the findings of a preliminary technology survey which reviewed available robotic and remote handling technologies developed within the nuclear, mining, rail and industrial robotics and automation industries, and at national laboratories, universities, and related research institutions and government agencies

Implementation of the full-scale emplacement (FE) experiment at the Mont Terri rock laboratory

Energy Technology Data Exchange (ETDEWEB)

Müller, H.R.; Garitte, B.; Vogt, T.; and others

2017-04-15

Opalinus Clay is currently being assessed as the host rock for a deep geological repository for high-level and low- and intermediate-level radioactive wastes in Switzerland. Within this framework, the 'Full-Scale Emplacement' (FE) experiment was initiated at the Mont Terri rock laboratory close to the small town of St-Ursanne in Switzerland. The FE experiment simulates, as realistically as possible, the construction, waste emplacement, backfilling and early post-closure evolution of a spent fuel/vitrified high-level waste disposal tunnel according to the Swiss repository concept. The main aim of this multiple heater test is the investigation of repository-induced thermo-hydro-mechanical (THM) coupled effects on the host rock at this scale and the validation of existing coupled THM models. For this, several hundred sensors were installed in the rock, the tunnel lining, the bentonite buffer, the heaters and the plug. This paper is structured according to the implementation timeline of the FE experiment. It documents relevant details about the instrumentation, the tunnel construction, the production of the bentonite blocks and the highly compacted 'granulated bentonite mixture' (GBM), the development and construction of the prototype 'backfilling machine' (BFM) and its testing for horizontal GBM emplacement. Finally, the plug construction and the start of all 3 heaters (with a thermal output of 1350 Watt each) in February 2015 are briefly described. In this paper, measurement results representative of the different experimental steps are also presented. Tunnel construction aspects are discussed on the basis of tunnel wall displacements, permeability testing and relative humidity measurements around the tunnel. GBM densities achieved with the BFM in the different off-site mock-up tests and, finally, in the FE tunnel are presented. Finally, in situ thermal conductivity and temperature measurements recorded during the first heating months

Investigation of stress in a circular tunnel due to overburden and thermal loading of horizontally placed 21 PWR multi purpose canisters

International Nuclear Information System (INIS)

Kandalaft-Ladkany, N.; Wyman, R.V.

1994-01-01

The drift of a High Level Nuclear Waste (HLNW) Repository were subjected to 2-D thermal loading resulting from the horizontal emplacement of 125 Ton Multi-Purpose Canisters (MPC). Ten 2-D temperature profiles, resulting from 57 Kw/acre and 114 Kw/acre thermal loading conditions, were used in a finite element analysis of the drift; in which a quadrant of the drift and surrounding rock ±100m above and below the drift were modeled. Our analysis shows that the 114 Kw/acre thermal loading results in compressive stresses around the drift, 60 years after emplacement, that exceed the unconfined compressive strength of the TSW tuff analyzed. Stresses resulting from a 57 Kw/acre thermal loading are within the acceptable limit in tunnel rock. A parametric analysis of the invert backfill material showed that Young's modulus for the invert backfill should closely match that of the surrounding unconfined rock in the tunnel in order to prevent an unacceptable stress rise in both rock and backfill

Pre-construction geologic section along the cross drift through the potential high-level radioactive waste repository, Yucca Mountain, Nye County, Nevada

International Nuclear Information System (INIS)

Potter, C.J.; Day, W.C.; Sweetkind, D.S.; Juan, C.S.; Drake, R.M. II

1998-01-01

As part of the Site Characterization effort for the US Department of Energy's Yucca Mountain Project, tunnels excavated by tunnel boring machines provide access to the volume of rock that is under consideration for possible underground storage of high-level nuclear waste beneath Yucca Mountain, Nevada. The Exploratory Studies Facility, a 7.8-km-long, 7.6-m-diameter tunnel, has been excavated, and a 2.8-km-long, 5-m-diameter Cross Drift will be excavated in 1998 as part of the geologic, hydrologic and geotechnical evaluation of the potential repository. The southwest-trending Cross Drift branches off of the north ramp of the horseshoe-shaped Exploratory Studies Facility. This report summarizes an interpretive geologic section that was prepared for the Yucca Mountain Project as a tool for use in the design and construction of the Cross Drift

Potential migration of buoyant LNAPL from intermediate level waste (ILW) emplaced in a geological disposal facility (GDF) for U.K. radioactive waste.

Science.gov (United States)

Benbow, Steven J; Rivett, Michael O; Chittenden, Neil; Herbert, Alan W; Watson, Sarah; Williams, Steve J; Norris, Simon

2014-10-15

A safety case for the disposal of Intermediate Level (radioactive) Waste (ILW) in a deep geological disposal facility (GDF) requires consideration of the potential for waste-derived light non-aqueous phase liquid (LNAPL) to migrate under positive buoyancy from disposed waste packages. Were entrainment of waste-derived radionuclides in LNAPL to occur, such migration could result in a shorter overall travel time to environmental or human receptors than radionuclide migration solely associated with the movement of groundwater. This paper provides a contribution to the assessment of this issue through multiphase-flow numerical modelling underpinned by a review of the UK's ILW inventory and literature to define the nature of the associated ILW LNAPL source term. Examination has been at the waste package-local GDF environment scale to determine whether proposed disposal of ILW would lead to significant likelihood of LNAPL migration, both from waste packages and from a GDF vault into the local host rock. Our review and numerical modelling support the proposition that the release of a discrete free phase LNAPL from ILW would not present a significant challenge to the safety case even with conservative approximations. 'As-disposed' LNAPL emplaced with the waste is not expected to pose a significant issue. 'Secondary LNAPL' generated in situ within the disposed ILW, arising from the decomposition of plastics, in particular PVC (polyvinyl chloride), could form the predominant LNAPL source term. Released high molecular weight phthalate plasticizers are judged to be the primary LNAPL potentially generated. These are expected to have low buoyancy-based mobility due to their very low density contrast with water and high viscosity. Due to the inherent uncertainties, significant conservatisms were adopted within the numerical modelling approach, including: the simulation of a deliberately high organic material--PVC content wastestream (2D03) within an annular grouted waste package

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

CROSS DRIFT ALCOVE/NICHE UTILITIES ANALYSIS

International Nuclear Information System (INIS)

S. Goodin

1999-01-01

The purpose of this analysis is to provide the design basis and general arrangement requirements of the non-potable water, waste water, compressed air and ventilation (post excavation) utilities required in support of the Cross Drift alcoves and niches

Waste Isolation Pilot Plant (WIPP) research and development program: in situ testing plan, March 1982

International Nuclear Information System (INIS)

Matalucci, R.V.; Christensen, C.L.; Hunter, T.O.; Molecke, M.A.; Munson, D.E.

1982-12-01

The WIPP in southeast New Mexico is being developed as an R and D facility to demonstrate the safe disposal of radioactive defense wastes in bedded salt. The tests are done first without radioactive materials and then with transuranic (TRU) waste and Defense High-Level Waste (DHLW). The thermal/structural itneraction experiments include (a) geomechanical evaluations of access drifts, vertical shafts, and isothermal TRU disposal rooms during the Site and Preliminary Validation Program, (b) tests that represent the reference DHLW room configuraton (5.5 m x 5.5 m) and areal thermal loading of 12 W/m 2 , (c) an overtest of the DHLW congfiguration heated to about four times the reference thermal loading; (d) geomechanical evaluations of various room widths up to 9.1 m, variable pillar widths, and a long-drift intersection, (e) an 11-m-dia axisymmetric heated pillar test, and (f) miscellaneous tests to determine stress field and clay seam sliding resistance. The plugging and sealing experiments include (a) salt permeability tests, (b) tests to determine effects of size and scale on behavior of plugs and to determine backfill material behavior and emplacement techniques, and (c) a plug test matrix to evaluate candidate sealing materials. Waste package interaction experiments include (a) simulated-waste package tests that use several design options and engineered barrier materials under reference and accelerated DHLW environments, (b) confirmatory brine migration tests, (c) TRU drum durability tests in dry and wet conditions, (d) options for radiation-source tests using cesium capsules, and (e) actual DHLW tests using up to 40 canisters for technical demonstrations and for addressing concerns of wasteform chemistry, leaching, and near-field radionuclide migration

Management of remote-handled defense transuranic wastes

International Nuclear Information System (INIS)

Ebra, M.A.; Pierce, G.D.; Carson, P.H.

1988-01-01

Transuranic (TRU) wastes generated by defense-related activities are scheduled for emplacement at the Waste Isolation Pilot Plant (WIPP) in New Mexico beginning in October 1988. After five years of operation as a research and development facility, the WIPP may be designated as a permanent repository for these wastes, if it has been demonstrated that this deep, geologically stable formation is a safe disposal option. Defense TRU wastes are currently stored at various Department of Energy (DOE) sites across the nation. Approximately 2% by volume of currently stored TRU wastes are defined, on the basis of dose rates, as remote-handled (RH). RH wastes continue to be generated at various locations operated by DOE contractors. They require special handling and processing prior to and during emplacement in the WIPP. This paper describes the strategy for managing defense RH TRU wastes

Nuclear hazardous waste cost control management

International Nuclear Information System (INIS)

Selg, R.A.

1991-01-01

The effects of the waste content of glass waste forms on Savannah River high-level waste disposal costs are currently under study to adjust the glass frit content to optimize the glass waste loadings and therefore significantly reduce the overall waste disposal cost. Changes in waste content affect onsite Defense Waste Changes in waste contents affect onsite Defense Waste Processing Facility (DWPF) costs as well as offsite shipping and repository emplacement charges. A nominal 1% increase over the 28 wt% waste loading of DWPF glass would reduce disposal costs by about $50 million for Savannah River wastes generated to the year 2000. Optimization of the glass waste forms to be produced in the SWPF is being supported by economic evaluations of the impact of the forms on waste disposal costs. Glass compositions are specified for acceptable melt processing and durability characteristics, with economic effects tracked by the number of waste canisters produced. This paper presents an evaluation of the effects of variations in waste content of the glass waste forms on the overall cost of the disposal, including offsite shipment and repository emplacement, of the Savannah River high-level wastes

Wire-rope emplacement of diagnostics systems

International Nuclear Information System (INIS)

Burden, W.L.

1982-01-01

The study reported here was initiated to determine if, with the Cable Downhole System (CDS) currently under development, there is an advantage to using continuous wire rope to lower the emplacement package to the bottom of the hole. A baseline design using two wire ropes as well as several alternatives are discussed in this report. It was concluded that the advantages of the wire-rope emplacement system do not justify the cost of converting to such a system, especially for LLNL's maximum emplacement package weights

Engineered Barrier System: Physical and Chemical Environment Model

International Nuclear Information System (INIS)

Jolley, D. M.; Jarek, R.; Mariner, P.

2004-01-01

The conceptual and predictive models documented in this Engineered Barrier System: Physical and Chemical Environment Model report describe the evolution of the physical and chemical conditions within the waste emplacement drifts of the repository. The modeling approaches and model output data will be used in the total system performance assessment (TSPA-LA) to assess the performance of the engineered barrier system and the waste form. These models evaluate the range of potential water compositions within the emplacement drifts, resulting from the interaction of introduced materials and minerals in dust with water seeping into the drifts and with aqueous solutions forming by deliquescence of dust (as influenced by atmospheric conditions), and from thermal-hydrological-chemical (THC) processes in the drift. These models also consider the uncertainty and variability in water chemistry inside the drift and the compositions of introduced materials within the drift. This report develops and documents a set of process- and abstraction-level models that constitute the engineered barrier system: physical and chemical environment model. Where possible, these models use information directly from other process model reports as input, which promotes integration among process models used for total system performance assessment. Specific tasks and activities of modeling the physical and chemical environment are included in the technical work plan ''Technical Work Plan for: In-Drift Geochemistry Modeling'' (BSC 2004 [DIRS 166519]). As described in the technical work plan, the development of this report is coordinated with the development of other engineered barrier system analysis model reports

Engineered Barrier System: Physical and Chemical Environment

International Nuclear Information System (INIS)

Dixon, P.

2004-01-01

The conceptual and predictive models documented in this Engineered Barrier System: Physical and Chemical Environment Model report describe the evolution of the physical and chemical conditions within the waste emplacement drifts of the repository. The modeling approaches and model output data will be used in the total system performance assessment (TSPA-LA) to assess the performance of the engineered barrier system and the waste form. These models evaluate the range of potential water compositions within the emplacement drifts, resulting from the interaction of introduced materials and minerals in dust with water seeping into the drifts and with aqueous solutions forming by deliquescence of dust (as influenced by atmospheric conditions), and from thermal-hydrological-chemical (THC) processes in the drift. These models also consider the uncertainty and variability in water chemistry inside the drift and the compositions of introduced materials within the drift. This report develops and documents a set of process- and abstraction-level models that constitute the engineered barrier system: physical and chemical environment model. Where possible, these models use information directly from other process model reports as input, which promotes integration among process models used for total system performance assessment. Specific tasks and activities of modeling the physical and chemical environment are included in the technical work plan ''Technical Work Plan for: In-Drift Geochemistry Modeling'' (BSC 2004 [DIRS 166519]). As described in the technical work plan, the development of this report is coordinated with the development of other engineered barrier system analysis model reports

High level waste canister emplacement and retrieval concepts study

International Nuclear Information System (INIS)

1975-09-01

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

Environment on the Surfaces of the Drip Shield and Waste Package Outer Barrier

Energy Technology Data Exchange (ETDEWEB)

T. Wolery

2005-02-22

This report provides supporting analysis of the conditions at which an aqueous solution can exist on the drip shield or waste package surfaces, including theoretical underpinning for the evolution of concentrated brines that could form by deliquescence or evaporation, and evaluation of the effects of acid-gas generation on brine composition. This analysis does not directly feed the total system performance assessment for the license application (TSPA-LA), but supports modeling and abstraction of the in-drift chemical environment (BSC 2004 [DIRS 169863]; BSC 2004 [DIRS 169860]). It also provides analyses that may support screening of features, events, and processes, and input for response to regulatory inquiries. This report emphasizes conditions of low relative humidity (RH) that, depending on temperature and chemical conditions, may be dry or may be associated with an aqueous phase containing concentrated electrolytes. Concentrated solutions at low RH may evolve by evaporative concentration of water that seeps into emplacement drifts, or by deliquescence of dust on the waste package or drip shield surfaces. The minimum RH for occurrence of aqueous conditions is calculated for various chemical systems based on current understanding of site geochemistry and equilibrium thermodynamics. The analysis makes use of known characteristics of Yucca Mountain waters and dust from existing tunnels, laboratory data, and relevant information from the technical literature and handbooks.

Environment on the Surfaces of the Drip Shield and Waste Package Outer Barrier

International Nuclear Information System (INIS)

T. Wolery

2005-01-01

This report provides supporting analysis of the conditions at which an aqueous solution can exist on the drip shield or waste package surfaces, including theoretical underpinning for the evolution of concentrated brines that could form by deliquescence or evaporation, and evaluation of the effects of acid-gas generation on brine composition. This analysis does not directly feed the total system performance assessment for the license application (TSPA-LA), but supports modeling and abstraction of the in-drift chemical environment (BSC 2004 [DIRS 169863]; BSC 2004 [DIRS 169860]). It also provides analyses that may support screening of features, events, and processes, and input for response to regulatory inquiries. This report emphasizes conditions of low relative humidity (RH) that, depending on temperature and chemical conditions, may be dry or may be associated with an aqueous phase containing concentrated electrolytes. Concentrated solutions at low RH may evolve by evaporative concentration of water that seeps into emplacement drifts, or by deliquescence of dust on the waste package or drip shield surfaces. The minimum RH for occurrence of aqueous conditions is calculated for various chemical systems based on current understanding of site geochemistry and equilibrium thermodynamics. The analysis makes use of known characteristics of Yucca Mountain waters and dust from existing tunnels, laboratory data, and relevant information from the technical literature and handbooks

Naval Waste Package Design Report

International Nuclear Information System (INIS)

M.M. Lewis

2004-01-01

A design methodology for the waste packages and ancillary components, viz., the emplacement pallets and drip shields, has been developed to provide designs that satisfy the safety and operational requirements of the Yucca Mountain Project. This methodology is described in the ''Waste Package Design Methodology Report'' Mecham 2004 [DIRS 166168]. To demonstrate the practicability of this design methodology, four waste package design configurations have been selected to illustrate the application of the methodology. These four design configurations are the 21-pressurized water reactor (PWR) Absorber Plate waste package, the 44-boiling water reactor (BWR) waste package, the 5-defense high-level waste (DHLW)/United States (U.S.) Department of Energy (DOE) spent nuclear fuel (SNF) Co-disposal Short waste package, and the Naval Canistered SNF Long waste package. Also included in this demonstration is the emplacement pallet and continuous drip shield. The purpose of this report is to document how that design methodology has been applied to the waste package design configurations intended to accommodate naval canistered SNF. This demonstrates that the design methodology can be applied successfully to this waste package design configuration and support the License Application for construction of the repository

Waste package/repository impact study: Final report

Energy Technology Data Exchange (ETDEWEB)

1985-09-01

The Waste Package/Repository Impact Study was conducted to evaluate the feasibility of using the current reference salt waste package in the salt repository conceptual design. All elements of the repository that may impact waste package parameters, i.e., (size, weight, heat load) were evaluated. The repository elements considered included waste hoist feasibility, transporter and emplacement machine feasibility, subsurface entry dimensions, feasibility of emplacement configuration, and temperature limits. The evaluations are discussed in detail with supplemental technical data included in Appendices to this report, as appropriate. Results and conclusions of the evaluations are discussed in light of the acceptability of the current reference waste package as the basis for salt conceptual design. Finally, recommendations are made relative to the salt project position on the application of the reference waste package as a basis for future design activities. 31 refs., 11 figs., 11 tabs.

Waste package/repository impact study: Final report

International Nuclear Information System (INIS)

1985-09-01

The Waste Package/Repository Impact Study was conducted to evaluate the feasibility of using the current reference salt waste package in the salt repository conceptual design. All elements of the repository that may impact waste package parameters, i.e., (size, weight, heat load) were evaluated. The repository elements considered included waste hoist feasibility, transporter and emplacement machine feasibility, subsurface entry dimensions, feasibility of emplacement configuration, and temperature limits. The evaluations are discussed in detail with supplemental technical data included in Appendices to this report, as appropriate. Results and conclusions of the evaluations are discussed in light of the acceptability of the current reference waste package as the basis for salt conceptual design. Finally, recommendations are made relative to the salt project position on the application of the reference waste package as a basis for future design activities. 31 refs., 11 figs., 11 tabs

The prospective usage of the multi-purpose canister and impacts on the waste management and disposal system

International Nuclear Information System (INIS)

McLeod, N.B.

1993-01-01

The Multi-Purpose Canister (MPC) is designed to be loaded with spent fuel and sealed at reactors and then serve the functions of transport, storage and disposal without reopening. It can be either self-shielded or unshielded, thus requiring compatible overpacks for transport, storage and disposal. The MPC is not a new concept but it may now be viable because of the particular characteristics at Yucca Mountain: larger MPCs are possible because of ramp access to the repository horizon, and the less difficult temperature limits because of in-drift emplacement, rather than borehole emplacement. This paper describes the advantages and disadvantages of adopting the MPC as the principal technology to be employed in the US program. Use of the MPC permits integration of the utility and DOE portions of the system as well as among the elements within the DOE portion. Paradoxically, the principal disadvantage of the MPC is a direct consequence of its merit as an integrating technology. Full integration includes disposability without reopening, and requires that disposability design decisions be made and implemented well in advance of when waste package licensing uncertainties are resolved. There is, therefore, a risk that MPCs loaded prior to waste package licensing will have to be opened. This risk is discussed in terms of probability and consequences and various alternatives for mitigating this risk are discussed

Repository Waste Package Transporter Shielding Weight Optimization

International Nuclear Information System (INIS)

C.E. Sanders; Shiaw-Der Su

2005-01-01

The Yucca Mountain repository requires the use of a waste package (WP) transporter to transport a WP from a process facility on the surface to the subsurface for underground emplacement. The transporter is a part of the waste emplacement transport systems, which includes a primary locomotive at the front end and a secondary locomotive at the rear end. The overall system with a WP on board weights over 350 metric tons (MT). With the shielding mass constituting approximately one-third of the total system weight, shielding optimization for minimal weight will benefit the overall transport system with reduced axle requirements and improved maneuverability. With a high contact dose rate on the WP external surface and minimal personnel shielding afforded by the WP, the transporter provides radiation shielding to workers during waste emplacement and retrieval operations. This paper presents the design approach and optimization method used in achieving a shielding configuration with minimal weight

Thermal modeling of nuclear waste package designs for disposal in tuff

International Nuclear Information System (INIS)

Hockman, J.N.; O'Neal, W.C.

1983-09-01

Lawrence Livermore National Laboratory is involved in the design and testing of high-level nuclear waste packages. Many of the aspects of waste package design and testing (e.g., corrosion and leaching) depend in part on the temperature history of the emplaced packages. This paper discusses thermal modeling and analysis of various emplaced waste package conceptual designs including the models used, the assumptions and approximations made, and the results obtained. 6 references, 6 figures, 3 tables

Thermal modeling of nuclear waste package designs for disposal in tuff

International Nuclear Information System (INIS)

Hockman, J.N.; O'Neal, W.C.

1984-02-01

Lawrence Livermore National Laboratory is involved in the design and testing of high-level nuclear waste packages. Many of the aspects of waste package design and testing (e.g., corrosion and leaching) depend in part on the temperature history of the emplaced packages. This paper discusses thermal modeling and analysis of various emplaced waste package conceptual designs including the models used, the assumptions and approximations made, and the results obtained. 6 references, 6 figures, 4 tables

Deep Borehole Emplacement Mode Hazard Analysis Revision 0

Energy Technology Data Exchange (ETDEWEB)

Sevougian, S. David [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2015-08-07

This letter report outlines a methodology and provides resource information for the Deep Borehole Emplacement Mode Hazard Analysis (DBEMHA). The main purpose is identify the accident hazards and accident event sequences associated with the two emplacement mode options (wireline or drillstring), to outline a methodology for computing accident probabilities and frequencies, and to point to available databases on the nature and frequency of accidents typically associated with standard borehole drilling and nuclear handling operations. Risk mitigation and prevention measures, which have been incorporated into the two emplacement designs (see Cochran and Hardin 2015), are also discussed. A key intent of this report is to provide background information to brief subject matter experts involved in the Emplacement Mode Design Study. [Note: Revision 0 of this report is concentrated more on the wireline emplacement mode. It is expected that Revision 1 will contain further development of the preliminary fault and event trees for the drill string emplacement mode.

WASTE PACKAGE TRANSPORTER DESIGN

International Nuclear Information System (INIS)

Weddle, D.C.; Novotny, R.; Cron, J.

1998-01-01

The purpose of this Design Analysis is to develop preliminary design of the waste package transporter used for waste package (WP) transport and related functions in the subsurface repository. This analysis refines the conceptual design that was started in Phase I of the Viability Assessment. This analysis supports the development of a reliable emplacement concept and a retrieval concept for license application design. The scope of this analysis includes the following activities: (1) Assess features of the transporter design and evaluate alternative design solutions for mechanical components. (2) Develop mechanical equipment details for the transporter. (3) Prepare a preliminary structural evaluation for the transporter. (4) Identify and recommend the equipment design for waste package transport and related functions. (5) Investigate transport equipment interface tolerances. This analysis supports the development of the waste package transporter for the transport, emplacement, and retrieval of packaged radioactive waste forms in the subsurface repository. Once the waste containers are closed and accepted, the packaged radioactive waste forms are termed waste packages (WP). This terminology was finalized as this analysis neared completion; therefore, the term disposal container is used in several references (i.e., the System Description Document (SDD)) (Ref. 5.6). In this analysis and the applicable reference documents, the term ''disposal container'' is synonymous with ''waste package''

Determination of performance criteria for high-level solidified nuclear waste

Energy Technology Data Exchange (ETDEWEB)

Heckman, R.A.; Holdsworth, T.

1979-05-07

To minimize radiological risk from the operation of a waste management system, performance limits on volatilization, particulate dispersion, and dissolution characteristics of solidified high level waste must be specified. The results show clearly that the pre-emplacement environs are more limiting in establishing the waste form performance criteria than the post-emplacement environs. Absolute values of expected risk are very sensitive to modeling assumptions. The transportation and interim storage operations appear to be most limiting in determining the performance characteristics required. The expected values of risk do not rely upon the repositories remaining intact over the potentially hazardous lifetime of the waste.

Determination of performance criteria for high-level solidified nuclear waste

International Nuclear Information System (INIS)

Heckman, R.A.; Holdsworth, T.

1979-01-01

To minimize radiological risk from the operation of a waste management system, performance limits on volatilization, particulate dispersion, and dissolution characteristics of solidified high level waste must be specified. The results show clearly that the pre-emplacement environs are more limiting in establishing the waste form performance criteria than the post-emplacement environs. Absolute values of expected risk are very sensitive to modeling assumptions. The transportation and interim storage operations appear to be most limiting in determining the performance characteristics required. The expected values of risk do not rely upon the repositories remaining intact over the potentially hazardous lifetime of the waste

Water Distribution and Removal Model

International Nuclear Information System (INIS)

Y. Deng; N. Chipman; E.L. Hardin

2005-01-01

The design of the Yucca Mountain high level radioactive waste repository depends on the performance of the engineered barrier system (EBS). To support the total system performance assessment (TSPA), the Engineered Barrier System Degradation, Flow, and Transport Process Model Report (EBS PMR) is developed to describe the thermal, mechanical, chemical, hydrological, biological, and radionuclide transport processes within the emplacement drifts, which includes the following major analysis/model reports (AMRs): (1) EBS Water Distribution and Removal (WD and R) Model; (2) EBS Physical and Chemical Environment (P and CE) Model; (3) EBS Radionuclide Transport (EBS RNT) Model; and (4) EBS Multiscale Thermohydrologic (TH) Model. Technical information, including data, analyses, models, software, and supporting documents will be provided to defend the applicability of these models for their intended purpose of evaluating the postclosure performance of the Yucca Mountain repository system. The WD and R model ARM is important to the site recommendation. Water distribution and removal represents one component of the overall EBS. Under some conditions, liquid water will seep into emplacement drifts through fractures in the host rock and move generally downward, potentially contacting waste packages. After waste packages are breached by corrosion, some of this seepage water will contact the waste, dissolve or suspend radionuclides, and ultimately carry radionuclides through the EBS to the near-field host rock. Lateral diversion of liquid water within the drift will occur at the inner drift surface, and more significantly from the operation of engineered structures such as drip shields and the outer surface of waste packages. If most of the seepage flux can be diverted laterally and removed from the drifts before contacting the wastes, the release of radionuclides from the EBS can be controlled, resulting in a proportional reduction in dose release at the accessible environment

Water Distribution and Removal Model

Energy Technology Data Exchange (ETDEWEB)

Y. Deng; N. Chipman; E.L. Hardin

2005-08-26

The design of the Yucca Mountain high level radioactive waste repository depends on the performance of the engineered barrier system (EBS). To support the total system performance assessment (TSPA), the Engineered Barrier System Degradation, Flow, and Transport Process Model Report (EBS PMR) is developed to describe the thermal, mechanical, chemical, hydrological, biological, and radionuclide transport processes within the emplacement drifts, which includes the following major analysis/model reports (AMRs): (1) EBS Water Distribution and Removal (WD&R) Model; (2) EBS Physical and Chemical Environment (P&CE) Model; (3) EBS Radionuclide Transport (EBS RNT) Model; and (4) EBS Multiscale Thermohydrologic (TH) Model. Technical information, including data, analyses, models, software, and supporting documents will be provided to defend the applicability of these models for their intended purpose of evaluating the postclosure performance of the Yucca Mountain repository system. The WD&R model ARM is important to the site recommendation. Water distribution and removal represents one component of the overall EBS. Under some conditions, liquid water will seep into emplacement drifts through fractures in the host rock and move generally downward, potentially contacting waste packages. After waste packages are breached by corrosion, some of this seepage water will contact the waste, dissolve or suspend radionuclides, and ultimately carry radionuclides through the EBS to the near-field host rock. Lateral diversion of liquid water within the drift will occur at the inner drift surface, and more significantly from the operation of engineered structures such as drip shields and the outer surface of waste packages. If most of the seepage flux can be diverted laterally and removed from the drifts before contacting the wastes, the release of radionuclides from the EBS can be controlled, resulting in a proportional reduction in dose release at the accessible environment. The purposes

Application of the air/water cushion technology for handling of heavy waste packages in Sweden and France

International Nuclear Information System (INIS)

Bosgiraud, Jean-Michel; Seidler, Wolf K.; Londe, Louis; Thurner, Erik; Pettersson, Stig

2008-01-01

The disposal of certain types of radioactive waste canisters in a deep repository involves handling and emplacement of very heavy loads. The weight of these particular canisters can be in the order of 20 to 50 metric tons. They generally have to be handled underground in openings that are not much larger than the canisters themselves as it is time consuming and expensive to excavate and backfill large openings in a repository. This therefore calls for the development of special technology that can meet the requirements for safe operation in an industrial scale in restrained operating spaces. Air/water cushion lifting systems are used world wide in the industry for moving heavy loads. However, until now the technology needed for emplacing heavy cylindrical radioactive waste packages in bored drifts (with narrow annular gaps) has not been developed or demonstrated previously. This paper describes the related R and D work carried out by ANDRA (for air cushion technology) and by SKB and Posiva (for water cushion technology) respectively, mainly within the framework of the European Commission (EC) funded Integrated Project called ESDRED (6th European Framework Programme). The background for both the air and the water cushion applications is presented. The specific characteristics of the two different emplacement concepts are also elaborated. The various phases of the Test Programmes (including the Prototype phases) are detailed and illustrated for the two lifting media. Conclusions are drawn for each system developed and evaluated. Finally, based on the R and D experience, improvements deemed necessary for an industrial application are listed. The tests performed so far have shown that the emplacement equipment developed is operating efficiently. However further tests are required to verify the availability and the reliability of the equipment over longer periods of time and to identify the modifications that would be needed for an industrial application in a nuclear

Engineering design study for storage and disposal of intermediate level waste

Energy Technology Data Exchange (ETDEWEB)

Griffin, J R; Hackney, S; Richardson, J A; Heafield, W

1982-11-01

A conceptual design study is presented which covers both the storage and disposal of intermediate level waste; repositories in several rock formations are considered at a 300m depth. A total system is proposed including an engineered trench for ..beta gamma.. waste, emplacement systems and off site transportation. Safety during the emplacement phase and the radiological effects of human intrusion and geological catastrophies are considered.

Waste Isolation Pilot Plant simulated RH TRU waste experiments: Data and interpretation pilot

International Nuclear Information System (INIS)

Molecke, M.A.; Argueello, G.J.; Beraun, R.

1993-04-01

The simulated, i.e., nonradioactive remote-handled transuranic waste (RH TRU) experiments being conducted underground in the Waste Isolation Pilot Plant (WIPP) were emplaced in mid-1986 and have been in heated test operation since 9/23/86. These experiments involve the in situ, waste package performance testing of eight full-size, reference RH TRU containers emplaced in horizontal, unlined test holes in the rock salt ribs (walls) of WIPP Room T. All of the test containers have internal electrical heaters; four of the test emplacements were filled with bentonite and silica sand backfill materials. We designed test conditions to be ''near-reference'' with respect to anticipated thermal outputs of RH TRU canisters and their geometrical spacing or layout in WIPP repository rooms, with RH TRU waste reference conditions current as of the start date of this test program. We also conducted some thermal overtest evaluations. This paper provides a: detailed test overview; comprehensive data update for the first 5 years of test operations; summary of experiment observations; initial data interpretations; and, several status; experimental objectives -- how these tests support WIPP TRU waste acceptance, performance assessment studies, underground operations, and the overall WIPP mission; and, in situ performance evaluations of RH TRU waste package materials plus design details and options. We provide instrument data and results for in situ waste container and borehole temperatures, pressures exerted on test containers through the backfill materials, and vertical and horizontal borehole-closure measurements and rates. The effects of heat on borehole closure, fracturing, and near-field materials (metals, backfills, rock salt, and intruding brine) interactions were closely monitored and are summarized, as are assorted test observations. Predictive 3-dimensional thermal and structural modeling studies of borehole and room closures and temperature fields were also performed

10 CFR 60.143 - Monitoring and testing waste packages.

Science.gov (United States)

2010-01-01

....143 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) DISPOSAL OF HIGH-LEVEL RADIOACTIVE WASTES IN... repository operations area, the environment of the waste packages selected for the waste package monitoring program shall be representative of the environment in which the wastes are to be emplaced. (c) The waste...

Engineered Barrier System: Physical and Chemical Environment Model

Energy Technology Data Exchange (ETDEWEB)

D. M. Jolley; R. Jarek; P. Mariner

2004-02-09

The conceptual and predictive models documented in this Engineered Barrier System: Physical and Chemical Environment Model report describe the evolution of the physical and chemical conditions within the waste emplacement drifts of the repository. The modeling approaches and model output data will be used in the total system performance assessment (TSPA-LA) to assess the performance of the engineered barrier system and the waste form. These models evaluate the range of potential water compositions within the emplacement drifts, resulting from the interaction of introduced materials and minerals in dust with water seeping into the drifts and with aqueous solutions forming by deliquescence of dust (as influenced by atmospheric conditions), and from thermal-hydrological-chemical (THC) processes in the drift. These models also consider the uncertainty and variability in water chemistry inside the drift and the compositions of introduced materials within the drift. This report develops and documents a set of process- and abstraction-level models that constitute the engineered barrier system: physical and chemical environment model. Where possible, these models use information directly from other process model reports as input, which promotes integration among process models used for total system performance assessment. Specific tasks and activities of modeling the physical and chemical environment are included in the technical work plan ''Technical Work Plan for: In-Drift Geochemistry Modeling'' (BSC 2004 [DIRS 166519]). As described in the technical work plan, the development of this report is coordinated with the development of other engineered barrier system analysis model reports.

Engineered Barrier System: Physical and Chemical Environment

Energy Technology Data Exchange (ETDEWEB)

P. Dixon

2004-04-26

The conceptual and predictive models documented in this Engineered Barrier System: Physical and Chemical Environment Model report describe the evolution of the physical and chemical conditions within the waste emplacement drifts of the repository. The modeling approaches and model output data will be used in the total system performance assessment (TSPA-LA) to assess the performance of the engineered barrier system and the waste form. These models evaluate the range of potential water compositions within the emplacement drifts, resulting from the interaction of introduced materials and minerals in dust with water seeping into the drifts and with aqueous solutions forming by deliquescence of dust (as influenced by atmospheric conditions), and from thermal-hydrological-chemical (THC) processes in the drift. These models also consider the uncertainty and variability in water chemistry inside the drift and the compositions of introduced materials within the drift. This report develops and documents a set of process- and abstraction-level models that constitute the engineered barrier system: physical and chemical environment model. Where possible, these models use information directly from other process model reports as input, which promotes integration among process models used for total system performance assessment. Specific tasks and activities of modeling the physical and chemical environment are included in the technical work plan ''Technical Work Plan for: In-Drift Geochemistry Modeling'' (BSC 2004 [DIRS 166519]). As described in the technical work plan, the development of this report is coordinated with the development of other engineered barrier system analysis model reports.

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

International Nuclear Information System (INIS)

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

1985-01-01

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

WASTE PACKAGE TRANSPORTER DESIGN

Energy Technology Data Exchange (ETDEWEB)

D.C. Weddle; R. Novotny; J. Cron

1998-09-23

The purpose of this Design Analysis is to develop preliminary design of the waste package transporter used for waste package (WP) transport and related functions in the subsurface repository. This analysis refines the conceptual design that was started in Phase I of the Viability Assessment. This analysis supports the development of a reliable emplacement concept and a retrieval concept for license application design. The scope of this analysis includes the following activities: (1) Assess features of the transporter design and evaluate alternative design solutions for mechanical components. (2) Develop mechanical equipment details for the transporter. (3) Prepare a preliminary structural evaluation for the transporter. (4) Identify and recommend the equipment design for waste package transport and related functions. (5) Investigate transport equipment interface tolerances. This analysis supports the development of the waste package transporter for the transport, emplacement, and retrieval of packaged radioactive waste forms in the subsurface repository. Once the waste containers are closed and accepted, the packaged radioactive waste forms are termed waste packages (WP). This terminology was finalized as this analysis neared completion; therefore, the term disposal container is used in several references (i.e., the System Description Document (SDD)) (Ref. 5.6). In this analysis and the applicable reference documents, the term ''disposal container'' is synonymous with ''waste package''.

Preliminary results of a coupled fracture-flow test at the 0.5 m scale

International Nuclear Information System (INIS)

Blair, S.C.; Costantino, M.S.

1997-01-01

Understanding the hydrologic response of fluids present in the proposed repository horizon to the construction of a repository and the subsequent storage of high-level radioactive waste is crucial to the evaluation of Yucca Mountain as a suitable repository location. Moreover, recent analysis of site measurements imply that the rate of infiltration of water into Yucca Mountain is higher than previously thought, and an in-depth review of all data and analyses has so far confirmed the new estimates (Taylor, 1997). It is also known that the potential repository horizon at Yucca Mountain contains a significant number of fractures. This fact, coupled with higher estimates of the infiltration rate, has resulted in a revision of the general hydrologic model for Yucca Mountain; and it is now thought that the significant flow regime at this site is episodic fast flow in fractures. This implies that surface water can travel deep into Yucca Mountain through fractures (Taylor, 1997). Given these recent developments, understanding flow in the fractured rock mass at Yucca Mountain is a critical requirement for viability assessment. Further, it is well known that the stress field in the rock surrounding the drifts will be altered by both the excavation of drifts and the heating of the rock associated with waste emplacement and storage. It also is generally accepted that flow in fractured rock is highly dependent on the stress field and temperature in the rock. This implies that the hydrologic behavior of rock surrounding emplacement drifts in the potential repository is dependent on the mechanical response of the fractured rock mass to excavation and waste emplacement

Richards Barrier LA Reference Design Feature Evaluation

International Nuclear Information System (INIS)

N.E. Kramer

1999-01-01

The Richards Barrier is one of the design features of the repository to be considered for the License Application (LA), Richards was a soil scientist who first described the diversion of moisture between two materials with different hydrologic properties. In this report, a Richards Barrier is a special type of backfill with a fine-grained material (such as sand) overlaying a coarse-grained material (such as gravel). Water that enters an emplacement drift will first encounter the fine-grained material and be transported around the coarse-grained material covering the waste package, thus protecting the waste package from contact with most of the groundwater. The objective of this report is to discuss the benefits and liabilities to the repository by the inclusion of a Richards Barrier type backfill in emplacement drifts. The Richards Barrier can act as a barrier to water flow, can reduce the waste package material dissolution rate, limit mobilization of the radionuclides, and can provide structural protection for the waste package. The scope of this report is to: (1) Analyze the behavior of barrier materials following the intrusion of groundwater for influxes of 1 to 300 mm per year. The report will demonstrate diversion of groundwater intrusions into the barrier over an extended time period when seismic activity and consolidation may cause the potential for liquefaction and settlement of the Richards Barrier. (2) Review the thermal effects of the Richards Barrier on material behavior. (3) Analyze the effect of rockfall on the performance of the Richards Barrier and the depth of the barrier required to protect waste packages under the barrier. (4) Review radiological and heating conditions on placement of multiple layers of the barrier. Subsurface Nuclear Safety personnel will perform calculations to determine the radiation reduction-time relationship and shielding capacity of the barrier. (5) Evaluate the effects of ventilation on cooling of emplacement drifts and

10 CFR 63.134 - Monitoring and testing waste packages.

Science.gov (United States)

2010-01-01

....134 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) DISPOSAL OF HIGH-LEVEL RADIOACTIVE WASTES IN A... geologic repository operations area, the environment of the waste packages selected for the waste package monitoring program must be representative of the environment in which the wastes are to be emplaced. (c) The...

Conceptual waste packaging options for deep borehole disposal

Energy Technology Data Exchange (ETDEWEB)

Su, Jiann -Cherng [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Hardin, Ernest L. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States)

2015-07-01

This report presents four concepts for packaging of radioactive waste for disposal in deep boreholes. Two of these are reference-size packages (11 inch outer diameter) and two are smaller (5 inch) for disposal of Cs/Sr capsules. All four have an assumed length of approximately 18.5 feet, which allows the internal length of the waste volume to be 16.4 feet. However, package length and volume can be scaled by changing the length of the middle, tubular section. The materials proposed for use are low-alloy steels, commonly used in the oil-and-gas industry. Threaded connections between packages, and internal threads used to seal the waste cavity, are common oilfield types. Two types of fill ports are proposed: flask-type and internal-flush. All four package design concepts would withstand hydrostatic pressure of 9,600 psi, with factor safety 2.0. The combined loading condition includes axial tension and compression from the weight of a string or stack of packages in the disposal borehole, either during lower and emplacement of a string, or after stacking of multiple packages emplaced singly. Combined loading also includes bending that may occur during emplacement, particularly for a string of packages threaded together. Flask-type packages would be fabricated and heat-treated, if necessary, before loading waste. The fill port would be narrower than the waste cavity inner diameter, so the flask type is suitable for directly loading bulk granular waste, or loading slim waste canisters (e.g., containing Cs/Sr capsules) that fit through the port. The fill port would be sealed with a tapered, threaded plug, with a welded cover plate (welded after loading). Threaded connections between packages and between packages and a drill string, would be standard drill pipe threads. The internal flush packaging concepts would use semi-flush oilfield tubing, which is internally flush but has a slight external upset at the joints. This type of tubing can be obtained with premium, low

SUBSURFACE CONSTRUCTION AND DEVELOPMENT ANALYSIS

International Nuclear Information System (INIS)

N.E. Kramer

1998-01-01

The purpose of this analysis is to identify appropriate construction methods and develop a feasible approach for construction and development of the repository subsurface facilities. The objective of this analysis is to support development of the subsurface repository layout for License Application (LA) design. The scope of the analysis for construction and development of the subsurface Repository facilities covers: (1) Excavation methods, including application of knowledge gained from construction of the Exploratory Studies Facility (ESF). (2) Muck removal from excavation headings to the surface. This task will examine ways of preventing interference with other subsurface construction activities. (3) The logistics and equipment for the construction and development rail haulage systems. (4) Impact of ground support installation on excavation and other construction activities. (5) Examination of how drift mapping will be accomplished. (6) Men and materials handling. (7) Installation and removal of construction utilities and ventilation systems. (8) Equipping and finishing of the emplacement drift mains and access ramps to fulfill waste emplacement operational needs. (9) Emplacement drift and access mains and ramps commissioning prior to handover for emplacement operations. (10) Examination of ways to structure the contracts for construction of the repository. (11) Discussion of different construction schemes and how to minimize the schedule risks implicit in those schemes. (12) Surface facilities needed for subsurface construction activities

SCALING ANALYSIS OF REPOSITORY HEAT LOAD FOR REDUCED DIMENSIONALITY MODELS

International Nuclear Information System (INIS)

MICHAEL T. ITAMUA AND CLIFFORD K. HO

1998-01-01

The thermal energy released from the waste packages emplaced in the potential Yucca Mountain repository is expected to result in changes in the repository temperature, relative humidity, air mass fraction, gas flow rates, and other parameters that are important input into the models used to calculate the performance of the engineered system components. In particular, the waste package degradation models require input from thermal-hydrologic models that have higher resolution than those currently used to simulate the T/H responses at the mountain-scale. Therefore, a combination of mountain- and drift-scale T/H models is being used to generate the drift thermal-hydrologic environment

Phenomena associated with magma expansion into a drift

International Nuclear Information System (INIS)

Gaffney, E.S.

2002-01-01

One of the significant threats to the proposed Yucca Mountain nuclear waste repository has been identified as the possibility of intersection of the underground structure by a basaltic intrusion. Based on the geology of the region, it is assumed that such an intrusion would consist of an alkali basalt similar to the nearby Lathrop Wells cone, which has been dated at about 78 ka. The threat of radioactive release may be either from eruption through the surface above the repository of basalt that had been contaminated or from migration through ground water of radionucleides released as a result of damage to waste packages that interact with the magma. As part of our study of these threats, we are analyzing the phenomena associated with magma expansion into drifts in tuff. The early phenomena of the encounter of volatile-rich basaltic magma with a drift are discussed here.

Engineered barrier emplacement experiment in Opalinus clay for the disposal of radioactive waste in underground repositories

International Nuclear Information System (INIS)

Mayor, J. C.; Garcia-Sineriz, J. L.; Alonso, E.; Alheid, H. J.; Blumbling, P.

2005-01-01

exclusively devoted to the GBM: fabrication, emplacement equipments, and hydraulic, mechanical and geophysical (seismic) characterization in conventional laboratory. Chapter 4 refers to the excavation of the experimental drift, and to the geophysical (seismic, electric) and hydrogeological characterization of the near field of the clay rock. Chapter 5 is devoted to the installation of the different in situ test components, systems and instruments. The experimental in situ test results and interpretation is presented in Chapter 6, while their modelling and interpretation is presented in Chapter 7. Chapter 8 includes an assessment of the experiment results and overall conclusions, in relation to the declared project objectives. Chapter 9 consists of a list of references. Although a large number of bibliographic references have been used in this project, each included in the corresponding specific reports, the only references included herein are the specific reports detailing the various tasks performed to date, plus a few other reports considered to be essential. (Author)

Engineered barrier emplacement experiment in opalinus clay for the disposal of radioactive waste in underground repositories

International Nuclear Information System (INIS)

Mayor, J. C.; Garcia-Sineriz, J. L.; Alonso, E.; Alheid, H. J.; Blumbling, P.

2005-01-01

exclusively devoted to the GBM: fabrication, emplacement equipments, and hydraulic, mechanical and geophysical (seismic) characterization in conventional laboratory. Chapter 4 refers to the excavation of the experimental drift, and to the geophysical (seismic, electric) and hydrogeological characterization of the near field of the clay rock. Chapter 5 is devoted to the installation of the different in situ test components, systems and instruments. The experimental in situ test results and interpretation is presented in Chapter 6, while their modelling and interpretation is presented in Chapter 7. Chapter 8 includes an assessment of the experiment results and overall conclusions, in relation to the declared project objectives. Chapter 9 consists of a list of references. Although a large number of bibliographic references have been used in this project, each included in the corresponding specific reports, the only references included herein are the specific reports detailing the various tasks performed to date, plus a few other reports considered to be essential. (Author)

Ocean disposal of heat generating radioactive waste

International Nuclear Information System (INIS)

1984-07-01

This report is based on an emplacement techniques review prepared for the Department of the Environment in February 1983, which appeared as Chapter III of the Nuclear Energy Agency, Seabed Working Group's Status Report. The original document (DOE/RW/83.032) has been amended to take account of the results of field trials carried out in March 1983 and to better reflect current UK Government policy on ocean disposal of HGW. In particular Figure 7 has been redrawn using more realistic drag factors for the calculation of the terminal velocity in water. This report reviews the work conducted by the SWG member countries into the different techniques of emplacing heat generating radioactive waste into the deep ocean sediments. It covers the waste handling from the port facilities to final emplacement in the seabed and verification of the integrity of the canister isolation system. The two techniques which are currently being considered in detail are drilled emplacement and the free fall penetrator. The feasibility study work in progress for both techniques as well as the mathematical and physical modelling work for embedment depth and hole closure behind the penetrator are reviewed. (author)

A review and synthesis of international proposals for the disposal of high-level radioactive wastes into crystalline rock formations

International Nuclear Information System (INIS)

1981-05-01

Examination of the broad range of international concepts for the disposal of high-level radioactive wastes into crystalline rock formations has indicated that systems based upon solid waste units provide the greatest degree of engineering control and security. Three particular disposal concepts are considered worthy of detailed evaluation. In order of priority these are:-tunnel networks with 'in-floor' waste emplacement; matrix of vertical emplacement holes drilled from the surface; tunnel networks with 'in-room' waste emplacement. A review of the international literature has shown that at least ten countries have embarked upon study programmes, but only five have developed detailed conceptual design proposals. These are:- Canada, France, Sweden, the United Kingdom, and the United States. Differing economic, environmental, historical and political circumstances have influenced the pattern of international studies and, to the uninitiated, these factors may obscure some of the relevant technical considerations. Nevertheless, a broad technical concensus is apparent in that all countries currently favour tunnel networks with 'in-floor' waste emplacement. The subject is discussed in detail. (author)

ERDA waste management program

International Nuclear Information System (INIS)

Kuhlman, C.W.

1976-01-01

The ERDA commercial waste program is summarized. It consists of three parts: terminal storage, processing, and preparation of the Generic Environmental Impact Statement. Emplacement in geologic formations is the best disposal method for high-level waste; migration would be essentially zero, as it was in the Oklo event. Solidification processes are needed. Relations with the states, etc. are touched upon

Buffer construction technique using granular bentonite

International Nuclear Information System (INIS)

Masuda, Ryoichi; Asano, Hidekazu; Toguri, Satohito; Mori, Takuo; Shimura, Tomoyuki; Matsuda, Takeshi; Uyama, Masao; Noda, Masaru

2007-01-01

Buffer construction using bentonite pellets as filling material is a promising technology for enhancing the ease of repository operation. In this study, a test of such technology was conducted in a full-scale simulated disposal drift, using a filling system which utilizes a screw conveyor system. The simulated drift, which contained two dummy overpacks, was configured as a half-cross-section model with a height of 2.22 m and a length of 6.0 m. The average dry density of the buffer obtained in the test was 1.29 Mg/m 3 , with an angle of repose of 35 to 40 degrees. These test results indicate that buffer construction using a screw conveyor system for pellet emplacement in a waste disposal drift is a promising technology for repositories for high level radioactive wastes. (author)

Performance Assessment Modeling and Sensitivity Analyses of Generic Disposal System Concepts.

Energy Technology Data Exchange (ETDEWEB)

Sevougian, S. David; Freeze, Geoffrey A.; Gardner, William Payton; Hammond, Glenn Edward; Mariner, Paul

2014-09-01

directly, rather than through simplified abstractions. It also a llows for complex representations of the source term, e.g., the explicit representation of many individual waste packages (i.e., meter - scale detail of an entire waste emplacement drift). This report fulfills the Generic Disposal System Analysis Work Packa ge Level 3 Milestone - Performance Assessment Modeling and Sensitivity Analyses of Generic Disposal System Concepts (M 3 FT - 1 4 SN08080 3 2 ).

The waste bin: nuclear waste dumping and storage in the Pacific

International Nuclear Information System (INIS)

Branch, J.B.

1984-01-01

Relatively small amounts of nuclear waste have been stored on Pacific islands and dumped into the Pacific Ocean since 1945. Governments of Pacific countries possessing nuclear power plants are presently seeking permanent waste storage and disposal solutions at Pacific sites including subseabed emplacement of high-level nuclear wastes and ocean dumping of low-level wastes. This article examines these plans and the response of Pacific islanders in their development of policies and international strategies to ban the proposed dumping on a regional basis. Island governments are preparing for a Regional Convention during which a treaty concerned with radioactive waste storage and disposal will be signed. (Author)

Nuclear waste disposal: technology and environmental hazards

International Nuclear Information System (INIS)

Hare, F.K.; Aikin, A.M.

1980-01-01

The subject is discussed under the headings: introduction; the nature and origin of wastes (fuel cycles; character of wastes; mining and milling operations; middle stages; irradiated fuel; reprocessing (waste generation); reactor wastes); disposal techniques and disposal of reprocessing wastes; siting of repositories; potential environmental impacts (impacts after emplacement in a rock repository; catastrophic effects; dispersion processes (by migrating ground water); thermal effects; future security; environmental survey, monitoring and modelling); conclusion. (U.K.)

Stability of disposal rooms during waste retrieval

International Nuclear Information System (INIS)

Brandshaug, T.

1989-03-01

This report presents the results of a numerical analysis to determine the stability of waste disposal rooms for vertical and horizontal emplacement during the period of waste retrieval. It is assumed that waste retrieval starts 50 years after the initial emplacement of the waste, and that access to and retrieval of the waste containers take place through the disposal rooms. It is further assumed that the disposal rooms are not back-filled. Convective cooling of the disposal rooms in preparation for waste retrieval is included in the analysis. Conditions and parameters used were taken from the Nevada Nuclear Waste Storage Investigation (NNWSI) Project Site Characterization Plan Conceptual Design Report (MacDougall et al., 1987). Thermal results are presented which illustrate the heat transfer response of the rock adjacent to the disposal rooms. Mechanical results are presented which illustrate the predicted distribution of stress, joint slip, and room deformations for the period of time investigated. Under the assumption that the host rock can be classified as ''fair to good'' using the Geomechanics Classification System (Bieniawski, 1974), only light ground support would appear to be necessary for the disposal rooms to remain stable. 23 refs., 28 figs., 2 tabs

Lessons learned in demonstration projects regarding operational safety during final disposal of vitrified waste and spent fuel

International Nuclear Information System (INIS)

Filbert, Wolfgang; Herold, Philipp

2015-01-01

The paper summarizes the lessons learned in demonstration projects regarding operational safety during the final disposal of vitrified waste and spent fuel. The three demonstration projects for the direct disposal of vitrified waste and spent fuel are described. The first two demonstration projects concern the shaft transport of heavy payloads of up to 85 t and the emplacement operations in the mine. The third demonstration project concerns the borehole emplacement operation. Finally, open issues for the next steps up to licensing of the emplacement and disposal systems are summarized.

Repository thermal response: A preliminary evaluation of the effects of modeled waste stream resolution

International Nuclear Information System (INIS)

Ryder, E.E.; Dunn, E.

1995-09-01

One of the primary factors that influences our predictions of host-rock thermal response within a high level waste repository is how the waste stream's represented in the models. In the context of thermal modeling, waste stream refers to an itemized listing of the type (pressurized-water or boiling-water reactor), age, burnup, and enrichment of the spent nuclear fuel assemblies entering the repository over the 25-year emplacement phase. The effect of package-by-package variations in spent fuel characteristics on predicted repository thermal response is the focus of this report. A three-year portion of the emplacement period was modeled using three approaches to waste stream resolution. The first assumes that each package type emplaced in a given year is adequately represented by average characteristics. For comparison, two models that explicitly account for each waste package's individual characteristics were run; the first assuming a random selection of packages and the second an ordered approach aimed at locating the higher power output packages toward the center of the emplacement area. Results indicate that the explicit representation of packages results in hot and cold spots that could have performance assessment and design implications. Furthermore, questions are raised regarding the representativeness of average characteristics with respect to integrated energy output and the possible implications of a mass-based repository loading approach

System analysis for a dual-purpose repository

International Nuclear Information System (INIS)

Bechthold, W.; Closs, K.D.; Knapp, U.; Papp, R.

1993-01-01

This study focuses on planning and evaluating alternative concepts for dealing with the back-end of the nuclear fuel cycle. Emphasis is laid on the underground part of this system consisting in the disposal of radioactive waste/spent fuel that generates considerable heat. A given amount of spent fuel discharged from LWRs is subject to disposal and reprocessing, respectively. For the emerging mix of waste and spent fuel packages a sizeable variety of disposal variants is designed conceptually, thereby providing the database for further near- and far-field analyses as well as dose and cost calculations. The technical alternatives were compared with one another by a panel of experts relying on a set of objectives (criteria) which belong to seven criteria fields. Concept BD1 (borehole and drift emplacement) and concept D3* (three-level drift emplacement) best meet the requirements as stated by the criteria. The comparative analysis also helped detect areas of insufficient knowledge which introduced a degree of uncertainty into the systems evaluation and gave rise to a continuation of the research programme

Review of the scientific and technical criteria for the Waste Isolation Pilot Plant (WIPP)

International Nuclear Information System (INIS)

1984-01-01

The panel has evaluated the scientific and technical adequacy of work being done on the Waste Isolation Pilot Plant (WIPP) project to satisfy the charge to the panel set out in Chapter 1. The panel concluded that the scientific work has been carried out with a high degree of professional competence. The panel notes that the geology revealed by shaft sinking and excavation of drifts and the preliminary measurements generally confirm the geologic expectations derived from surface explorations and boreholes. The purity and volume of the salt, the absence of brine pockets at the repository horizon in the areas excavated, the absence of breccia pipes and of toxic gases, and the nearly horizontal bedding of the salt indicate that a repository can be constructed that will meet the geologic criteria for site selection. Thus, the important issues about the geology at the site have been resolved, but there remain some issues about the hydrology and design of the facility that should be resolved before large-scale transuranic (TRU) waste emplacement begins. The panel's conclusions and recommendations regarding the following studies are presented: site selection and characterization; in-situ tests and experiments; waste acceptance criteria; design and construction of underground facilities; and performance assessment. 65 references, 17 figures, 3 tables

Test plan for buried waste containment system materials

International Nuclear Information System (INIS)

Weidner, J.; Shaw, P.

1997-03-01

The objectives of the FY 1997 barrier material work at the Idaho National Engineering and Environmental Laboratory are to (1) select a waste barrier material and verify that it is compatible with the Buried Waste Containment System Process, and (2) determine if, and how, the Buried Waste Containment System emplacement process affects the material properties and performance (on proof of principle scale). This test plan describes a set of measurements and procedures used to validate a waste barrier material for the Buried Waste Containment System. A latex modified proprietary cement manufactured by CTS Cement Manufacturing Company will be tested. Emplacement properties required for the Buried Waste Containment System process are: slump between 8 and 10 in., set time between 15 and 30 minutes, compressive strength at set of 20 psi minimum, and set temperature less than 100 degrees C. Durability properties include resistance to degradation from carbonate, sulfate, and waste-site soil leachates. A set of baseline barrier material properties will be determined to provide a data base for comparison with the barrier materials when tested in the field. The measurements include permeability, petrographic analysis to determine separation and/or segregation of mix components, and a set of mechanical properties. The measurements will be repeated on specimens from the field test material. The data will be used to determine if the Buried Waste Containment System equipment changes the material. The emplacement properties will be determined using standard laboratory procedures and instruments. Durability of the barrier material will be evaluated by determining the effect of carbonate, sulfate, and waste-site soil leachates on the compressive strength of the barrier material. The baseline properties will be determined using standard ASTM procedures. 9 refs., 1 fig., 2 tabs

Alternative-waste-form evaluation for Savannah River Plant high-level waste

International Nuclear Information System (INIS)

Gould, T.H. Jr.; Crandall, J.L.

1982-01-01

Results of the waste form evaluation are summarized as: risks of human exposure are comparable and extremely small for either borosilicate glass or Synroc ceramic. Waste form properties are more than adequate for either form. The waste form decision can therefore be made on the basis of practicality and cost effectiveness. Synroc offers lower costs for transportation and emplacement. The borosilicate glass form offers the lowest total disposal cost, much simpler and less costly production, an established and proven process, lower future development costs, and an earlier startup of the DWPF

RIP Input From WAPDEG for LA Design Selection: Enhanced Design Alternative II

International Nuclear Information System (INIS)

B.E. Bullard

1999-01-01

The purpose of this analysis is to identify and analyze concepts for the acquisition of data in support of the Performance Confirmation (PC) program at the potential subsurface nuclear waste repository at Yucca Mountain. This analysis is being prepared to document an investigation of design concepts, current available technology, technology trends, and technical issues associated with data acquisition during the PC period. This analysis utilizes the ''Performance Confirmation Plan'' (CRWMS M and O 2000b) to help define the scope for the PC data acquisition system. The focus of this analysis is primarily on the PC period for a minimum of 30 years after emplacement of the last waste package. The design of the data acquisition system shall allow for a closure deferral up to 300 years from initiation of waste emplacement. (CRWMS M and O 2000h, page 5-1). This analysis is a revision to and supercedes analysis, ''Performance Confirmation Data Acquisition System'', DI No. BCAI00000-017 17-0200-00002 Rev 00 (CRWMS M and O 1997), and incorporates the latest repository design changes following the M and O and DOE evaluation of a series of Enhanced Design Alternatives (EDAs), as described in the ''Enhanced Design Alternatives II Report'' (CRWMS M and O 1999d). Significant design changes include: thermal line loading of the emplacement drifts, closer spacing of the waste packages (WPs), wider spacing and fewer emplacement drifts, continuous ventilation of all active emplacement drifts, thinner walled WP designs which will increase external radiation levels, a 50-year repository closure option, inclusion of a drip-shield, exclusion of backfill, and new conceptual designs for the waste emplacement vehicles and equipment (Stroupe 2000). The scope and primary objectives of this analysis are to: (1) Review the criteria for design as presented in the Performance Confirmation Data Acquisition/Monitoring System Description Document, by way of the Input Transmittal, ''Performance

Review of DOE waste package program. Subtask 1.1. National waste package program, April-September 1982

International Nuclear Information System (INIS)

Soo, P.

1983-03-01

The current effort is part of an ongoing task to evaluate the national high-level waste package effort. It includes evaluations of reference waste form, container, and packing material components with respect to determining how they may contribute to the containment and controlled release of radionuclides after waste packages have been emplaced in salt and basalt repositories. Chemical and mechanical failure/degradation modes for the waste package have been reviewed and the licensing data requirements to demonstrate compliance with NRC performance objectives specified

Design Feature 7: Continuous Preclosure Ventilation

International Nuclear Information System (INIS)

A.T. Watkins

1999-01-01

This design feature (DF) is intended to evaluate the effects of continuous ventilation in the emplacement drifts during preclosure and how the effects, if any, compare to the Viability Assessment (VA) reference design for postclosure long term performance. This DF will be evaluated against a set of criteria provided by the License Application Design Selection (LADS) group. The VA reference design included a continuous ventilation airflow quantity of 0.1 m 3 /s in the emplacement drifts in the design of the repository subsurface facilities. The effects of this continuous ventilation during the preclosure was considered to have a negligible effect on postclosure performance and therefore is not included during postclosure in the assessment of the long term performance. This DF discusses the effects of continuous ventilation on the emplacement drift environment and surrounding rock conditions during preclosure for three increased airflow quantities. The three cases of continuous ventilation systems are: System A, 1.0 m 3 /s (Section 8), System B, 5.0 m 3 /s (Section 9), and System C, 10.0 m 3 /s (Section 10) in each emplacement drift split. An emplacement drift split is half total length of emplacement drift going from the east or west main to the exhaust main. The difference in each system is the quantity of airflow in the emplacement drifts

Mont Terri Project - Engineered barrier emplacement experiment in Opalinus Clay for the disposal of radioactive waste in underground repositories

International Nuclear Information System (INIS)

Mayor, J. C.; Garcia-Sineriz, J.; Alonso, E.; Alheid, H.-J.; Bluemling, P.

2007-01-01

The Engineered Barrier (EB) experiment was a full-scale test for the demonstration, in a horizontal drift, of an emplacement technics of the clay barrier, using a granular bentonite material in the upper part of this barrier and bentonite blocks at the bottom. The test has been carried out in a 6 m long section of a niche excavated in Opalinus Clay of the Mont Terri underground laboratory. A steel dummy canister, with the same dimensions and weight as the real reference canisters, was placed on top of a bed of highly compacted bentonite blocks (in turn lying on a concrete bed), and the rest of the clay barrier volume was backfilled with a Granular Bentonite Material (GBM), made of very highly compacted pellets of different sizes. Hydro-mechanical instrumentation and an artificial hydration system (to accelerate the saturation of the clay barrier) were installed, and the test section sealed with a concrete plug. The evolution of the hydro-mechanical parameters along the hydration, both in the barrier and in the clayey rock formation, has been monitored during about 1.5 years, and modelled using the CODE-BRIGHT code. The EB experiment has proved that fully automated production of a Granular Bentonite Material (GBM) is possible and large quantities can be produced in due time in the required quality. Only minor modifications of existing production lines in industry for other applications were necessary to achieve this result. In the EB test section, a dry density of 1.36 g/cm 3 of the emplaced GBM has been obtained. With this value it is estimated that the hydraulic conductivity of this material is lower than 5 x 10 -12 m/s and the swelling pressure is about 1.3 MPa. Even though the EB test section conditions are now not considered as representative of a true demonstration, it is deemed that the model emplacement testing results (dry density of about 1.40 g/cm 3 ) serve well to demonstrate the achievable densities expected in the real world setting. The artificial

Mont Terri Project - Engineered barrier emplacement experiment in Opalinus Clay for the disposal of radioactive waste in underground repositories

Energy Technology Data Exchange (ETDEWEB)

Mayor, J. C. [Empresa Nacional de Residuos Radioactivos SA (ENRESA), Madrid (Spain); Garcia-Sineriz, J. [Asociacion para la Investigacion y Desarollo Industrial de los Recursos Naturales (AITEMIN), Madrid (Spain); Alonso, E. [Centre Internacional de Metodos Numerics en Ingenyeria (CIMNE), Barcelona (Spain); Alheid, H.-J. [Bundesanstalt fuer Geowissenschaften und Rohstoffe (BGR), Hannover (Germany); Bluemling, P. [National Cooperative for the Disposal of Radioactive Waste (Nagra), Wettingen (Switzerland)

2007-07-01

The Engineered Barrier (EB) experiment was a full-scale test for the demonstration, in a horizontal drift, of an emplacement technics of the clay barrier, using a granular bentonite material in the upper part of this barrier and bentonite blocks at the bottom. The test has been carried out in a 6 m long section of a niche excavated in Opalinus Clay of the Mont Terri underground laboratory. A steel dummy canister, with the same dimensions and weight as the real reference canisters, was placed on top of a bed of highly compacted bentonite blocks (in turn lying on a concrete bed), and the rest of the clay barrier volume was backfilled with a Granular Bentonite Material (GBM), made of very highly compacted pellets of different sizes. Hydro-mechanical instrumentation and an artificial hydration system (to accelerate the saturation of the clay barrier) were installed, and the test section sealed with a concrete plug. The evolution of the hydro-mechanical parameters along the hydration, both in the barrier and in the clayey rock formation, has been monitored during about 1.5 years, and modelled using the CODE-BRIGHT code. The EB experiment has proved that fully automated production of a Granular Bentonite Material (GBM) is possible and large quantities can be produced in due time in the required quality. Only minor modifications of existing production lines in industry for other applications were necessary to achieve this result. In the EB test section, a dry density of 1.36 g/cm{sup 3} of the emplaced GBM has been obtained. With this value it is estimated that the hydraulic conductivity of this material is lower than 5 x 10{sup -12} m/s and the swelling pressure is about 1.3 MPa. Even though the EB test section conditions are now not considered as representative of a true demonstration, it is deemed that the model emplacement testing results (dry density of about 1.40 g/cm{sup 3}) serve well to demonstrate the achievable densities expected in the real world setting. The

Albedo Neutron Dosimetry in a Deep Geological Disposal Repository for High-Level Nuclear Waste.

Science.gov (United States)

Pang, Bo; Becker, Frank

2017-04-28

Albedo neutron dosemeter is the German official personal neutron dosemeter in mixed radiation fields where neutrons contribute to personal dose. In deep geological repositories for high-level nuclear waste, where neutrons can dominate the radiation field, it is of interest to investigate the performance of albedo neutron dosemeter in such facilities. In this study, the deep geological repository is represented by a shielding cask loaded with spent nuclear fuel placed inside a rock salt emplacement drift. Due to the backscattering of neutrons in the drift, issues concerning calibration of the dosemeter arise. Field-specific calibration of the albedo neutron dosemeter was hence performed with Monte Carlo simulations. In order to assess the applicability of the albedo neutron dosemeter in a deep geological repository over a long time scale, spent nuclear fuel with different ages of 50, 100 and 500 years were investigated. It was found out, that the neutron radiation field in a deep geological repository can be assigned to the application area 'N1' of the albedo neutron dosemeter, which is typical in reactors and accelerators with heavy shielding. © The Author 2016. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

Engineering for a disposal facility using the in-room emplacement method

Energy Technology Data Exchange (ETDEWEB)

Baumgartner, P; Bilinsky, D M; Ates, Y; Read, R S; Crosthwaite, J L; Dixon, D A

1996-06-01

This report describes three nuclear fuel waste disposal vaults using the in-room emplacement method. First, a generic disposal vault design is provided which is suitable for a depth range of 500 m to 1000 m in highly stressed, sparsely fractured rock. The design process is described for all components of the system. The generic design is then applied to two different disposal vaults, one at a depth of 750 m in a low hydraulically conductive, sparsely fractured rock mass and another at a depth of 500 m in a higher conductivity, moderately fractured rock mass. In the in-room emplacement method, the disposal containers with used-fuel bundles are emplaced within the confines of the excavated rooms of a disposal vault. The discussion of the disposal-facility design process begins with a detailed description of a copper-shell, packed-particulate disposal container and the factors that influenced its design. The disposal-room generic design is presented including the detailed specifications, the scoping and numerical thermal and thermal mechanical analyses, the backfilling and sealing materials, and the operational processes. One room design is provided that meets all the requirements for a vault depth range of 500 to 1000 m. A disposal-vault layout and the factors that influenced its design are also presented, including materials handling, general logistics, and separation of radiological and nonradiological operations. Modifications to the used-fuel packaging plant for the filling and sealing of the copper-shell, packed-particulate disposal containers and a brief description of the common surface facilities needed by the disposal vault and the packaging plant are provided. The implementation of the disposal facility is outlined, describing the project stages and activities and itemizing a specific plan for each of the project stages: siting, construction, operation; decommissioning; and closure. (author). 72 refs., 15 tabs., 63 figs.

Engineering for a disposal facility using the in-room emplacement method

International Nuclear Information System (INIS)

Baumgartner, P.; Bilinsky, D.M.; Ates, Y.; Read, R.S.; Crosthwaite, J.L.; Dixon, D.A.

1996-06-01

This report describes three nuclear fuel waste disposal vaults using the in-room emplacement method. First, a generic disposal vault design is provided which is suitable for a depth range of 500 m to 1000 m in highly stressed, sparsely fractured rock. The design process is described for all components of the system. The generic design is then applied to two different disposal vaults, one at a depth of 750 m in a low hydraulically conductive, sparsely fractured rock mass and another at a depth of 500 m in a higher conductivity, moderately fractured rock mass. In the in-room emplacement method, the disposal containers with used-fuel bundles are emplaced within the confines of the excavated rooms of a disposal vault. The discussion of the disposal-facility design process begins with a detailed description of a copper-shell, packed-particulate disposal container and the factors that influenced its design. The disposal-room generic design is presented including the detailed specifications, the scoping and numerical thermal and thermal mechanical analyses, the backfilling and sealing materials, and the operational processes. One room design is provided that meets all the requirements for a vault depth range of 500 to 1000 m. A disposal-vault layout and the factors that influenced its design are also presented, including materials handling, general logistics, and separation of radiological and nonradiological operations. Modifications to the used-fuel packaging plant for the filling and sealing of the copper-shell, packed-particulate disposal containers and a brief description of the common surface facilities needed by the disposal vault and the packaging plant are provided. The implementation of the disposal facility is outlined, describing the project stages and activities and itemizing a specific plan for each of the project stages: siting, construction, operation; decommissioning; and closure. (author)

RETRIEVAL EQUIPMENT DESCRIPTIONS

International Nuclear Information System (INIS)

J. Steinhoff

1997-01-01

The objective and the scope of this document are to list and briefly describe the major mobile equipment necessary for waste package (WP) retrieval from the proposed subsurface nuclear waste repository at Yucca Mountain. Primary performance characteristics and some specialized design features of the equipment are explained and summarized in the individual subsections of this document. There are no quality assurance requirements or QA controls in this document. Retrieval under normal conditions is accomplished with the same fleet of equipment as is used for emplacement. Descriptions of equipment used for retrieval under normal conditions is found in Emplacement Equipment Descriptions, DI: BCAF00000-01717-5705-00002 (a document in progress). Equipment used for retrieval under abnormal conditions is addressed in this document and consists of the following: (1) Inclined Plane Hauler; (2) Bottom Lift Transporter; (3) Load Haul Dump (LHD) Loader; (4) Heavy Duty Forklift for Emplacement Drifts; (5) Covered Shuttle Car; (6) Multipurpose Vehicle; and (7) Scaler

Numerical experiments on the probability of seepage into underground openings in heterogeneous fractured rock

International Nuclear Information System (INIS)

Birkholzer, J.; Li, G.; Tsang, C.F.; Tsang, Y.

1998-01-01

An important issue for the performance of underground nuclear waste repositories is the rate of seepage into the waste emplacement drifts. A prediction of this rate is particularly complicated for the potential repository site at Yucca Mountain, Nevada, because it is located in thick, unsaturated, fractured tuff formations. Underground opening in unsaturated media might act as capillary barriers, diverting water around them. In the present work, they study the potential rate of seepage into drifts as a function of the percolation flux at Yucca Mountain, based on a stochastic model of the fractured rock mass in the drift vicinity. A variety of flow scenarios are considered, assuming present-day and possible future climate conditions. They show that the heterogeneity in the flow domain is a key factor controlling seepage rates, since it causes channelized flow and local ponding in the unsaturated flow field

United Kingdom government policy towards radioactive waste

International Nuclear Information System (INIS)

Pritchard, G.

1986-01-01

There are three areas of radioactive waste management which exemplify, beyond any reasonable doubt, that the United Kingdom has in the past (and intends in the future), to pursue a policy of dispersal and disposal of radioactive wastes: These are: (I) dumping of low-level waste in the deep ocean and, on a parallel, seabed emplacement of highly active waste; (II) the liquid discharges from Windscale into the Irish Sea; and (III) land dumping of low- and intermediate-level waste

Technical concept for test of geologic storage of spent reactor fuel in the Climax granite, Nevada Test Site

International Nuclear Information System (INIS)

Ramspott, L.D.; Ballou, L.B.; Carlson, R.C.; Montan, D.N.; Butkovich, T.R.; Duncan, J.E.; Patrick, W.C.; Wilder, D.G.; Brough, W.G.; Mayr, M.C.

1979-01-01

The Spent Fuel Test in the Climax granite at the Nevada Test Site is a generic test in which spent fuel assemblies from an operating commercial nuclear reactor are emplaced at, and retrieved from, a plausible waste repository depth in a typical granite. Eleven canisters of spent fuel are emplaced in a storage drift 420 m below the surface along with six electrical simulator canisters. Two adjacent drifts contain electrical heaters which are operated so as to simulate the initial five years of the temperature-stress-displacement fields of a large repository. The site is described, and the pre-operational measurement program and characteristics of the spent fuel are given. Both thermal and mechanical response calculations are summarized. The field instrumentation and data acquisition systems are described, as well as the system for handling the spent fuel

Measurement methods for radiological characterisation of low-active and mid-active radioactive waste for emplacement; Messmethoden fuer die radiologische Charakterisierung von niedrig - und mittelaktiven radioaktiven Abfaellen fuer die Einlagerung

Energy Technology Data Exchange (ETDEWEB)

Sokcic-Kostic, Marina; Schultheis, Roland [NUKEM Technologies GmbH, Alzenau (Germany)

2014-03-15

For radiological classification and characterisation of radioactive waste - as it is here considered - the specification of a multitude of parameters is necessary. The measurement or rather definition of parameter ought to guarantee, that waste packages can be handled safely, that radioactive waste repository corresponds to the respective waste and that safety of emplacement is assured. From the parameters further properties of waste, such as the share of long-living isotopes, as well as activity limiting values, are deducted. For this purpose necessary measurements can be divided into non-destructive and destructive ones. The validity, sensitiveness and accuracy of both measurement methods differ. For the destructive methods, samples from the waste packages are retrieved and examined at the laboratory. In case of non-destructive methods, the entire package is scanned, whereby depending on the nuclides and their specific emissions (Gamma- and Neutron radiation, both the Beta- as well as Alpha- radiation) specific measurement methods arise. Available methods are evaluated and introduced with regard to accuracy, reliability as well as handling. Regarding the hardware- with exception of neutron evaluation technics - progress lies less in the development of new methods, rather than in the production line of robust and reliable measurement devices, which can be applied in automated infrastructures. During the evaluation routine simulation with the Monte-Carlo-Methods establishes itself more and more. Main focus regarding changes lies nevertheless in the introduction of the Bayes-Theory, which calculates consistently, and reliably measurement values and their errors, as well as trust intervals. (orig.)

Low-level waste disposal performance assessments - Total source-term analysis

Energy Technology Data Exchange (ETDEWEB)

Wilhite, E.L.

1995-12-31

Disposal of low-level radioactive waste at Department of Energy (DOE) facilities is regulated by DOE. DOE Order 5820.2A establishes policies, guidelines, and minimum requirements for managing radioactive waste. Requirements for disposal of low-level waste emplaced after September 1988 include providing reasonable assurance of meeting stated performance objectives by completing a radiological performance assessment. Recently, the Defense Nuclear Facilities Safety Board issued Recommendation 94-2, {open_quotes}Conformance with Safety Standards at Department of Energy Low-Level Nuclear Waste and Disposal Sites.{close_quotes} One of the elements of the recommendation is that low-level waste performance assessments do not include the entire source term because low-level waste emplaced prior to September 1988, as well as other DOE sources of radioactivity in the ground, are excluded. DOE has developed and issued guidance for preliminary assessments of the impact of including the total source term in performance assessments. This paper will present issues resulting from the inclusion of all DOE sources of radioactivity in performance assessments of low-level waste disposal facilities.

State of the art for fabricating and emplacing concrete containers into large horizontal disposal caverns in the french geological repository - 59267

International Nuclear Information System (INIS)

Bosgiraud, Jean-Michel; Guariso, Maurice; Pineau, Francois

2012-01-01

The research and development work presented in this paper was initialized by Andra in 2007. The work necessary for manufacturing and testing a full scale demonstrator is presently implemented. The case story is twofold. The first part is related to the initial development of a high performance concrete formulation used for fabricating concrete storage containers (containing Intermediate Level and Long Lived Waste primary canisters) to be stacked and emplaced into 400-m long concrete lined horizontal disposal vaults (also called cavern), excavated in the Callovo-Oxfordian clay host formation at a 550 to 600-m depth, with an inside diameter of approximately 8-m. The fabrication of the concrete boxes is illustrated. The second part presents the outcome at the end of the detailed design phase, for a system which is now being manufactured (for further test and assembly), for the emplacement of the concrete containers inside the vault. The application was engineered for remote emplacing a pile of 2 concrete containers (the containers are preliminarily stacked in a pile of 2, inside a hot cell, thanks to a ground travelling gantry crane). The emplacement process is justified and the related emplacement synoptic is illustrated. The test campaign is scheduled in 2011-2012. The successful completion of the technical trials is mandatory to confirm the mechanical feasibility of remotely emplacing concrete containers into large horizontal disposal caverns over long distances. The later display of the machinery at work in Andra's showroom will be instrumental for the confidence building process involving the various stakeholders concerned by the public enquiry period (mid-2013) preceding the deep geological repository license application (2014-2015). (authors)

Abstraction of Models for Pitting and Crevice Corrosion of Drip Shield and Waste Package Outer Barrier

Energy Technology Data Exchange (ETDEWEB)

K. Mon

2001-08-29

This analyses and models report (AMR) was conducted in response to written work direction (CRWMS M and O 1999a). ICN 01 of this AMR was developed following guidelines provided in TWP-MGR-MD-000004 REV 01, ''Technical Work Plan for: Integrated Management of Technical Product Input Department'' (BSC 2001, Addendum B). The purpose and scope of this AMR is to review and analyze upstream process-level models (CRWMS M and O 2000a and CRWMS M and O 2000b) and information relevant to pitting and crevice corrosion degradation of waste package outer barrier (Alloy 22) and drip shield (Titanium Grade 7) materials, and to develop abstractions of the important processes in a form that is suitable for input to the WAPDEG analysis for long-term degradation of waste package outer barrier and drip shield in the repository. The abstraction is developed in a manner that ensures consistency with the process-level models and information and captures the essential behavior of the processes represented. Also considered in the model abstraction are the probably range of exposure conditions in emplacement drifts and local exposure conditions on drip shield and waste package surfaces. The approach, method, and assumptions that are employed in the model abstraction are documented and justified.

Abstraction of Models for Pitting and Crevice Corrosion of Drip Shield and Waste Package Outer Barrier

International Nuclear Information System (INIS)

Mon, K.

2001-01-01

This analyses and models report (AMR) was conducted in response to written work direction (CRWMS M and O 1999a). ICN 01 of this AMR was developed following guidelines provided in TWP-MGR-MD-000004 REV 01, ''Technical Work Plan for: Integrated Management of Technical Product Input Department'' (BSC 2001, Addendum B). The purpose and scope of this AMR is to review and analyze upstream process-level models (CRWMS M and O 2000a and CRWMS M and O 2000b) and information relevant to pitting and crevice corrosion degradation of waste package outer barrier (Alloy 22) and drip shield (Titanium Grade 7) materials, and to develop abstractions of the important processes in a form that is suitable for input to the WAPDEG analysis for long-term degradation of waste package outer barrier and drip shield in the repository. The abstraction is developed in a manner that ensures consistency with the process-level models and information and captures the essential behavior of the processes represented. Also considered in the model abstraction are the probably range of exposure conditions in emplacement drifts and local exposure conditions on drip shield and waste package surfaces. The approach, method, and assumptions that are employed in the model abstraction are documented and justified

Ventilation System Strategy for a Prospective Korean Radioactive Waste Repository

International Nuclear Information System (INIS)

Kim, Jin; Kwon, Sang Ki

2005-01-01

In the stage of conceptual design for the construction and operation of the geologic repository for radioactive wastes, it is important to consider a repository ventilation system which serves the repository working environment, hygiene and safety of the public at large, and will allow safe maintenance like moisture content elimination in repository for the duration of the repositories life, construction/operation/closure, also allowing safe waste transportation and emplacement. This paper describes the possible ventilation system design criteria and requirements for the prospective Korean radioactive waste repositories with emphasis on the underground rock cavity disposal method in the both cases of low and medium-level and high-level wastes. It was found that the most important concept is separate ventilation systems for the construction (development) and waste emplacement (storage) activities. In addition, ventilation network system modeling, natural ventilation, ventilation monitoring systems and real time ventilation simulation, and fire simulation and emergency system in the repository are briefly discussed.

Co-disposal of mixed waste materials

International Nuclear Information System (INIS)

Phillips, S.J.; Alexander, R.G.; Crane, P.J.; England, J.L.; Kemp, C.J.; Stewart, W.E.

1993-08-01

Co-disposal of process waste streams with hazardous and radioactive materials in landfills results in large, use-efficiencies waste minimization and considerable cost savings. Wasterock, produced from nuclear and chemical process waste streams, is segregated, treated, tested to ensure regulatory compliance, and then is placed in mixed waste landfills, burial trenches, or existing environmental restoration sites. Large geotechnical unit operations are used to pretreat, stabilize, transport, and emplace wasterock into landfill or equivalent subsurface structures. Prototype system components currently are being developed for demonstration of co-disposal

Thermal analyses for a nuclear-waste repository in tuff using USW-G1 borehole data

International Nuclear Information System (INIS)

Johnson, R.L.

1982-10-01

Thermal calculations using properties of tuffs obtained from the USW-G1 borehole, located near the SW margin of the Nevada Test Site (NTS), have been completed for a nuclear waste repository sited in welded tuff below the water table. The analyses considered two wasteforms, high level waste and spent fuel, emplaced at two different, gross thermal loadings, 50 and 75 kW/Acre (20.24 and 30.36 kW/ha). Calculations were made assuming that no boiling of the groundwater occurs; i.e., that the hydrostatic head potential was reestablished soon after waste emplacement. 23 figures, 2 tables

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

Effect on localized waste-container failure on radionuclide transport from an underground nuclear waste vault

International Nuclear Information System (INIS)

Cheung, S.C.H.; Chan, T.

1983-07-01

In the geological disposal of nuclear fuel waste, one option is to emplace the waste container in a borehole drilled into the floor of the underground vault. In the borehole, the waste container is surrounded by a compacted soil material known as the buffer. A finite-element simulation has been performed to study the effect of localized partial failure of the waste container on the steady-state radionuclide transport by diffusion from the container through the buffer to the surrounding rock and/or backfill. In this study, the radionuclide concentration at the buffer-backfill interface is assumed to be zero. Two cases are considered at the interface between the buffer and the rock. In case 1, a no-flux boundary condition is used to simulate intact rock. In case 2, a constant radionuclide concentration condition is used to simulate fractured rock with groundwater flow. The results show that the effect of localized partial failure of the waste container on the total flux is dependent on the boundary condition at the buffer-rock interface. For the intact rock condition, the total flux is mainly dependent on the location of the failure. The total flux increases as the location changes from the bottom to the top of the emplaced waste container. For a given localized failure of the waste container, the total flux remains unaffected by the area of failed surface below the top of the failure. For fractured rock, the total flux is directly proportional to the failed surface area of the waste container regardless of the failure location

Emplacement and stemming of nuclear explosives for Plowshare applications

Energy Technology Data Exchange (ETDEWEB)

Cramer, J L [Lawrence Radiation Laboratory, University of California, Livermore, CA (United States)

1970-05-15

This paper will discuss the various methods used for emplacement and design considerations that must be taken into account when the emplacement and stemming method is selected. The step-by-step field procedure will not be discussed in this paper. The task of emplacing and stemming the nuclear explosive is common to all Plowshare experiments today. All present-day applications of a nuclear explosive for Plowshare experiments require that the detonation take place some distance below the surface of the ground. This is normally done by lowering the explosive into an emplacement hole to a desired depth and then backfilling the hole with a suitable stemming material. At first glance it scenes like a very straightforward, simple task to perform. It would appear to be a task that could become a standard procedure for all experiments; however, this is not the case. In actuality, the emplacement and stemming of a nuclear explosive must almost be a custom design. It varies with the application of the experiment, i.e., cratering or underground engineering. It also varies with the condition of the hole, the available equipment to do the job, the actual purpose of the stemming, possible postshot reentry, hydrology, geology, and future production. A very important item that must always be considered is the protection of the firing and signal cables during the downhole and stemming operation. Each of these things must be considered; ignoring any one of them could jeopardize one of the objectives of the experiment or perhaps even the experiment itself. It should be emphasized that for a multiple-shot program such as would be used to develop a gas field where the geology, depths of burial etc. are the same, the emplacement and stemming operation would be standardized, as would all other parts of the program. However, for individual experiments in totally different areas, complete standardization of the emplacement and stemming is impossible.

Emplacement and stemming of nuclear explosives for Plowshare applications

International Nuclear Information System (INIS)

Cramer, J.L.

1970-01-01

This paper will discuss the various methods used for emplacement and design considerations that must be taken into account when the emplacement and stemming method is selected. The step-by-step field procedure will not be discussed in this paper. The task of emplacing and stemming the nuclear explosive is common to all Plowshare experiments today. All present-day applications of a nuclear explosive for Plowshare experiments require that the detonation take place some distance below the surface of the ground. This is normally done by lowering the explosive into an emplacement hole to a desired depth and then backfilling the hole with a suitable stemming material. At first glance it scenes like a very straightforward, simple task to perform. It would appear to be a task that could become a standard procedure for all experiments; however, this is not the case. In actuality, the emplacement and stemming of a nuclear explosive must almost be a custom design. It varies with the application of the experiment, i.e., cratering or underground engineering. It also varies with the condition of the hole, the available equipment to do the job, the actual purpose of the stemming, possible postshot reentry, hydrology, geology, and future production. A very important item that must always be considered is the protection of the firing and signal cables during the downhole and stemming operation. Each of these things must be considered; ignoring any one of them could jeopardize one of the objectives of the experiment or perhaps even the experiment itself. It should be emphasized that for a multiple-shot program such as would be used to develop a gas field where the geology, depths of burial etc. are the same, the emplacement and stemming operation would be standardized, as would all other parts of the program. However, for individual experiments in totally different areas, complete standardization of the emplacement and stemming is impossible

Reducing Pesticide Drift

Science.gov (United States)

Provides information about pesticide spray drift, including problems associated with drift, managing risks from drift and the voluntary Drift Reduction Technology program that seeks to reduce spray drift through improved spray equipment design.

Mechanisms Of Saucer-Shaped Sill Emplacement: Insight From Experimental Modeling

Science.gov (United States)

Galland, O.; Planke, S.; Malthe-Sørenssen, A.; Polteau, S.; Svensen, H.; Podladchikov, Y. Y.

2006-12-01

It has been recently demonstrated that magma intrusions in sedimentary basins had a strong impact on petroleum systems. Most of these intrusions are sills, and especially saucer-shaped sills. These features can be observed in many sedimentary basins (i.e. the Karoo basin, South Africa; the Norwegian and North Sea; the Tunguska basin, Siberia; the Neuquén basin in Argentina). The occurrence of such features in so various settings suggests that their emplacement results from fundamental processes. However, the mechanisms that govern their formation remain poorly constrained. Experiments were conducted to simulate the emplacement of saucer-shaped magma intrusions in sedimentary basins. The model rock and magma were fine-grained silica flour and molten vegetable oil, respectively. This modeling technique allows simultaneous simulation of magma emplacement and brittle deformation at a basin scale. For our purpose, we performed our experiments without external deformation. During the experiments, the oil was injected horizontally at constant flow rate within the silica flour. Then the oil initially emplaced in a sill, whereas the surface of the model inflated into a smooth dome. Subsequently, the oil propagated upwards along inclined sheets, finally reaching the surface at the edge of the dome. The resulting geometries of the intrusions were saucer-shaped sills. Then the oil solidified, and the model was cut in serial cross-sections through which the structures of the intrusive body and of the overburden can be observed. In order to constraint the processes governing the emplacement of such features, we performed a parametric study based on a set of experiments in which we systematically varied parameters such as the depth of emplacement and the injection flow rate of the oil. Our results showed that saucer diameters are larger at deeper level of emplacement. Opposite trend was obtained with varying injection flow rates. Based on our results, we conducted a detailed

10 CFR 60.135 - Criteria for the waste package and its components.

Science.gov (United States)

2010-01-01

... Section 60.135 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) DISPOSAL OF HIGH-LEVEL RADIOACTIVE WASTES... for HLW shall be designed so that the in situ chemical, physical, and nuclear properties of the waste package and its interactions with the emplacement environment do not compromise the function of the waste...

Transuranic waste management at Savannah River - past, present, and future

International Nuclear Information System (INIS)

D'Ambrosia, J.

1985-01-01

The major objective of the TRU program at Savannah River is to support the TRU National Program, which is dedicated to preparing waste for, and emplacing waste in, the Waste Isolation Pilot Plant, (WIPP). Thus, the Savannah River Program also supports WIPP operations. The Savannah River site specific goals to phase out the indefinite storage of TRU waste, which has been the mode of waste management since 1974, and to dispose of Savannah River's Defense TRU waste

Thermomechanical scoping calculations for the waste package environment tests

International Nuclear Information System (INIS)

Butkovich, T.R.; Yow, J.L. Jr.

1986-03-01

During the site characterization phase of the Nevada Nuclear Waste Storage Investigation Project, tests are planned to provide field information on the hydrological and thermomechanical environment. These results are needed for assessing performance of stored waste packages emplaced at depth in excavations in a rock mass. Scoping calculations were performed to provide information on displacements and stress levels attained around excavations in the rock mass from imposing a thermal load designed to simulate the heat produced by radioactive decay. In this way, approximate levels of stresses and displacements are available for choosing instrumentation type and sensitivity as well as providing indications for optimizing instrument emplacement during the test. 7 refs., 9 figs., 1 tab

Emplacement of small and large buffer blocks

International Nuclear Information System (INIS)

Saari, H.; Nikula, M.; Suikki, M.

2010-05-01

The report describes emplacement of a buffer structure encircling a spent fuel canister to be deposited in a vertical hole. The report deals with installability of various size blocks and with an emplacement gear, as well as evaluates the achieved quality of emplacement and the time needed for installing the buffer. Two block assembly of unequal size were chosen for examination. A first option involved small blocks, the use of which resulted in a buffer structure consisting of small sector blocks 200 mm in height. A second option involved large blocks, resulting in a buffer structure which consists of eight blocks. In these tests, the material chosen for both block options was concrete instead of bentonite. The emplacement test was a three-phase process. A first phase included stacking a two meter high buffer structure with small blocks for ensuring the operation of test equipment and blocks. A second phase included installing buffer structures with both block options to a height matching that of a canister-encircling cylindrical component. A third phase included testing also the installability of blocks to be placed above the canister by using small blocks. In emplacement tests, special attention was paid to the installability of blocks as well as to the time required for emplacement. Lifters for both blocks worked well. Due to the mass to be lifted, the lifter for large blocks had a more heavy-duty frame structure (and other lifting gear). The employed lifters were suspended in the tests on a single steel wire rope. Stacking was managed with both block sizes at adequate precision and stacked-up towers were steady. The stacking of large blocks was considerably faster. Therefore it is probably that the overall handling of the large blocks will be more convenient at a final disposal site. From the standpoint of reliability in lifting, the small blocks were safer to install above the canister. In large blocks, there are strict shape-related requirements which are

Engineering considerations for the Waste Isolation Pilot Plant

International Nuclear Information System (INIS)

Scully, L.W.

1978-01-01

The WIPP, located at Los Medanos in New Mexico, is to be used for DOE transuranic and high-level defense wastes. On the surface, there are contact-handled and remote-handled waste facilities. Package size, delivery rates, shipping, shielding and thermal considerations, underground transport and emplacement, retrievability, ventilation, and hoist conveyence safety are discussed

Determination of performance criteria for high-level solidified nuclear waste from the commercial nuclear fuel cycle: a probabilistic safety analysis

International Nuclear Information System (INIS)

Heckman, R.A.

1978-01-01

To minimize the radiological risk from the operation of a waste management system for the safe disposal of high-level waste, performance characteristics of the solidified waste form must be specified. The minimum waste form characteristics that must be specified are the radionuclide volatilization fraction, airborne particulate dispersion fraction, and the aqueous dissolution characteristics. The results indicate that the pre-emplacement environs are more limiting in establishing the waste form performance criteria than the post-emplacement environs. The actual values of expected risk are sensitive to modeling assumptions and data base uncertainties. The transportation step appears to be the most limiting in determining the required performance characteristics

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

Deep borehole disposal of high-level radioactive waste.

Energy Technology Data Exchange (ETDEWEB)

Stein, Joshua S.; Freeze, Geoffrey A.; Brady, Patrick Vane; Swift, Peter N.; Rechard, Robert Paul; Arnold, Bill Walter; Kanney, Joseph F.; Bauer, Stephen J.

2009-07-01

Preliminary evaluation of deep borehole disposal of high-level radioactive waste and spent nuclear fuel indicates the potential for excellent long-term safety performance at costs competitive with mined repositories. Significant fluid flow through basement rock is prevented, in part, by low permeabilities, poorly connected transport pathways, and overburden self-sealing. Deep fluids also resist vertical movement because they are density stratified. Thermal hydrologic calculations estimate the thermal pulse from emplaced waste to be small (less than 20 C at 10 meters from the borehole, for less than a few hundred years), and to result in maximum total vertical fluid movement of {approx}100 m. Reducing conditions will sharply limit solubilities of most dose-critical radionuclides at depth, and high ionic strengths of deep fluids will prevent colloidal transport. For the bounding analysis of this report, waste is envisioned to be emplaced as fuel assemblies stacked inside drill casing that are lowered, and emplaced using off-the-shelf oilfield and geothermal drilling techniques, into the lower 1-2 km portion of a vertical borehole {approx}45 cm in diameter and 3-5 km deep, followed by borehole sealing. Deep borehole disposal of radioactive waste in the United States would require modifications to the Nuclear Waste Policy Act and to applicable regulatory standards for long-term performance set by the US Environmental Protection Agency (40 CFR part 191) and US Nuclear Regulatory Commission (10 CFR part 60). The performance analysis described here is based on the assumption that long-term standards for deep borehole disposal would be identical in the key regards to those prescribed for existing repositories (40 CFR part 197 and 10 CFR part 63).

Ventilation Model

International Nuclear Information System (INIS)

Yang, H.

1999-01-01

The purpose of this analysis and model report (AMR) for the Ventilation Model is to analyze the effects of pre-closure continuous ventilation in the Engineered Barrier System (EBS) emplacement drifts and provide heat removal data to support EBS design. It will also provide input data (initial conditions, and time varying boundary conditions) for the EBS post-closure performance assessment and the EBS Water Distribution and Removal Process Model. The objective of the analysis is to develop, describe, and apply calculation methods and models that can be used to predict thermal conditions within emplacement drifts under forced ventilation during the pre-closure period. The scope of this analysis includes: (1) Provide a general description of effects and heat transfer process of emplacement drift ventilation. (2) Develop a modeling approach to simulate the impacts of pre-closure ventilation on the thermal conditions in emplacement drifts. (3) Identify and document inputs to be used for modeling emplacement ventilation. (4) Perform calculations of temperatures and heat removal in the emplacement drift. (5) Address general considerations of the effect of water/moisture removal by ventilation on the repository thermal conditions. The numerical modeling in this document will be limited to heat-only modeling and calculations. Only a preliminary assessment of the heat/moisture ventilation effects and modeling method will be performed in this revision. Modeling of moisture effects on heat removal and emplacement drift temperature may be performed in the future

Waste salt disposal at the Savannah River Plant

International Nuclear Information System (INIS)

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

1986-01-01

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

Uncertainties in sealing a nuclear waste repository in partially saturated tuff

International Nuclear Information System (INIS)

Tillerson, J.R.; Fernandez, J.A.; Hinkebein, T.E.

1989-01-01

Sealing a nuclear waste repository in partially saturated tuff presents unique challenges to assuring performance of sealing components. Design and performance of components for sealing shafts, ramps, drifts, and exploratory boreholes depend on specific features of both the repository design and the site; of particular importance is the hydrologic environment in the unsaturated zone, including the role of fracture flow. Repository design features important to sealing of a repository include the size and location of shaft and ramp accesses, excavation methods, and the underground layout features such as grade (drainage direction) and location relative to geologic structure. Uncertainties about seal components relate to the postclosure environment for the seals, the emplacement methods, the material properties, and the potential performance of the components. An approach has been developed to reduce uncertainties and to increase confidence in seal performance; it includes gathering extensive site characterization data, establishing conservative design requirements, testing seal components in laboratory and field environments, and refining designs of both the seals and the repository before seals are installed. 9 refs., 5 figs., 2 tabs

Geotechnical aspects of deep ocean radioactive waste disposal

International Nuclear Information System (INIS)

Freeman, T.J.

1990-01-01

The methods that might be used to bury radioactive waste in the deep ocean, and their likely effect on the sediment barrier, have been the subject of an international research program performed during the last ten years. This paper reviews the geotechnical aspects of deep ocean disposal and discusses how far the research performed has gone towards providing the information needed to assess this form of disposal. Considerable progress has been made during the course of the international program towards understanding the processes involved in the emplacement of heat generating waste (HGW) into the deep ocean bed and the subsequent interactions between the waste and the sediments. These processes do not appear to have a deleterious effect on the barrier properties of the sediments, and it is concluded that it is likely that HGW could be emplaced in the deep ocean in such a way that the seabed would provide an effective containment for the radionuclides

Site characterization plan: Conceptual design report: Volume 4, Appendices F-O: Nevada Nuclear Waste Storage Investigations Project

Energy Technology Data Exchange (ETDEWEB)

MacDougall, H R; Scully, L W; Tillerson, J R [comps.

1987-09-01

The site for the prospective repository is located at Yucca Mountain in southwestern Nevada, and the waste emplacement area will be constructed in the underlying volcanic tuffs. The target horizon for waste emplacement is a sloping bed of densely welded tuff more than 650 ft below the surface and typically more than 600 ft above the water table. The conceptual design described in this report is unique among repository designs in that it uses ramps in addition to shafts to gain access to the underground facility, the emplacement horizon is located above the water table, and it is possible that 300- to 400-ft-long horizontal waste emplacement boreholes will be used. This report summarizes the design bases, design and performance criteria, and the design analyses performed. The current status of meeting the preclosure performance objectives for licensing and of resolving the repository design and preclosure issues is presented. The repository design presented in this report will be expanded and refined during the advanced conceptual design, the license application design, and the final procurement and construction design phases. Volume 4 contains Appendices F to O.

Site characterization plan: Conceptual design report: Volume 4, Appendices F-O: Nevada Nuclear Waste Storage Investigations Project

International Nuclear Information System (INIS)

MacDougall, H.R.; Scully, L.W.; Tillerson, J.R.

1987-09-01

The site for the prospective repository is located at Yucca Mountain in southwestern Nevada, and the waste emplacement area will be constructed in the underlying volcanic tuffs. The target horizon for waste emplacement is a sloping bed of densely welded tuff more than 650 ft below the surface and typically more than 600 ft above the water table. The conceptual design described in this report is unique among repository designs in that it uses ramps in addition to shafts to gain access to the underground facility, the emplacement horizon is located above the water table, and it is possible that 300- to 400-ft-long horizontal waste emplacement boreholes will be used. This report summarizes the design bases, design and performance criteria, and the design analyses performed. The current status of meeting the preclosure performance objectives for licensing and of resolving the repository design and preclosure issues is presented. The repository design presented in this report will be expanded and refined during the advanced conceptual design, the license application design, and the final procurement and construction design phases. Volume 4 contains Appendices F to O

Intermediate depth burial of classified transuranic wastes in arid alluvium

International Nuclear Information System (INIS)

Cochran, J.R.; Crowe, B.M.; Di Sanza, F.

1999-01-01

Intermediate depth disposal operations were conducted by the US Department of Energy (DOE) at the DOE's Nevada Test Site (NTS) from 1984 through 1989. These operations emplaced high-specific activity low-level wastes (LLW) and limited quantities of classified transuranic (TRU) wastes in 37 m (120-ft) deep, Greater Confinement Disposal (GCD) boreholes. The GCD boreholes are 3 m (10 ft) in diameter and founded in a thick sequence of arid alluvium. The bottom 15 m (50 ft) of each borehole was used for waste emplacement and the upper 21 m (70 ft) was backfilled with native alluvium. The bottom of each GCD borehole is almost 200 m (650 ft) above the water table. The GCD boreholes are located in one of the most arid portions of the US, with an average precipitation of 13 cm (5 inches) per year. The limited precipitation, coupled with generally warm temperatures and low humidities results in a hydrologic system dominated by evapotranspiration. The US Environmental Protection Agency's (EPA's) 40 CFR 191 defines the requirements for protection of human health from disposed TRU wastes. This EPA standard sets a number of requirements, including probabilistic limits on the cumulative releases of radionuclides to the accessible environment for 10,000 years. The DOE Nevada Operations Office (DOE/NV) has contracted with Sandia National Laboratories (Sandia) to conduct a performance assessment (PA) to determine if the TRU wastes emplaced in the GCD boreholes complies with the EPA's 40 CFR 191 requirements. This paper describes DOE's actions undertaken to evaluate whether the TRU wastes in the GCD boreholes will, or will not, endanger human health. Based on preliminary modeling, the TRU wastes in the GCD boreholes meet the EPA's requirements, and are, therefore, protective of human health

ENGINEERED BARRIER SYSTEM FEATURES, EVENTS AND PROCESSES

International Nuclear Information System (INIS)

Jaros, W.

2005-01-01

The purpose of this report is to evaluate and document the inclusion or exclusion of engineered barrier system (EBS) features, events, and processes (FEPs) with respect to models and analyses used to support the total system performance assessment for the license application (TSPA-LA). A screening decision, either Included or Excluded, is given for each FEP along with the technical basis for exclusion screening decisions. This information is required by the U.S. Nuclear Regulatory Commission (NRC) at 10 CFR 63.114 (d, e, and f) [DIRS 173273]. The FEPs addressed in this report deal with those features, events, and processes relevant to the EBS focusing mainly on those components and conditions exterior to the waste package and within the rock mass surrounding emplacement drifts. The components of the EBS are the drip shield, waste package, waste form, cladding, emplacement pallet, emplacement drift excavated opening (also referred to as drift opening in this report), and invert. FEPs specific to the waste package, cladding, and drip shield are addressed in separate FEP reports: for example, ''Screening of Features, Events, and Processes in Drip Shield and Waste Package Degradation'' (BSC 2005 [DIRS 174995]), ''Clad Degradation--FEPs Screening Arguments (BSC 2004 [DIRS 170019]), and Waste-Form Features, Events, and Processes'' (BSC 2004 [DIRS 170020]). For included FEPs, this report summarizes the implementation of the FEP in the TSPA-LA (i.e., how the FEP is included). For excluded FEPs, this analysis provides the technical basis for exclusion from TSPA-LA (i.e., why the FEP is excluded). This report also documents changes to the EBS FEPs list that have occurred since the previous versions of this report. These changes have resulted due to a reevaluation of the FEPs for TSPA-LA as identified in Section 1.2 of this report and described in more detail in Section 6.1.1. This revision addresses updates in Yucca Mountain Project (YMP) administrative procedures as they

Impact of partitioning and transmutation on high-level waste disposal for the fast breeder reactor fuel cycle

International Nuclear Information System (INIS)

Nishihara, Kenji; Oigawa, Hiroyuki; Nakayama, Shinichi; Ono, Kiyoshi; Shiotani, Hiroki

2010-01-01

The impact of partitioning and/or transmutation (PT) technology on high-level waste management was investigated for the equilibrium state of several potential fast breeder reactor (FBR) fuel cycles. Three different fuel cycle scenarios involving PT technology were analyzed: 1) partitioning process only (separation of some fission products), 2) transmutation process only (separation and transmutation of minor actinides), and 3) both partitioning and transmutation processes. The conventional light water reactor (LWR) fuel cycle without PT technology, on which the current repository design is based, was also included for comparison. We focused on the thermal constraints in a geological repository and determined the necessary predisposal storage quantities and time periods (by defining a storage capacity index) for several predefined emplacement configurations through transient thermal analysis. The relation between this storage capacity index and the required repository emplacement area was obtained. We found that the introduction of the FBR fuel cycle without PT can yield a 35% smaller repository per unit electricity generation than the LWR fuel cycle, although the predisposal storage period is prolonged from 50 years for the LWR fuel cycle to 65 years for the FBR fuel cycle without PT. The introduction of the partitioning-only process does not result in a significant reduction of the repository emplacement area from that for the FBR fuel cycle without PT, but the introduction of the transmutation-only process can reduce the emplacement area by a factor of 5 when the storage period is extended from 65 to 95 years. When a coupled partitioning and transmutation system is introduced, the repository emplacement area can be reduced by up to two orders of magnitude by assuming a predisposal storage of 60 years for glass waste and 295 years for calcined waste containing the Sr and Cs fraction. The storage period of 295 years for the calcined waste does not require a large

Mechanical behavior of an instrumented shotcrete drifts definitive lining in a 500 m deep clay formation

International Nuclear Information System (INIS)

Zghondi, Jad; Armand, Gilles; Noiret, Aurelien

2012-01-01

Document available in extended abstract form only. At the Meuse/Haute Marne Underground Research Laboratory (URL), Andra has developed a technical and scientific program to test excavation methods in a 500 m deep Callovo Oxfordian clay-stone to demonstrate feasibility of nuclear waste disposal Different types of drift excavations and reinforcements methods has been and will be tested at the URL,in order to evaluate the impact on the surrounding rock behavior, especially the EDZ, and to optimize the design of the reinforcement. At the beginning soft support has been used to let drifts converge, and from time to time the stiffness of support has been increase up to emplace gasketed pre-cast concrete segmental rings just after an open face tunneling excavation end of 2013. In this previous experiment, the target was to apply and on a short time a stiff reinforcement that can have a similar behavior as a pre-cast concrete ring. This paper will present the experimental layout, the measurement tools as well as the first results. The instrumented drift section 'BPE' is 15 m long and 6,3 m diameter; it was excavated by a BRH machine. The excavation sequence was realized with a one meter excavation pass. After each pass, a 10 cm layer of wet mixed fiber reinforced shotcrete was applied on the vault, and 45 cm on the counter vault. The vault 45 cm thickness was reached after three other layers added respectively while proceeding with the three following pass of excavation. Different kinds of measurements were carried out before, during and after excavation, in a way to evaluate the loading of the shotcrete reinforcement as well as the hydro-mechanical behavior of the host rock. Before the excavation of the drift, three standard diameter boreholes have been drilled around the planned drift. They have been equipped with pressure and deformation measurements in a way to monitor the hydro-mechanical impact of the excavation on the surrounding rock. While excavating, the

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

Rooted in two Countries? Migrant Heritage, emplacement and the ‘Canon van Nederland’.

NARCIS (Netherlands)

van Faassen, M.; Hoekstra, F.G.

2017-01-01

Rooted in two Countries? Migrant Heritage, emplacement and the ‘Canon van Nederland’. In the cultural heritage sector theorizing emplacement is considered to be vital for identity: “people need to anchor there identity concretely to a location (emplacement)” (Grever&Van Boxtel, 2014). The question

Electron drift time in silicon drift detectors: A technique for high precision measurement of electron drift mobility

International Nuclear Information System (INIS)

Castoldi, A.; Rehak, P.

1995-01-01

This paper presents a precise absolute measurement of the drift velocity and mobility of electrons in high resistivity silicon at room temperature. The electron velocity is obtained from the differential measurement of the drift time of an electron cloud in a silicon drift detector. The main features of the transport scheme of this class of detectors are: the high uniformity of the electron motion, the transport of the signal electrons entirely contained in the high-purity bulk, the low noise timing due to the very small anode capacitance (typical value 100 fF), and the possibility to measure different drift distances, up to the wafer diameter, in the same semiconductor sample. These features make the silicon drift detector an optimal device for high precision measurements of carrier drift properties. The electron drift velocity and mobility in a 10 kΩ cm NTD n-type silicon wafer have been measured as a function of the electric field in the range of possible operation of a typical drift detector (167--633 V/cm). The electron ohmic mobility is found to be 1394 cm 2 /V s. The measurement precision is better than 1%. copyright 1995 American Institute of Physics

Westinghouse Hanford Company plan for certifying newly generated contact-handled transuranic waste for emplacement in the Waste Isolation Pilot Plant

International Nuclear Information System (INIS)

Lipinski, R.M.; Sheehan, J.S.

1992-07-01

Westinghouse Hanford Company (Westinghouse Hanford) currently manages an interim storage site for Westinghouse Hanford and non-Westinghouse Hanford-generated transuranic (TRU) waste and operates TRU waste generating facilities within the Hanford Site in Washington State. Approval has been received from the Waste Acceptance Criteria Certification Committee (WACCC) and Westinghouse Hanford TRU waste generating facilities to certify newly generated contact-handled TRU (CH-TRU) solid waste to meet the Waste Acceptance Criteria (WAC). This document describes the plan for certifying newly generated CH-TRU solid waste to meet the WAC requirements for storage at the Waste Isolation Pilot Plant (WIPP) site. Attached to this document are facility-specific certification plans for the Westinghouse Hanford TRU waste generators that have received WACCC approval. The certification plans describe operations that generate CH-TRU solid waste and the specific procedures by which these wastes will be certified and segregated from uncertified wastes at the generating facilities. All newly generated CH-TRU solid waste is being transferred to the Transuranic Storage and Assay Facility (TRUSAF) and/or a controlled storage facility. These facilities will store the waste until the certified TRU waste can be sent to the WIPP site and the non-certified TRU waste can be sent to the Waste Receiving and Processing Facility. All non-certifiable TRU waste will be segregated and clearly identified

Preliminary Analysis of Remote Monitoring and Robotic Concepts for Performance Confirmation

International Nuclear Information System (INIS)

McAffee, D.A.

1997-01-01

As defined in 10 CFR Part 60.2, Performance Confirmation is the ''program of tests, experiments and analyses which is conducted to evaluate the accuracy and adequacy of the information used to determine with reasonable assurance that the performance objectives for the period after permanent closure will be met''. The overall Performance Confirmation program begins during site characterization and continues up to repository closure. The main purpose of this document is to develop, explore and analyze initial concepts for using remotely operated and robotic systems in gathering repository performance information during Performance Confirmation. This analysis focuses primarily on possible Performance Confirmation related applications within the emplacement drifts after waste packages have been emplaced (post-emplacement) and before permanent closure of the repository (preclosure). This will be a period of time lasting approximately 100 years and basically coincides with the Caretaker phase of the project. This analysis also examines, to a lesser extent, some applications related to Caretaker operations. A previous report examined remote handling and robotic technologies that could be employed during the waste package emplacement phase of the project (Reference 5.1). This analysis is being prepared to provide an early investigation of possible design concepts and technical challenges associated with developing remote systems for monitoring and inspecting activities during Performance Confirmation. The writing of this analysis preceded formal development of Performance Confirmation functional requirements and program plans and therefore examines, in part, the fundamental Performance Confirmation monitoring needs and operating conditions. The scope and primary objectives of this analysis are to: (1) Describe the operating environment and conditions expected in the emplacement drifts during the preclosure period. (Presented in Section 7.2). (2) Identify and discuss the

Ventilation planning for a prospective nuclear waste repository

International Nuclear Information System (INIS)

Wallace, K.G. Jr.

1987-01-01

In 1982, the US Congress passed the Nuclear Waste Policy Act to provide for the development of underground repositories for spent nuclear fuel. This development will be managed by the United States Department of Energy. In 1986, the President selected three areas for site characterization to determine their suitability for the development of an underground repository; those sites were: (1) A site in volcanic tuff located at Yucca Mountain in Nevada, (2) a site in bedded salt located in Deaf Smith County in Texas, and (3) a site in basalt located in Hanford, Washington. At present conceptual repository designs are being developed for each site. A key element of a repository design is the underground ventilation system required to support construction, nuclear waste emplacement, and potential waste retrieval. This paper describes the preliminary ventilation systems designed for the repository in tuff. The concept provides separate ventilation systems for the construction and waste emplacement activities. The paper further describes the means by which acceptable environmental conditions will be re-established to allow re-entry into previously closed rooms for the purpose of inspection, maintenance or retrieval

Radioactive waste disposal. Facts, problems and responsible action

International Nuclear Information System (INIS)

Finckh, E.; Seitz, M.

1994-01-01

In a first part, natural science and technology aspects of waste management are outlined: basic concepts of radioactivity; properties, detection and primary effects of radioactive radiation; biological effect of radioactivity and radiation; general geological bases; composition of spent fuel elements; interim storage and transport; reprocessing of spent fuels; classification and treatment of radioactive wastes; emplacement possibilities for radioactive wastes; possible ways of radionuclides from the repository back into the biosphere; comparative consideration of the risks involved in nuclear waste management. The second part of the paper deals with ethical and theological aspects of radioactive waste management. (orig./HP) [de

A disposal centre for immobilized nuclear waste

International Nuclear Information System (INIS)

1980-02-01

This report describes a conceptual design of a disposal centre for immobilized nuclear waste. The surface facilities consist of plants for the preparation of steel cylinders containing nuclear waste immobilized in glass, shaft headframe buildings and all necessary support facilities. The underground disposal vault is located on one level at a depth of 1000 m. The waste cylinders are emplaced into boreholes in the tunnel floors. All surface and subsurface facilities are described, operations and schedules are summarized, and cost estimates and manpower requirements are given. (auth)

Drift velocity monitoring of the CMS muon drift chambers

CERN Document Server

Sonnenschein, Lars

2010-01-01

The drift velocity in drift tubes of the CMS muon chambers is a key parameter for the muon track reconstruction and trigger. It needs to be monitored precisely in order to detect any deviation from its nominal value. A change in absolute pressure, a variation of the gas admixture or a contamination of the chamber gas by air affect the drift velocity. Furthermore the temperature and magnetic field influence its value. First data, taken with a dedicated Velocity Drift Chamber (VDC) built by RWTH Aachen IIIA are presented.

Preoperational checkout of the remote-handled transuranic waste handling at the Waste Isolation Pilot Plant

International Nuclear Information System (INIS)

1987-09-01

This plan describes the preoperational checkout for handling Remote-Handled Transuranic (RH-TRU) Wastes from their receipt at the Waste Isolation Pilot Plant (WIPP) to their emplacement underground. This plan identifies the handling operations to be performed, personnel groups responsible for executing these operations, and required equipment items. In addition, this plan describes the quality assurance that will be exercised throughout the checkout, and finally, it establishes criteria by which to measure the success of the checkout. 7 refs., 5 figs

Waste Isolation Pilot Plant RH TRU waste preoperational checkout: Final report

International Nuclear Information System (INIS)

1988-06-01

This report documents the results of the Waste Isolation Pilot Plant (WIPP) Remote-Handled Transuranic (RH TRU) Waste Preoperational Checkout. The primary objective of this checkout was to demonstrate the process of handling RH TRU waste packages, from receipt through emplacement underground, using equipment, personnel, procedures, and methods to be used with actual waste packages. A further objective was to measure operational time lines to provide bases for confirming the WIPP design through put capability and for projecting operator radiation doses. Successful completion of this checkout is a prerequisite to the receipt of actual RH TRU waste. This checkout was witnessed in part by members of the Environmental Evaluation Group (EEG) of the state of New Mexico. Further, this report satisfies a key milestone contained in the Agreement for Consultation and Cooperation with the state of New Mexico. 4 refs., 26 figs., 4 tabs

Modeling studies of unsaturated flow with long-term permeability change at Yucca Mountain

International Nuclear Information System (INIS)

Zhang Chengyuan; Liu Xiaoyan; Liu Quansheng

2008-01-01

The amount of water seeping into the waste emplacement drifts is crucial for the performance of underground nuclear waste repository, since it controls the corrosion rates of waste packages and the mobilization rate of radionuclides. It is limited by water flow through drift vicinity. In the present work we study the potential rates of water flow around drifts as a function of predicted long-term change of permeability at Yucca Mountain, based on a dual-continuum model of the unsaturated flow in fractured rock mass. For stage of DECOVALEX Ⅳ, we used a simplified practical model on unsaturated flow in Yucca Mountain case simulation. These models contain main physical processes that should be considered, including thermal expansion, thermal radiation, water-rock coupling and stress-induced change of permeability. Comparative study with other DECOVALEX team's results shows that they are both good enough and flexible enough to include more physical processes. We can draw the conclusion that it is necessary to model stress-induced changes in permeability and relative processes in future studies, because there are obvious differences (in water saturation and water flux) between simulation cases with and without variable permeability, especially in areas very close to the drift. (authors)

Measurement of the positron-drift time relation of a high-pressure drift chamber

International Nuclear Information System (INIS)

Pruefert, W.

1989-04-01

As a test of its performance, the measurement of the drift time versus drift distance relation of a high pressure drift chamber using cosmic rays is described. Two multiwire proportional chambers, mounted above and below the detector, are used to define the track of the cosmic particle in the drift chamber. The drift chamber is read out by FADCs (Flash Analog to Digital Converter), and the drift time is determined from the FADC signals by the DOS- (Difference Of Samples) method. The measured drift time versus drift distance relation showed good agreement with the relation, which is expected from the spatial dependence of the electric field and the dependence of the drift velocity on this field. (orig.) [de

Quality Assurance Program Plan for the Waste Isolation Pilot Plant Experimental-Waste Characterization Program

International Nuclear Information System (INIS)

1991-01-01

This Quality Assurance Program Plan (QAPP) identifies the quality of data necessary to meet the specific objectives associated with the Department of Energy (DOE) Waste Isolation Pilot Plant (WIPP) Experimental-Waste Characterization Program (the Program). This experimental-waste characterization program is only one part of the WIPP Test Phase, both in the short- and long-term, to quantify and evaluate the characteristics and behavior of transuranic (TRU) wastes in the repository environment. Other parts include the bin-scale and alcove tests, drum-scale tests, and laboratory experiments. In simplified terms, the purpose of the Program is to provide chemical, physical, and radiochemical data describing the characteristics of the wastes that will be emplaced in the WIPP, while the remaining WIPP Test Phase is directed at examining the behavior of these wastes in the repository environment. 50 refs., 35 figs., 33 tabs

Cryovolcanic Emplacement of Domes on Europa

Science.gov (United States)

Quick, Lynnae C.; Glaze, Lori S.; Baloga, Stephen M.

2016-01-01

Here we explore the hypothesis that certain domes on Europa may have been produced by the extrusion of viscous cryolavas. A new mathematical method for the emplacement and relaxation of viscous lava domes is presented and applied to putative cryovolcanic domes on Europa. A similarity solution approach is applied to the governing equation for fluid flow in a cylindrical geometry, and dome relaxation is explored assuming a volume of cryolava has been rapidly emplaced onto the surface. Nonphysical sin- gularities inherent in previous models for dome relaxation have been eliminated, and cryolava cooling is represented by a time-variable viscosity. We find that at the onset of relaxation, bulk kinematic viscosities may lie in the range between 10(exp 3) and 10(exp 6) sq m/s, while the actual fluid lava viscosity may be much lower. Plausible relaxation times to form the domes, which are linked to bulk cryolava rheology, are found to range from 3.6 days to 7.5 years. We find that cooling of the cryolava, while dominated by conduction through an icy skin, should not prevent fluids from advancing and relaxing to form domes within the timescales considered. Determining the range of emplacement conditions for putative cryolava domes will shed light on Europa's resurfacing history. In addition, the rheologies and compositions of erupted cryolavas have implications for subsurface cryomagma ascent and local surface stress conditions on Europa.

Site characterization plan: Conceptual design report, Volume 2: Chapters 4-9: Nevada Nuclear Waste Storage Investigations Project

International Nuclear Information System (INIS)

MacDougall, H.R.; Scully, L.W.; Tillerson, J.R.

1987-09-01

This document presents a description of a prospective geologic repository for high-level radioactive waste to support the development of the Site Characterization Plan for the Yucca Mountain site. The target horizon for waste emplacement is a sloping bed of densely welded tuff more than 650 ft below the surface and typically more than 600 ft above the water table. The conceptual design described in this report is unique among repository designs in that it uses ramps in addition to shafts to gain access to the underground facility, the emplacement horizon is located above the water table, and it is possible that 300- to 400-ft-long horizontal waste emplacement boreholes will be used. This report summarizes the design bases (site and properties of the waste package), design and performance criteria, and the design analyses performed. The current status of meeting the preclosure performance objectives for licensing and of resolving the repository design and preclosure issues is presented. The repository design presented in this report will be expanded and refined during the advanced conceptual design, the license application design, and the final procurement and construction design phases. 147 refs., 145 figs., 83 tabs

Developing an institutional strategy for transporting defense transuranic waste materials

International Nuclear Information System (INIS)

Guerrero, J.V.; Kresny, H.S.

1986-01-01

In late 1988, the US Department of Energy (DOE) expects to begin emplacing transuranic waste materials in the Waste Isolation Pilot Plant (WIPP), an R and D facility to demonstrate the safe disposal of radioactive wastes resulting from defense program activities. Transuranic wastes are production-related materials, e.g., clothes, rags, tools, and similar items. These materials are contaminated with alpha-emitting transuranium radionuclides with half-lives of > 20 yr and concentrations > 100 nCi/g. Much of the institutional groundwork has been done with local communities and the State of New Mexico on the siting and construction of the facility. A key to the success of the emplacement demonstration, however, will be a qualified transportation system together with institutional acceptance of the proposed shipments. The DOE's Defense Transuranic Waste Program, and its contractors, has lead responsibility for achieving this goal. The Joint Integration Office (JIO) of the DOE, located in Albuquerque, New Mexico, is taking the lead in implementing an integrated strategy for assessing nationwide institutional concerns over transportation of defense transuranic wastes and in developing ways to resolve or mitigate these concerns. Parallel prototype programs are under way to introduce both the new packaging systems and the institutional strategy to interested publics and organizations

Thermal analysis of a heat generating waste repository on the seabed

International Nuclear Information System (INIS)

Maynard, M.J.; Butler, T.P.; Firmin, G.H.

1987-02-01

The time dependent thermal behaviour of a repository containing heat generating waste has been investigated during loading, transport, and subsequent emplacement on the seabed. Variations of less than 1 0 C in the sealed repository water temperature were calculated to be sufficient to create adequate water circulation. Conservative 1-D analyses were used to estimate a maximum repository water temperature of 256 0 C, occuring about 3 years after emplacement. The temperature distributions within the heat generating waste canisters and grouted titanium tubes have been calculated using 2-D axisymmetric finite element models. Peak temperatures at the waste centre-line were found to be approx. 40 0 C above the repository water temperature. The sensitivity of the results to assumed thermal parameters and to the effect of sediment accumulation have been considered. The possibility and consequences of steam formation within a vented repository have been discussed. (author)

Drift velocity and pressure monitoring of the CMS muon drift chambers

CERN Document Server

Sonnenschein, Lars

2011-01-01

The drift velocity in drift tubes of the CMS muon chambers is a key parameter for the muon track reconstruction and trigger. It needs to be monitored precisely in order to detect any deviation from its nominal value. A change in absolute pressure, a variation of the gas admixture or a contamination of the chamber gas by air affect the drift velocity. Furthermore, the temperature and magnetic field influence its value. First data, taken with a dedicated Velocity Drift Chamber (VDC) built by RWTH Aachen IIIA are presented. Another important parameter to be monitored is the pressure inside the muon drift tube chambers. The differential pressure must not exceed a certain value and the absolute pressure has to be kept slightly above ambient pressure to prevent air from entering into the muon drift tube chambers in case of a leak. Latest drift velocity monitoring results are discussed.

A feasibility study of the offshore disposal of radioactive waste by drilled emplacement

International Nuclear Information System (INIS)

1984-01-01

This report describes the third phase of a study of the feasibility of disposal and isolation of high level radioactive waste in holes drilled deep into the sediments of the ocean. In this phase, work has concentrated on establishing the logistics of disposing of up to 400 cubic metres of vitrified waste per year, and on the capital and running costs of doing so. The report concludes that the disposal of waste in the form produced by the AVM process is operationally feasible, and that disposal in this way will add approximately 0.2% to the cost of generation of the energy contributing to the waste. (author)

INDIVIDUAL DOSIMETRY IN DISPOSAL REPOSITORY OF HEAT-GENERATING NUCLEAR WASTE.

Science.gov (United States)

Pang, Bo; Saurí Suárez, Héctor; Becker, Frank

2016-09-01

Certain working scenarios in a disposal facility of heat-generating nuclear waste might lead to an enhanced level of radiation exposure for workers in such facilities. Hence, a realistic estimation of the personal dose during individual working scenarios is desired. In this study, the general-purpose Monte Carlo N-Particle code MCNP6 (Pelowitz, D. B. (ed). MCNP6 user manual LA-CP-13-00634, Rev. 0 (2013)) was applied to simulate a representative radiation field in a disposal facility. A tool to estimate the personal dose was then proposed by taking into account the influence of individual motion sequences during working scenarios. As basis for this approach, a movable whole-body phantom was developed to describe individual body gestures of the workers during motion sequences. In this study, the proposed method was applied to the German concept of geological disposal in rock salt. The feasibility of the proposed approach was demonstrated with an example of working scenario in an emplacement drift of a rock salt mine. © The Author 2016. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

Stokes drift

Science.gov (United States)

van den Bremer, T. S.; Breivik, Ø.

2017-12-01

During its periodic motion, a particle floating at the free surface of a water wave experiences a net drift velocity in the direction of wave propagation, known as the Stokes drift (Stokes 1847 Trans. Camb. Philos. Soc. 8, 441-455). More generally, the Stokes drift velocity is the difference between the average Lagrangian flow velocity of a fluid parcel and the average Eulerian flow velocity of the fluid. This paper reviews progress in fundamental and applied research on the induced mean flow associated with surface gravity waves since the first description of the Stokes drift, now 170 years ago. After briefly reviewing the fundamental physical processes, most of which have been established for decades, the review addresses progress in laboratory and field observations of the Stokes drift. Despite more than a century of experimental studies, laboratory studies of the mean circulation set up by waves in a laboratory flume remain somewhat contentious. In the field, rapid advances are expected due to increasingly small and cheap sensors and transmitters, making widespread use of small surface-following drifters possible. We also discuss remote sensing of the Stokes drift from high-frequency radar. Finally, the paper discusses the three main areas of application of the Stokes drift: in the coastal zone, in Eulerian models of the upper ocean layer and in the modelling of tracer transport, such as oil and plastic pollution. Future climate models will probably involve full coupling of ocean and atmosphere systems, in which the wave model provides consistent forcing on the ocean surface boundary layer. Together with the advent of new space-borne instruments that can measure surface Stokes drift, such models hold the promise of quantifying the impact of wave effects on the global atmosphere-ocean system and hopefully contribute to improved climate projections. This article is part of the theme issue 'Nonlinear water waves'.

Drift velocity and pressure monitoring of the CMS muon drift chambers

CERN Document Server

Sonnenschein, Lars

2010-01-01

The drift velocity in drift tubes of the CMS muon chambers is a key parameter for the muon track reconstruction and trigger. It needs to be monitored precisely in order to detect any deviation from its nominal value. A change in absolute pressure, a variation of the gas admixture or a contamination of the chamber gas by air affect the drift velocity. Furthermore, the temperature and magnetic field influence its value. First data, taken with a dedicated Velocity Drift Chamber (VDC) built by RWTH Aachen IIIA are presented. Another important parameter to be monitored is the pressure inside the muon drift tube chambers because the drift velocity depends on it. Furthermore the differential pressure must not exceed a certain value and the absolute pressure has to be kept slightly above ambient pressure to prevent air from entering into the muon drift tube chambers in case of a leak. Latest pressure monitoring results are discussed.

Disposal of Radioactive Waste

International Nuclear Information System (INIS)

2011-01-01

This Safety Requirements publication applies to the disposal of radioactive waste of all types by means of emplacement in designed disposal facilities, subject to the necessary limitations and controls being placed on the disposal of the waste and on the development, operation and closure of facilities. The classification of radioactive waste is discussed. This Safety Requirements publication establishes requirements to provide assurance of the radiation safety of the disposal of radioactive waste, in the operation of a disposal facility and especially after its closure. The fundamental safety objective is to protect people and the environment from harmful effects of ionizing radiation. This is achieved by setting requirements on the site selection and evaluation and design of a disposal facility, and on its construction, operation and closure, including organizational and regulatory requirements.

POST-PROCESSING ANALYSIS FOR THC SEEPAGE

International Nuclear Information System (INIS)

SUN, Y.

2004-01-01

This report describes the selection of water compositions for the total system performance assessment (TSPA) model of results from the thermal-hydrological-chemical (THC) seepage model documented in ''Drift-Scale THC Seepage Model'' (BSC 2004 [DIRS 169856]). The selection has been conducted in accordance with ''Technical Work Plan for: Near-Field Environment and Transport: Coupled Processes (Mountain-Scale TH/THC/THM, Drift-Scale THC Seepage, and Post-Processing Analysis for THC Seepage) Report Integration'' (BSC 2004 [DIRS 171334]). This technical work plan (TWP) was prepared in accordance with AP-2.27Q, ''Planning for Science Activities''. Section 1.2.3 of the TWP describes planning information pertaining to the technical scope, content, and management of this report. The post-processing analysis for THC seepage (THC-PPA) documented in this report provides a methodology for evaluating the near-field compositions of water and gas around a typical waste emplacement drift as these relate to the chemistry of seepage, if any, into the drift. The THC-PPA inherits the conceptual basis of the THC seepage model, but is an independently developed process. The relationship between the post-processing analysis and other closely related models, together with their main functions in providing seepage chemistry information for the Total System Performance Assessment for the License Application (TSPA-LA), are illustrated in Figure 1-1. The THC-PPA provides a data selection concept and direct input to the physical and chemical environment (P and CE) report that supports the TSPA model. The purpose of the THC-PPA is further discussed in Section 1.2. The data selection methodology of the post-processing analysis (Section 6.2.1) was initially applied to results of the THC seepage model as presented in ''Drift-Scale THC Seepage Model'' (BSC 2004 [DIRS 169856]). Other outputs from the THC seepage model (DTN: LB0302DSCPTHCS.002 [DIRS 161976]) used in the P and CE (BSC 2004 [DIRS 169860

ENGINEERED BARRIER SYSTEM: PHYSICAL AND CHEMICAL ENVIRONMENT

International Nuclear Information System (INIS)

Jarek, R.

2004-01-01

The purpose of this report is to describe the evolution of the physical and chemical environmental conditions within the waste emplacement drifts of the repository, including the drip shield and waste package surfaces. The abstraction model is used in the total system performance assessment for the license application (TSPA LA) to assess the performance of the engineered barrier system and the waste form. This report develops and documents a set of these abstraction-level models that describe the engineered barrier system physical and chemical environment. Where possible, these models use information directly from other reports as input, which promotes integration among process models used for TSPA-LA. Specific tasks and activities of modeling the physical and chemical environment are included in ''Technical Work Plan for: Near-Field Environment and Transport In-Drift Geochemistry Model Report Integration'' (BSC 2004 [DIRS 171156], Section 1.2.2). As described in the technical work plan, the development of this report is coordinated with the development of other engineered barrier system reports

ENGINEERED BARRIER SYSTEM: PHYSICAL AND CHEMICAL ENVIRONMENT

International Nuclear Information System (INIS)

G.H. Nieder-Westermann

2005-01-01

The purpose of this report is to describe the evolution of the physical and chemical environmental conditions within the waste emplacement drifts of the repository, including the drip shield and waste package surfaces. The abstraction model is used in the total system performance assessment for the license application (TSPA LA) to assess the performance of the engineered barrier system and the waste form. This report develops and documents a set of these abstraction-level models that describe the engineered barrier system physical and chemical environment. Where possible, these models use information directly from other reports as input, which promotes integration among process models used for TSPA-LA. Specific tasks and activities of modeling the physical and chemical environment are included in ''Technical Work Plan for: Near-Field Environment and Transport In-Drift Geochemistry Model Report Integration'' (BSC 2004 [DIRS 171156], Section 1.2.2). As described in the technical work plan, the development of this report is coordinated with the development of other engineered barrier system reports

EMPLACEMENT GANTRY ITS STANDARDS IDENTIFICATION STUDY

International Nuclear Information System (INIS)

Voegele, M.

2005-01-01

To date, the project has established ITS performance requirements for SSCs based on identification and categorization of event sequences that may result in a radiological release. These performance requirements are defined within the NSDB. Further, SSCs credited with performing safe functions are classified as ITS. In turn, perform confirmation for these SSCs is sought through the use of consensus code and standards. The purpose of this study is to identify applicable codes and standards for the WP Emplacement Gantry ITS SSCs. Further, this study will form the basis for selection and the extent of applicability of each code and standard. This study is based on the design development completed for LA only. Accordingly, identification of ITS SSCs beyond those defined within the NSDB are based on designs that may be subject to further development during detail design. Furthermore, several design alternatives may still be under consideration to satisfy certain safety functions, and that final selection will not be determined until further design development has occurred. Therefore, for completeness, throughout this study alternative designs currently under considered will be discussed. Further, the results of this study will be subject to evaluation as part of a follow-on GAP analysis study. Based on the results of this study the GAP analysis will evaluate each code and standard to ensure each ITS performance requirement is fully satisfied. When a performance requirement is not fully satisfied a ''gap'' is highlighted. Thereafter, the study will identify supplemental requirements to augment the code or standard to meet performance requirements. Further, the GAP analysis will identify non-standard areas of the design that will be subject to a Development Plan. Non-standard components and non-standard design configurations are defined as areas of the design that do not follow standard industry practices or codes and standards. Whereby, performance confirmation cannot be

Earth science developments in support of waste isolation

International Nuclear Information System (INIS)

Duguid, J.O.

1981-01-01

Earth science issues in geologic waste isolation can be subdivided into smaller questions that are resolvable. This approach provides a mechanism for focusing research on topics of definable priority and monitoring progress through the status of issue resolution. The status of resolution of major issues in borehole sealing, interpretation of groundwater hydrology, geochemistry, and repository performance assessment is presented. The Waste Terminal Storage Program has reached a point where the selection of sites, underground testing, and emplacement of waste can proceed on a well-defined schedule

Performance testing of waste forms in a tuff environment

International Nuclear Information System (INIS)

Oversby, V.M.

1983-11-01

This paper describes experimental work conducted to establish the chemical composition of water which will have reacted with Topopah Spring Member tuff prior to contact with waste packages. The experimental program to determine the behavior of spent fuel and borosilicate glass in the presence of this water is then described. Preliminary results of experiments using spent fuel segments with defects in the Zircaloy cladding are presented. Some results from parametric testing of a borosilicate glass with tuff and 304L stainless steel are also discussed. Experiments conducted using Topopah Spring tuff and J-13 well water have been conducted to provide an estimate of the post-emplacement environment for waste packages in a repository at Yucca Mountain. The results show that emplacement of waste packages should cause only small changes in the water chemistry and rock mineralogy. The changes in environment should not have any detrimental effects on the performance of metal barriers or waste forms. The NNWSI waste form testing program has provided preliminary results related to the release rate of radionuclides from the waste package. Those results indicate that release rates from both spent fuel and borosilicate glass should be below 1 part in 10 5 per year. Future testing will be directed toward making release rate testing more closely relevant to site specific conditions. 17 references, 7 figures

Demonstration of close-coupled barriers for subsurface containment of buried waste

International Nuclear Information System (INIS)

Dwyer, B.P.

1996-05-01

A close-coupled barrier is produced by first installing a conventional cement grout curtain followed by a thin inner lining of a polymer grout. The resultant barrier is a cement polymer composite that has economic benefits derived from the cement and performance benefits from the durable and resistant polymer layer. Close-coupled barrier technology is applicable for final, interim, or emergency containment of subsurface waste forms. Consequently, when considering the diversity of technology application, the construction emplacement and material technology maturity, general site operational requirements, and regulatory compliance incentives, the close-coupled barrier system provides an alternative for any hazardous or mixed waste remediation plan. This paper discusses the installation of a close-coupled barrier and the subsequent integrity verification. The demonstration was installed at a benign site at the Hanford Geotechnical Test Facility, 400 Area, Hanford, Washington. The composite barrier was emplaced beneath a 7,500 liter tank. The tank was chosen to simulate a typical DOE Complex waste form. The stresses induced on the waste form were evaluated during barrier construction. The barrier was constructed using conventional jet grouting techniques. Drilling was completed at a 45 degree angle to the ground, forming a conical shaped barrier with the waste form inside the cone. Two overlapping rows of cylindrical cement columns were grouted in a honeycomb fashion to form the secondary backdrop barrier layer. The primary barrier, a high molecular weight polymer manufactured by 3M Company, was then installed providing a relatively thin inner liner for the secondary barrier. The primary barrier was emplaced by panel jet grouting with a dual wall drill stem, two phase jet grouting system

Safety analysis of the proposed Canadian geologic nuclear waste repository

International Nuclear Information System (INIS)

Prowse, D.R.

1977-01-01

The Canadian program for development and qualification of a geologic repository for emplacement of high-level and long-lived, alpha-emitting waste from irradiated nuclear fuel has been inititiated and is in its initial development stage. Fieldwork programs to locate candidate sites with suitable geological characteristics have begun. Laboratory studies and development of models for use in safety analysis of the emplaced nuclear waste have been initiated. The immediate objective is to complete a simplified safety analysis of a model geologic repository by mid-1978. This analysis will be progressively updated and will form part of an environmental Assessment Report of a Model Fuel Center which will be issued in mid-1979. The long-term objectives are to develop advanced safety assessment models of a geologic repository which will be available by 1980

Design of a nuclear-waste package for emplacement in tuff

International Nuclear Information System (INIS)

O'Neal, W.C.; Rothman, A.J.; Gregg, D.W.; Hockman, J.N.; Revelli, M.A.; Russell, E.W.; Schornhorst, J.R.

1983-01-01

Design, modeling, and testing activities are under way at LLNL in the development of high level nuclear waste package designs. We discuss the geological characteristics affecting design, the 10CFR60 design requirements, conceptual designs, metals for containment barriers, economic analysis, thermal modeling, and performance modeling

Defense waste salt disposal at the Savannah River Plant

International Nuclear Information System (INIS)

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

1984-01-01

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

Disposal costs for SRP high-level wastes in borosilicate glass and crystalline ceramic waste forms

International Nuclear Information System (INIS)

Rozsa, R.B.; Campbell, J.H.

1982-01-01

Purpose of this document is to compare and contrast the overall burial costs of the glass and ceramic waste forms, including processing, storage, transportation, packaging, and emplacement in a repository. Amount of waste will require approximately 10,300 standard (24 in. i.d. x 9-5/6 ft length) canisters of waste glass, each containing about 3260 lb of waste at 28% waste loading. The ceramic waste form requires about one-third the above number of standard canisters. Approximately $2.5 billion is required to process and dispose of this waste, and the total cost is independent of waste form (glass or ceramic). The major cost items (about 80% of the total cost) for all cases are capital and operating expenses. The capital and 20-year operating costs for the processing facility are the same order of magnitude, and their sum ranges from about one-half of the total for the reference glass case to two-thirds of the total for the ceramic cases

Overview of the waste/barrier/rock interactions program of the basalt waste isolation project

International Nuclear Information System (INIS)

Salter, P.F.; Burnell, J.R.; Lane, D.L.

1986-01-01

The waste package waste/barrier/rock interactions testing program of the Basalt Waste Isolation Project is designed to assess the interactions between nuclear waste forms, other waste package components, and the environment in order to evaluate long-term waste package isolation (radionuclide release) behavior. The program involves reacting fully radioactive waste forms with combinations of steel or copper container material and basalt/bentonite packing material in site-specific ground water under anticipated repository conditions to obtain the steady state radionuclide concentrations required to predictively model waste package radionuclide concentrations required to predictively model waste package radionuclide releases. Both static and flow-through autoclaves are being used in the test program to determine radionuclide concentrations as a function of time and groundwater flow rate, and to evaluate the solid phase and water chemistry changes that control those concentrations. This test program, when combined with project hydrologic and geochemical testing and modeling efforts, and natural analog studies, provides the information required to evaluate long-term radionuclide mobility within a waste package emplaced in a basalt repository

Solar energy emplacement developer

Science.gov (United States)

Mortensen, Michael; Sauls, Bob

1991-01-01

A preliminary design was developed for a Lunar Power System (LPS) composed of photovoltaic arrays and microwave reflectors fabricated from lunar materials. The LPS will collect solar energy on the surface of the Moon, transform it into microwave energy, and beam it back to Earth where it will be converted into usable energy. The Solar Energy Emplacement Developer (SEED) proposed will use a similar sort of solar energy collection and dispersement to power the systems that will construct the LPS.

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

Ventilation Model and Analysis Report

International Nuclear Information System (INIS)

Chipman, V.

2003-01-01

This model and analysis report develops, validates, and implements a conceptual model for heat transfer in and around a ventilated emplacement drift. This conceptual model includes thermal radiation between the waste package and the drift wall, convection from the waste package and drift wall surfaces into the flowing air, and conduction in the surrounding host rock. These heat transfer processes are coupled and vary both temporally and spatially, so numerical and analytical methods are used to implement the mathematical equations which describe the conceptual model. These numerical and analytical methods predict the transient response of the system, at the drift scale, in terms of spatially varying temperatures and ventilation efficiencies. The ventilation efficiency describes the effectiveness of the ventilation process in removing radionuclide decay heat from the drift environment. An alternative conceptual model is also developed which evaluates the influence of water and water vapor mass transport on the ventilation efficiency. These effects are described using analytical methods which bound the contribution of latent heat to the system, quantify the effects of varying degrees of host rock saturation (and hence host rock thermal conductivity) on the ventilation efficiency, and evaluate the effects of vapor and enhanced vapor diffusion on the host rock thermal conductivity

ENGINEERED BARRIER SYSTEM FEATURES, EVENTS AND PROCESSES

Energy Technology Data Exchange (ETDEWEB)

Jaros, W.

2005-08-30

The purpose of this report is to evaluate and document the inclusion or exclusion of engineered barrier system (EBS) features, events, and processes (FEPs) with respect to models and analyses used to support the total system performance assessment for the license application (TSPA-LA). A screening decision, either Included or Excluded, is given for each FEP along with the technical basis for exclusion screening decisions. This information is required by the U.S. Nuclear Regulatory Commission (NRC) at 10 CFR 63.114 (d, e, and f) [DIRS 173273]. The FEPs addressed in this report deal with those features, events, and processes relevant to the EBS focusing mainly on those components and conditions exterior to the waste package and within the rock mass surrounding emplacement drifts. The components of the EBS are the drip shield, waste package, waste form, cladding, emplacement pallet, emplacement drift excavated opening (also referred to as drift opening in this report), and invert. FEPs specific to the waste package, cladding, and drip shield are addressed in separate FEP reports: for example, ''Screening of Features, Events, and Processes in Drip Shield and Waste Package Degradation'' (BSC 2005 [DIRS 174995]), ''Clad Degradation--FEPs Screening Arguments (BSC 2004 [DIRS 170019]), and Waste-Form Features, Events, and Processes'' (BSC 2004 [DIRS 170020]). For included FEPs, this report summarizes the implementation of the FEP in the TSPA-LA (i.e., how the FEP is included). For excluded FEPs, this analysis provides the technical basis for exclusion from TSPA-LA (i.e., why the FEP is excluded). This report also documents changes to the EBS FEPs list that have occurred since the previous versions of this report. These changes have resulted due to a reevaluation of the FEPs for TSPA-LA as identified in Section 1.2 of this report and described in more detail in Section 6.1.1. This revision addresses updates in Yucca Mountain Project

United States National Waste Terminal Storage argillaceous rock studies

International Nuclear Information System (INIS)

Brunton, G.D.

1981-01-01

The past and present argillaceous rock studies for the US National Waste Terminal Storage Program consist of: (1) evaluation of the geological characteristics of several widespread argillaceous formations in the United States; (2) laboratory studies of the physical and chemical properties of selected argillaceous rock samples; and (3) two full-scale in situ surface heater experiments that simulate the emplacement of heat-generating radioactive waste in argillaceous rock

United States National Waste Terminal Storage argillaceous rock studies

International Nuclear Information System (INIS)

Brunton, G.D.

1979-01-01

The past and present argillaceous rock studies for the US National Waste Terminal Storage Program consist of: (1) evaluation of the geological characteristics of several widespread argillaceous formations in the United States; (2) laboratory studies of the physical and chemical properties of selected argillaceous rock samples; and (3) two full-scale in-situ surface heater experiments that simulate the emplacement of heat-generating radioactive waste in argillaceous rock

Extreme scenarios for nuclear waste repositories

Energy Technology Data Exchange (ETDEWEB)

Brown, M J [Harvard Univ., Cambridge, MA (USA). Div. of Applied Sciences; Crouch, E [Harvard Univ., Cambridge, MA (USA). Energy and Environmental Policy Center

1982-09-01

Two extreme scenarios for release of radioactive waste have been constructed. In the first, a volcanic eruption releases 1 km/sup 2/ of an underground nuclear waste repository, while in the second, waste enters the drinking water reservoir of a major city. With pessimistic assumptions, upper bounds on the number of cancers due to radiation are calculated. In the volcano scenario, the effects of the waste are smaller than the effects of natural radioactivity in the volcanic dust if the delay between emplacement and eruption exceeds 2000 yr. The consequences of the waste in drinking water depend on the survival time of the canisters and the rate of leaching of the nuclides from the waste matrix. For a canister life of 400 yr and a leach time of 6300 yr the cancer rate in the affected area would increase by 25%.

Waste Isolation Pilot Plant Safety Analysis Report

International Nuclear Information System (INIS)

1995-11-01

The following provides a summary of the specific issues addressed in this FY-95 Annual Update as they relate to the CH TRU safety bases: Executive Summary; Site Characteristics; Principal Design and Safety Criteria; Facility Design and Operation; Hazards and Accident Analysis; Derivation of Technical Safety Requirements; Radiological and Hazardous Material Protection; Institutional Programs; Quality Assurance; and Decontamination and Decommissioning. The System Design Descriptions'' (SDDS) for the WIPP were reviewed and incorporated into Chapter 3, Principal Design and Safety Criteria and Chapter 4, Facility Design and Operation. This provides the most currently available final engineering design information on waste emplacement operations throughout the disposal phase up to the point of permanent closure. Also, the criteria which define the TRU waste to be accepted for disposal at the WIPP facility were summarized in Chapter 3 based on the WAC for the Waste Isolation Pilot Plant.'' This Safety Analysis Report (SAR) documents the safety analyses that develop and evaluate the adequacy of the Waste Isolation Pilot Plant Contact-Handled Transuranic Wastes (WIPP CH TRU) safety bases necessary to ensure the safety of workers, the public and the environment from the hazards posed by WIPP waste handling and emplacement operations during the disposal phase and hazards associated with the decommissioning and decontamination phase. The analyses of the hazards associated with the long-term (10,000 year) disposal of TRU and TRU mixed waste, and demonstration of compliance with the requirements of 40 CFR 191, Subpart B and 40 CFR 268.6 will be addressed in detail in the WIPP Final Certification Application scheduled for submittal in October 1996 (40 CFR 191) and the No-Migration Variance Petition (40 CFR 268.6) scheduled for submittal in June 1996. Section 5.4, Long-Term Waste Isolation Assessment summarizes the current status of the assessment

Waste Isolation Pilot Plant Safety Analysis Report

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-11-01

The following provides a summary of the specific issues addressed in this FY-95 Annual Update as they relate to the CH TRU safety bases: Executive Summary; Site Characteristics; Principal Design and Safety Criteria; Facility Design and Operation; Hazards and Accident Analysis; Derivation of Technical Safety Requirements; Radiological and Hazardous Material Protection; Institutional Programs; Quality Assurance; and Decontamination and Decommissioning. The System Design Descriptions`` (SDDS) for the WIPP were reviewed and incorporated into Chapter 3, Principal Design and Safety Criteria and Chapter 4, Facility Design and Operation. This provides the most currently available final engineering design information on waste emplacement operations throughout the disposal phase up to the point of permanent closure. Also, the criteria which define the TRU waste to be accepted for disposal at the WIPP facility were summarized in Chapter 3 based on the WAC for the Waste Isolation Pilot Plant.`` This Safety Analysis Report (SAR) documents the safety analyses that develop and evaluate the adequacy of the Waste Isolation Pilot Plant Contact-Handled Transuranic Wastes (WIPP CH TRU) safety bases necessary to ensure the safety of workers, the public and the environment from the hazards posed by WIPP waste handling and emplacement operations during the disposal phase and hazards associated with the decommissioning and decontamination phase. The analyses of the hazards associated with the long-term (10,000 year) disposal of TRU and TRU mixed waste, and demonstration of compliance with the requirements of 40 CFR 191, Subpart B and 40 CFR 268.6 will be addressed in detail in the WIPP Final Certification Application scheduled for submittal in October 1996 (40 CFR 191) and the No-Migration Variance Petition (40 CFR 268.6) scheduled for submittal in June 1996. Section 5.4, Long-Term Waste Isolation Assessment summarizes the current status of the assessment.

Discrete fracture network modelling of a KBS-3H repository at Olkiluoto

Energy Technology Data Exchange (ETDEWEB)

Lanyon, G.W. (Fracture Systems Ltd, St Ives (United Kingdom)); Marschall, P. (Nagra, Wettingen (Switzerland))

2008-06-15

This report presents Discrete Fracture Network (DFN) models of groundwater flow around a KBS-3H repository situated at Olkiluoto. The study was performed in support of the Safety Case for the KBS-3H Concept, being jointly studied by SKB and Posiva. As part of the preliminary assessment of long term safety of a KBS-3H repository, a Process Report and an Evolution Report (evolution of the disposal system from the emplacement of the first canister to the long term) are being produced. In the course of the task definition the project team identified the need for complementary modelling studies aimed at increasing insight into the hydrodynamic evolution of the disposal system after waste emplacement. In particular, the following issues were identified as requiring input from hydrodynamic models: Probability of high inflow points which may cause buffer erosion. Time transients of inflows after construction of deposition drifts. Interference between deposition drifts and transport tunnels. The DFN models represent the fault and fracture system in the planned repository volume at Olkiluoto. In particular, they represent the hydro geologically significant features. The types of hydrogeological features included in the models are: Major Fracture Zones (MFZs). Local Fracture Zones (LFZs) and associated water conducting features (LFZ-WCFs). Water Conducting Features in the background rock (BR-WCFs). These feature types are derived from the current geological and hydrogeological interpretations developed by Posiva. Several model variants were developed during the study and these variants were used for geometric simulations of the WCF network around the deposition drifts. A simple layout adaptation scheme has been applied to the network models to derive statistics for performance measures relating to the deposition drifts, compartments, plugs and super-containers. A single fracture transient flow model was developed to provide insight to transient flow behaviour around

Discrete fracture network modelling of a KBS-3H repository at Olkiluoto

International Nuclear Information System (INIS)

Lanyon, G.W.; Marschall, P.

2008-06-01

This report presents Discrete Fracture Network (DFN) models of groundwater flow around a KBS-3H repository situated at Olkiluoto. The study was performed in support of the Safety Case for the KBS-3H Concept, being jointly studied by SKB and Posiva. As part of the preliminary assessment of long term safety of a KBS-3H repository, a Process Report and an Evolution Report (evolution of the disposal system from the emplacement of the first canister to the long term) are being produced. In the course of the task definition the project team identified the need for complementary modelling studies aimed at increasing insight into the hydrodynamic evolution of the disposal system after waste emplacement. In particular, the following issues were identified as requiring input from hydrodynamic models: Probability of high inflow points which may cause buffer erosion. Time transients of inflows after construction of deposition drifts. Interference between deposition drifts and transport tunnels. The DFN models represent the fault and fracture system in the planned repository volume at Olkiluoto. In particular, they represent the hydro geologically significant features. The types of hydrogeological features included in the models are: Major Fracture Zones (MFZs). Local Fracture Zones (LFZs) and associated water conducting features (LFZ-WCFs). Water Conducting Features in the background rock (BR-WCFs). These feature types are derived from the current geological and hydrogeological interpretations developed by Posiva. Several model variants were developed during the study and these variants were used for geometric simulations of the WCF network around the deposition drifts. A simple layout adaptation scheme has been applied to the network models to derive statistics for performance measures relating to the deposition drifts, compartments, plugs and super-containers. A single fracture transient flow model was developed to provide insight to transient flow behaviour around

Radiological protection criteria risk assessments for waste disposal options

International Nuclear Information System (INIS)

Hill, M.D.

1982-01-01

Radiological protection criteria for waste disposal options are currently being developed at the National Radiological Protection Board (NRPB), and, in parallel, methodologies to be used in assessing the radiological impact of these options are being evolved. The criteria and methodologies under development are intended to apply to all solid radioactive wastes, including the high-level waste arising from reprocessing of spent nuclear fuel (because this waste will be solidified prior to disposal) and gaseous or liquid wastes which have been converted to solid form. It is envisaged that the same criteria will be applied to all solid waste disposal options, including shallow land burial, emplacement on the ocean bed (sea dumping), geological disposal on land and sub-seabed disposal

Reconsolidation of salt as applied to permanent seals for the Waste Isolation Pilot Plant

International Nuclear Information System (INIS)

Hansen, F.D.; Callahan, G.D.; Van Sembeek, L.L.

1993-01-01

Reconsolidated salt is a fundamental component of the permanent seals for the Waste Isolation Pilot Plant. As regulations are currently understood and seal concepts envisioned, emplaced salt is the sole long-term seal component designed to prevent the shafts from becoming preferred pathways for rating gases or liquids. Studies under way in support of the sealing function of emplaced salt include laboratory testing of crushed salt small-scale in situ tests, constitutive modeling of crushed salt, calculations of the opening responses during operation and closure, and design practicalities including emplacement techniques. This paper briefly summarizes aspects of these efforts and key areas of future work

Spent fuel test-climax: a test of geologic storage of high-level waste in granite

International Nuclear Information System (INIS)

Ramspott, L.D.; Ballou, L.B.; Patrick, W.C.

1981-01-01

A test of retrievable geologic storage of spent fuel assemblies from an operating commercial nuclear reactor is underway at the Nevada Test Site (NTS) of the US Department of Energy. This generic test is located 420 m below the surface in the Climax granitic stock. Eleven canisters of spent fuel approximately 2.5 years out of reactor core (about 1.6 kW/canister thermal output) were emplaced in a storage drift along with 6 electrical simulator canisters. Two adjacent drifts contain electrical heaters, which are operated to simulate within the test array the thermal field of a large repository. Fuel was loaded during April to May 1980 and initial results of the test will be presented

Disposal of radioactive wastes by UK NIREX Ltd

International Nuclear Information System (INIS)

Ginniff, M.E.

1989-01-01

In the United Kingdom UK Nirex Ltd., provides a comprehensive, long-term radioactive waste disposal service for low and intermediate level solid radioactive wastes arising from all radioactive operations in the country. The high level wastes which are not the responsibility of Nirex, are to be vitrified and stored for some 50 years. The low and intermediate wastes are to be emplaced in a deep underground repository and the developments during 1988 towards this objective are presented. Following the publication of a widely circulated consultation document entitled 'The Way Forward', design studies and site selection exercises for a deep underground repository were started. (author)

Self-shielding flex-circuit drift tube, drift tube assembly and method of making

Science.gov (United States)

Jones, David Alexander

2016-04-26

The present disclosure is directed to an ion mobility drift tube fabricated using flex-circuit technology in which every other drift electrode is on a different layer of the flex-circuit and each drift electrode partially overlaps the adjacent electrodes on the other layer. This results in a self-shielding effect where the drift electrodes themselves shield the interior of the drift tube from unwanted electro-magnetic noise. In addition, this drift tube can be manufactured with an integral flex-heater for temperature control. This design will significantly improve the noise immunity, size, weight, and power requirements of hand-held ion mobility systems such as those used for explosive detection.

Drift chamber detectors

International Nuclear Information System (INIS)

Duran, I.; Martinez Laso, L.

1989-01-01

A review of High Energy Physics detectors based on drift chambers is presented. The ionization, drift diffusion, multiplication and detection principles are described. Most common drift media are analysied, and a classification of the detectors according to its geometry is done. Finally the standard read-out methods are displayed and the limits of the spatial resolution are discussed. (Author)

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

Construction update and drift velocity calibration for the CLAS drift chamber system

International Nuclear Information System (INIS)

Mestayer, M.D.; Barbosa, F.J.; Bonneau, P.; Burtin, E.; Christo, S.; Doolittle, G.; Dytman, S.A.; Gilfoyle, G.P.; Hyde-Wright, C.E.; Klein, A.; Kossov, M.V.; Kuhn, S.E.; Magahiz, R.; Miskimen, R.A.; Murphy, L.Y.; O'Meara, J.E.; Pyron, T.D.; Qin, L.; Raue, B.A.; Schumacher, R.A.; Tuzel, W.; Weinstein, L.B.; Yegneswaran, A.

1995-01-01

We briefly describe the drift chamber system for the CLAS detector at CEBAF, concentrating on the method which will be used to calibrate the drift velocity function. We identify key features of the function which should apply to any small-cell drift chamber geometry in which the cathode and anode surfaces are wires. Using these ideas, we describe a simple method to compensate for variations in the drift velocity function due to environmental changes. (orig.)

Construction update and drift velocity calibration for the CLAS drift chamber system

Energy Technology Data Exchange (ETDEWEB)

Mestayer, M.D. [Continuous Electron Beam Accelerator Facility, Newport News, VA (United States); Barbosa, F.J. [Continuous Electron Beam Accelerator Facility, Newport News, VA (United States); Bonneau, P. [Continuous Electron Beam Accelerator Facility, Newport News, VA (United States); Burtin, E. [University of South Carolina, Columbia, SC (United States); Christo, S. [Continuous Electron Beam Accelerator Facility, Newport News, VA (United States); Doolittle, G. [Continuous Electron Beam Accelerator Facility, Newport News, VA (United States); Dytman, S.A. [University of Pittsburg, Pittsburg, PA (United States); Gilfoyle, G.P. [University of Richmond, Richmond, VA (United States); Hyde-Wright, C.E. [Old Dominion University, Norfolk, VA (United States); Klein, A. [Old Dominion University, Norfolk, VA (United States); Kossov, M.V. [Christopher Newport University, Newport News, VA (United States); Kuhn, S.E. [Old Dominion University, Norfolk, VA (United States); Magahiz, R. [Carnegie-Mellon Univ., Pittsburgh, PA (United States); Miskimen, R.A. [University of Massachussetts, Amherst, MA (United States); Murphy, L.Y. [CE Saclay, Gif sur Yvette (France); O`Meara, J.E. [Continuous Electron Beam Accelerator Facility, Newport News, VA (United States); Pyron, T.D. [Old Dominion University, Norfolk, VA (United States); Qin, L. [Old Dominion University, Norfolk, VA (United States); Raue, B.A. [Old Dominion University, Norfolk, VA (United States); Schumacher, R.A. [Carnegie-Mellon Univ., Pittsburgh, PA (United States); Tuzel, W. [Continuous Electron Beam Accelerator Facility, Newport News, VA (United States); Weinstein, L.B. [Continuous Electron Beam Accelerator Facility, Newport News, VA (United States); Yegneswaran, A. [Continuous Electron Beam Accelerator Facility, Newport News, VA (United States)

1995-12-11

We briefly describe the drift chamber system for the CLAS detector at CEBAF, concentrating on the method which will be used to calibrate the drift velocity function. We identify key features of the function which should apply to any small-cell drift chamber geometry in which the cathode and anode surfaces are wires. Using these ideas, we describe a simple method to compensate for variations in the drift velocity function due to environmental changes. (orig.).

Salt brickwork as long-term sealing in salt formations

International Nuclear Information System (INIS)

Walter, F.; Yaramanci, U.

1993-01-01

Radioactive wastes can be disposed of in deep salt formations. Rock salt is a suitable geologic medium because of its unique characteristics. Open boreholes, shafts and drifts are created to provide physical access to the repository. Long-term seals must be emplaced in these potential pathways to prevent radioactive release into the biosphere. The sealing materials must be mechanically and, most important, geochemically stable within the host rock. Salt bricks made from compressed salt-powder are understood to be the first choice long-term sealing material. Seals built of salt bricks will be ductile. Large sealing systems are built by combining the individual bricks with mortar. Raw materials for mortar are fine-grained halite powder and ground saliferous clay. This provides for the good adhesive strength of the mortar to the bricks and the high shear-strength of the mortar itself. To test the interaction of rock salt with an emplaced long-term seal, experiments will be carried out in situ, in the Asse salt mine in Germany. Simple borehole sealing experiments will be performed in horizontal holes and a complicated drift sealing experiment is planned, to demonstrate the technology of sealing a standard size drift or shaft inside a disturbed rock mass. Especially, the mechanical stability of the sealing system has to be demonstrated

Coupled THM processes in EDZ of crystalline rocks using an elasto-plastic cellular automaton

Science.gov (United States)

Pan, Peng-Zhi; Feng, Xia-Ting; Huang, Xiao-Hua; Cui, Qiang; Zhou, Hui

2009-05-01

This paper aims at a numerical study of coupled thermal, hydrological and mechanical processes in the excavation disturbed zones (EDZ) around nuclear waste emplacement drifts in fractured crystalline rocks. The study was conducted for two model domains close to an emplacement tunnel; (1) a near-field domain and (2) a smaller wall-block domain. Goodman element and weak element were used to represent the fractures in the rock mass and the rock matrix was represented as elasto-visco-plastic material. Mohr-Coulomb criterion and a non-associated plastic flow rule were adopted to consider the viscoplastic deformation in the EDZ. A relation between volumetric strain and permeability was established. Using a self-developed EPCA2D code, the elastic, elasto-plastic and creep analyses to study the evolution of stress and deformations, as well as failure and permeability evolution in the EDZ were conducted. Results indicate a strong impact of fractures, plastic deformation and time effects on the behavior of EDZ especially the evolution of permeability around the drift.

Drift Chambers detectors

International Nuclear Information System (INIS)

Duran, I.; Martinez laso, L.

1989-01-01

We present here a review of High Energy Physics detectors based on drift chambers. The ionization, drift diffusion, multiplication and detection principles are described. Most common drift media are analysed, and a classification of the detectors according to its geometry is done. Finally the standard read-out methods are displayed and the limits of the spatial resolution are discussed. (Author) 115 refs

Permeability of granular beds emplaced in vertical drill holes

International Nuclear Information System (INIS)

Griffiths, S.K.; Morrison, F.A. Jr.

1979-01-01

To determine the permeabilities of granular materials emplaced in vertical drill holes used for underground nuclear tests, an experiment at the USDOE Nevada Test Site (NTS) was conducted. As the hole is being filled, falling material increases pressure above and within the granular beds beneath. When the filling operation starts or stops, a transient pressure response occurs within the beds; measurements of this response in beds of various compositions were made. The permeabilities after emplacement were found by matching analytical predictions of the response to these data. This information is useful in assuring the containment of nuclear tests conducted in such drill holes

Sampling of Stochastic Input Parameters for Rockfall Calculations and for Structural Response Calculations Under Vibratory Ground Motion

International Nuclear Information System (INIS)

M. Gross

2004-01-01

The purpose of this scientific analysis is to define the sampled values of stochastic (random) input parameters for (1) rockfall calculations in the lithophysal and nonlithophysal zones under vibratory ground motions, and (2) structural response calculations for the drip shield and waste package under vibratory ground motions. This analysis supplies: (1) Sampled values of ground motion time history and synthetic fracture pattern for analysis of rockfall in emplacement drifts in nonlithophysal rock (Section 6.3 of ''Drift Degradation Analysis'', BSC 2004 [DIRS 166107]); (2) Sampled values of ground motion time history and rock mechanical properties category for analysis of rockfall in emplacement drifts in lithophysal rock (Section 6.4 of ''Drift Degradation Analysis'', BSC 2004 [DIRS 166107]); (3) Sampled values of ground motion time history and metal to metal and metal to rock friction coefficient for analysis of waste package and drip shield damage to vibratory motion in ''Structural Calculations of Waste Package Exposed to Vibratory Ground Motion'' (BSC 2004 [DIRS 167083]) and in ''Structural Calculations of Drip Shield Exposed to Vibratory Ground Motion'' (BSC 2003 [DIRS 163425]). The sampled values are indices representing the number of ground motion time histories, number of fracture patterns and rock mass properties categories. These indices are translated into actual values within the respective analysis and model reports or calculations. This report identifies the uncertain parameters and documents the sampled values for these parameters. The sampled values are determined by GoldSim V6.04.007 [DIRS 151202] calculations using appropriate distribution types and parameter ranges. No software development or model development was required for these calculations. The calculation of the sampled values allows parameter uncertainty to be incorporated into the rockfall and structural response calculations that support development of the seismic scenario for the

Waste form dissolution in bedded salt

International Nuclear Information System (INIS)

Kaufman, A.M.

1980-01-01

A model was devised for waste dissolution in bedded salt, a hydrologically tight medium. For a typical Spent UnReprocessed Fuel (SURF) emplacement, the dissolution rate wll be diffusion limited and will rise to a steady state value after t/sub eq/ approx. = 250 (1+(1-epsilon 0 ) K/sub D//epsilon 0 ) (years) epsilon 0 is the overpack porosity and K/sub d/ is the overpack sorption coefficient. The steady state dissolution rate itself is dominated by the solubility of UO 2 . Steady state rates between 5 x 10 -5 and .5 (g/year) are achievable by SURF emplacements in bedded salt without overpack, and rates between 5 x 10 -7 and 5 x 10 -3 (g/year) with an overpack having porosity of 10 -2

A perspective on the management of low-level radioactive waste

International Nuclear Information System (INIS)

Champ, D.R.; Charlesworth, D.H.

1994-01-01

In Canada, low-level radioactive waste (LLRW) is defined as all radioactive waste except spent fuel waste and tailings. At the time of the conference, the current practice was storage, but programs are underway to dispose of LLRW. AECL has applied for licensing of an intrusion-resistant underground structure. A comprehensive approach to LLRW management calls for: waste stream identification, waste characterization, waste segregation and characterization, waste processing, waste emplacement (storage or disposal); general principles are discussed under these headings. Performance assessment of disposal involves mathematical modelling. Progress has been slow, so if the Canadian nuclear industry does not eventually decide on a joint strategy for LLRW disposal, the federal government may have to impose a solution. 10 refs., 2 figs

Acceptance criteria considerations for miscellaneous wastes

International Nuclear Information System (INIS)

Irvine, A.R.; Forsberg, C.W.

1987-01-01

EPA standards set forth limitations regarding releases to the accessible environment adjacent to a geologic repository. The NRC criteria pertaining to waste form and engineered barrier performance place certain restrictions on the physical and chemical nature of the waste form and require substantially complete confinement of radioactivity until the high-heat-production period is past. After this period, the annual release of radionuclides from the waste package is normally limited to 1 part in 100,000 of the amounts calculated to be present at 1000-y decay. The regulation permits deviation from these criteria in exceptional circumstances. One such circumstance might be the absence of a significant perturbation in temperature around the stored waste. The lack of significant heat release will eliminate the hydrologic driving force for dispersal of radionuclides. Exceptional circumstances which potentially could justify a less stringent long-term release criterion are: small quantity of radioactivity, the nature of the radioactive species, and the nature of the geology in which the waste is to be emplaced. Because the MW after a suitable decay period have low heat release rates per unit volume, they apparently could be so emplaced in a repository that there would be no compelling need, according to the reasoning presented in 10 CFR 60, for a 1000-y container. Regarding attainment of the specified long-term release rate criterion, neither the solubility limits for the various waste forms nor the conductance of potential migration barriers are currently adequately characterized. The relatively small total heat generation rate for the MW in combination with the usual low volumetric heat generation rate apparently will allow application of migration barriers in a low temperature environment where barrier performance would be expected to be unchanged with time

Geological Disposal of Radioactive Waste: Technological Implications for Retrievability

International Nuclear Information System (INIS)

2009-01-01

Various IAEA Member States are discussing whether and to what degree reversibility (including retrievability) might be built into management strategies for radioactive waste. This is particularly the case in relation to the disposal of long lived and/or high level waste and spent nuclear fuel (SNF) in geological repositories. It is generally accepted that such repositories should be designed to be passively safe with no intention of retrieving the waste. Nevertheless, various reasons have been advanced for including the concept of reversibility and the ability to retrieve the emplaced wastes in the disposal strategy. The intention is to increase the level of flexibility and to provide the ability to cope with, or to benefit from, new technical advances in waste management and materials technologies, and to respond to changing social, economic and political opinion. The technological implications of retrievability in geological disposal concepts are explored in this report. Scenarios for retrieving emplaced waste packages are considered and the report aims to identify and describe any related technological provisions that should be incorporated into the design, construction, operational and closure phases of the repository. This is based on a number of reference concepts for the geological disposal of radioactive waste (including SNF) which are currently being developed in Member States with advanced development programmes. The report begins with a brief overview of various repository concepts, starting with a summary of the types of radioactive waste that are typically considered for deep geological disposal. The main host rocks considered are igneous crystalline and volcanic rocks, argillaceous clay rocks and salts. The typical design features of repositories are provided with a description of repository layouts, an overview of the key features of the major repository components, comprising the waste package, the emplacement cells and repository access facilities

Evaluation of design feature No.20 -- Ground support options

International Nuclear Information System (INIS)

Duan, F.

2000-01-01

Ground support options are primarily evaluated for emplacement drifts while ground support systems for non-emplacement openings such as access mains and ventilation drifts are not evaluated against LADS evaluation criteria in this report. Considerations include functional requirements for ground support, the use of a steel-lined system, and the feasibility of using an unlined ground support system principally with grouted rock bolts for permanent ground support. The feature evaluation also emphasizes the postclosure effects of ground support materials on waste isolation and the preclosure aspects such as durability, maintainability, constructibility, safety, engineering acceptability, and cost. This evaluation is to: (A) Review the existing analyses, reports, and studies regarding this design feature, and compile relevant information on performance characteristics. (B) Develop an appropriate evaluation approach for evaluating ground support options against evaluation criteria provided by the LADS team. (C) Evaluate ground support options not only for their preclosure performance in terms of drift stability, material durability, maintenance, constructibility, and cost, but also for their postclosure performance in terms of chemical effects of ground support materials (i.e., concrete, steel) on waste isolation and radionuclide transport. Specifically, the scope for ground support options evaluation include: (1) all steel-lined drifts (no cementitious materials), (2) unlined drifts with minimum cementitious materials (e.g., grout for rockbolts), and (3) concrete-lined drifts, with the focus on the postclosure acceptability evaluation. In addition, unlined drifts with zero cementitious materials (e.g., use of frictional bolts such as split sets, Swellex bolts) are briefly discussed. (D) Identify candidate ground support systems that have the potential to enhance the repository performance based on the feature evaluation. and (E) Provide conclusions and recommendations

Department of Energy Operational Readiness Review for the Waste Isolation Pilot Plant

Energy Technology Data Exchange (ETDEWEB)

None

2016-12-01

The U.S. Department of Energy (DOE) has completed an Operational Readiness Review (ORR) for the restart of Contact Handled (CH) waste emplacement at the Waste Isolation Pilot Plant (WIPP) located near Carlsbad, New Mexico. The ORR team assessed the readiness of Nuclear Waste Partnership, LLC (NWP) to manage and perform receipt through CH waste emplacement, and associated waste handling and management activities, including the ability of the National TRU Program (NTP) to evaluate the waste currently stored at the WIPP site against the revised and enhanced Waste Acceptance Criteria (WAC). Field work for this review began on November 14, 2015 and was completed on November 30, 2016. The DOE ORR was conducted in accordance with the Department of Energy Operational Readiness Review Implementation Plan for the Waste Isolation Pilot Plant, dated November 8, 2016, and DOE Order 425.1D, Verification of Readiness to Start Up or Restart Nuclear Facilities. The review activities included personnel interviews, record reviews, direct observation of operations and maintenance demonstrations, and observation of multiple operational and emergency drills/exercises. The DOE ORR also evaluated the adequacy of the contractor’s ORR (CORR) and the readiness of the DOE Carlsbad field Office (CBFO) to oversee the startup and execution of CH waste emplacement activities at the WIPP facility. The WIPP facility is categorized as a Hazard Category 2 DOE Nonreactor Nuclear Facility for all surface and Underground (UG) operations per DOE-STD-1027-92, Hazard Categorization and Accident Analysis Techniques for Compliance with DOE Order 5480.23, Nuclear Safety Analysis Reports. In addition, the WIPP experienced two events in February, 2014 that resulted in Accident Investigations being performed in accordance with the requirements of DOE Order 225.1B, Accident Investigations. Based upon the results of the accident investigations and hazard categorization of the facility, the team placed

Extreme scenarios for nuclear waste repositories

Energy Technology Data Exchange (ETDEWEB)

Brown, M J; Crouch, E

1982-09-01

Two extreme scenarios for release of radioactive waste have been constructed. In the first, a volcanic eruption releases 1 km2 of an underground nuclear waste repository, while in the second, waste enters the drinking water reservoir of a major city. With pessimistic assumptions, upper bounds on the number of cancers due to radiation are calculated. In the volcano scenario, the effects of the water are smaller than the effects of natural radioactivity in the volcanic dust if the delay between emplacement and eruption exceeds 2000 yr. The consequences of the waste in drinking water depend on the survival time of the canisters and the rate of leaching of the nuclides from the waste matrix. For a canister life of 400 yr and a leach time of 6300 yr the cancer rate in the affected area would increase by 25%.

Multiscale Thermohydrologic Model

Energy Technology Data Exchange (ETDEWEB)

T. Buscheck

2004-10-12

The purpose of the multiscale thermohydrologic model (MSTHM) is to predict the possible range of thermal-hydrologic conditions, resulting from uncertainty and variability, in the repository emplacement drifts, including the invert, and in the adjoining host rock for the repository at Yucca Mountain. Thus, the goal is to predict the range of possible thermal-hydrologic conditions across the repository; this is quite different from predicting a single expected thermal-hydrologic response. The MSTHM calculates the following thermal-hydrologic parameters: temperature, relative humidity, liquid-phase saturation, evaporation rate, air-mass fraction, gas-phase pressure, capillary pressure, and liquid- and gas-phase fluxes (Table 1-1). These thermal-hydrologic parameters are required to support ''Total System Performance Assessment (TSPA) Model/Analysis for the License Application'' (BSC 2004 [DIRS 168504]). The thermal-hydrologic parameters are determined as a function of position along each of the emplacement drifts and as a function of waste package type. These parameters are determined at various reference locations within the emplacement drifts, including the waste package and drip-shield surfaces and in the invert. The parameters are also determined at various defined locations in the adjoining host rock. The MSTHM uses data obtained from the data tracking numbers (DTNs) listed in Table 4.1-1. The majority of those DTNs were generated from the following analyses and model reports: (1) ''UZ Flow Model and Submodels'' (BSC 2004 [DIRS 169861]); (2) ''Development of Numerical Grids for UZ Flow and Transport Modeling'' (BSC 2004); (3) ''Calibrated Properties Model'' (BSC 2004 [DIRS 169857]); (4) ''Thermal Conductivity of the Potential Repository Horizon'' (BSC 2004 [DIRS 169854]); (5) ''Thermal Conductivity of the Non-Repository Lithostratigraphic Layers

Review of DOE waste package program. Subtask 1.1. National waste package program, April-September 1983. Volume 5

Energy Technology Data Exchange (ETDEWEB)

Soo, P. (ed.)

1984-08-01

The current effort is part of an ongoing task to review the national high-level waste package effort. It includes evaluations of reference waste form, container, and packing material components with respect to determining how they may contribute to the containment and controlled release of radionuclides after waste packages have been emplaced in salt, basalt, and tuff repositories. In the current Biannual Report a section on carbon steel container corrosion has been included to complement prior work on TiCode-12 and Type 304 stainless steel. The use of crushed tuff as a packing material is discussed and waste package component interaction test data are included. Licensing data requirements to estimate the degree of compliance with NRC performance objectives are specified. 41 figures, 24 tables.

Review of DOE waste package program. Subtask 1.1. National waste package program, April-September 1983. Volume 5

International Nuclear Information System (INIS)

Soo, P.

1984-08-01

The current effort is part of an ongoing task to review the national high-level waste package effort. It includes evaluations of reference waste form, container, and packing material components with respect to determining how they may contribute to the containment and controlled release of radionuclides after waste packages have been emplaced in salt, basalt, and tuff repositories. In the current Biannual Report a section on carbon steel container corrosion has been included to complement prior work on TiCode-12 and Type 304 stainless steel. The use of crushed tuff as a packing material is discussed and waste package component interaction test data are included. Licensing data requirements to estimate the degree of compliance with NRC performance objectives are specified. 41 figures, 24 tables

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

Reversibility and retrievability in geologic disposal of radioactive waste. A new Nea report

International Nuclear Information System (INIS)

Brown, P.A.; Pascatore, C.; Sumerling, T.

2001-01-01

Radioactive waste needs to be managed responsibly to ensure public safety and the protection of the environment, as well as security from unauthorized interference, now and in the future. One of the most challenging tasks is the management of long-lived radioactive waste that must be isolated from the human environment for many thousands, or even hundreds of thousands, of years. There is a consensus among the engaged technical community that engineered geologic disposal provides a safe and ethical method for the long term management of such waste. This method is also cited in the national policies of several countries as either a promising or appropriate method for dealing with long-lived radioactive waste. Engineered geologic disposal means emplacement of waste in repositories constructed deep underground in suitable geologic media. Thus the waste is contained, and safety assured by passive barriers with multiple safety functions, so that there is no need for any further actions by future generations. Primary principles of the engineered geologic disposal concept are that waste will only be emplaced in a repository when there is high confidence in the ultimate long-term safety, and that the long-term safety must not rely on actions following the closure of the repository. This does not mean, however, that actions cannot be taken. Most repository development programmes include the possibility of post-closure activities for security and monitoring purposes. (authors)

Design management and stress analysis of a circular rock tunnel and emplacement holes for storage of spent nuclear fuel

International Nuclear Information System (INIS)

Kandalaft-Ladkany, N.; Wyman, R.V.

1992-01-01

This paper discusses a critical path method (CPM) diagram and logic net which are used for the design cycle of the rock tunnel system for a high level nuclear waste repository. In the analysis the design tunnel is subjected to pre-existing temperature and overburden loads at time of construction. high thermal stresses develop later due to the long term influx of heat from the canisters stored in vertical emplacement holes. Results indicate that thermal stresses reach a critical level for the rock in the vicinity of the canisters which could lead to local collapse of the rock and damage to the canisters

Study on retrievability of waste package in geological disposal

International Nuclear Information System (INIS)

Hasegawa, Hiroshi; Noda, Masaru

2002-02-01

Retrievability of waste packages in geological disposal of high-level radioactive waste has been investigated from a technical aspect in various foreign countries, reflecting a social concern while retrievability is not provided as a technical requirement. This study investigates the concept of reversibility and retrievability in foreign countries and a technical feasibility on retrievability of waste packages in the geological disposal concept shown in the H12 report. The conclusion obtained through this study is as follows: 1. Concept of reversibility and retrievability in foreign countries. Many organizations have reconsidered the retrievability as one option in the geological disposal to improve the reversibility of the stepwise decision-making process and provide the flexibility, even based upon the principle of the geological disposal that retrieval of waste from the repository is not intended. 2. Technical feasibility on the retrievability in disposal concept in the H12 report. It is confirmed to be able to remove the buffer and to retrieve the waste packages by currently available technologies even after the stages following emplacement of the buffer. It must be noted that a large effort and expense would be required for some activities such as the reconstruction of access route if the activities started after a stage of backfilling disposal tunnels. 3. Evaluation of feasibility on the retrievability and extraction of the issues. In the near future, it is necessary to study and confirm the practical workability and economical efficiency for the retrieving method of waste packages proposed in this study, the handling and processing method of removed buffer materials, and the retrieving method of waste packages in the case of degrading the integrity of waste packages or not emplacing the waste packages in the assumed attitude, etc. (author)

Emplacement Gantry Gap Analysis Study

International Nuclear Information System (INIS)

Thornley, R.

2005-01-01

To date, the project has established important to safety (ITS) performance requirements for structures, systems, and components (SSCs) based on the identification and categorization of event sequences that may result in a radiological release. These performance requirements are defined within the ''Nuclear Safety Design Bases for License Application'' (NSDB) (BSC 2005 [DIRS 171512], Table A-11). Further, SSCs credited with performing safety functions are classified as ITS. In turn, assurance that these SSCs will perform as required is sought through the use of consensus codes and standards. This gap analysis is based on the design completed for license application only. Accordingly, identification of ITS SSCs beyond those defined within the NSDB are based on designs that may be subject to further development during detail design. Furthermore, several design alternatives may still be under consideration to satisfy certain safety functions, and final selection will not be determined until further design development has occurred. Therefore, for completeness, alternative designs currently under consideration will be discussed throughout this study. This gap analysis will evaluate each code and standard identified within the ''Emplacement Gantry ITS Standards Identification Study'' (BSC 2005 [DIRS 173586]) to ensure each ITS performance requirement is fully satisfied. When a performance requirement is not fully satisfied, a gap is highlighted. This study will identify requirements to supplement or augment the code or standard to meet performance requirements. Further, this gap analysis will identify nonstandard areas of the design that will be subject to a design development plan. Nonstandard components and nonstandard design configurations are defined as areas of the design that do not follow standard industry practices or codes and standards. Whereby, assurance that an SSC will perform as required may not be readily sought though the use of consensus standards. This

Modeling and Field Test Planning Activities in Support of Disposal of Heat-Generating Waste in Salt

Energy Technology Data Exchange (ETDEWEB)

Rutqvist, Jonny; Blanco Martin, Laura; Mukhopadhyay, Sumit; Houseworth, Jim; Birkholzer, Jens

2014-09-26

The modeling efforts in support of the field test planning conducted at LBNL leverage on recent developments of tools for modeling coupled thermal-hydrological-mechanical-chemical (THMC) processes in salt and their effect on brine migration at high temperatures. This work includes development related to, and implementation of, essential capabilities, as well as testing the model against relevant information and published experimental data related to the fate and transport of water. These are modeling capabilities that will be suitable for assisting in the design of field experiment, especially related to multiphase flow processes coupled with mechanical deformations, at high temperature. In this report, we first examine previous generic repository modeling results, focusing on the first 20 years to investigate the expected evolution of the different processes that could be monitored in a full-scale heater experiment, and then present new results from ongoing modeling of the Thermal Simulation for Drift Emplacement (TSDE) experiment, a heater experiment on the in-drift emplacement concept at the Asse Mine, Germany, and provide an update on the ongoing model developments for modeling brine migration. LBNL also supported field test planning activities via contributions to and technical review of framework documents and test plans, as well as participation in workshops associated with field test planning.

Time dependent drift Hamiltonian

International Nuclear Information System (INIS)

Boozer, A.H.

1982-04-01

The motion of individual charged particles in a given magnetic and an electric fields is discussed. An idea of a guiding center distribution function f is introduced. The guiding center distribution function is connected to the asymptotic Hamiltonian through the drift kinetic equation. The general non-stochastic magnetic field can be written in a contravariant and a covariant forms. The drift Hamiltonian is proposed, and the canonical gyroradius is presented. The proposed drift Hamiltonian agrees with Alfven's drift velocity to lowest non-vanishing order in the gyroradius. The relation between the exact, time dependent equations of motion and the guiding center equation is clarified by a Lagrangian analysis. The deduced Lagrangian represents the drift motion. (Kato, T.)

Annual technology assessment and progress report for the Buried Transuranic Waste Program at the Idaho National Engineering Laboratory

International Nuclear Information System (INIS)

Low, J.O.

1985-12-01

An improved-confinement technology as applied to the Idaho National Engineering Laboratory (INEL) shallow-land-buried transuranic (TRU) waste is being investigated. An improved-confinement technology, in situ grouting, is being demonstrated in a 2-year engineering feasibility test at the INEL. Grout formulation and development were completed by Oak Ridge National Laboratory in Tennessee to support the in situ grouting test. Three grout formulations have been adapted to the arid, unsaturated soil conditions at the INEL: ordinary particulate grout; microfine penetration grout; soil grout. Three test trenches were constructed north of the INEL's Subsurface Disposal Area (SDA). Nonradioactive waste forms closely resembling TRU waste buried at the INEL have been fabricated and are ready for emplacement into these test trenches. A literature search for a simulated (analog) TRU tracer was completed as well as a chemical characterization of the INEL soil. Data developed from the chemistry characterization and literature search have been inputed into the selection and laboratory testing of the TRU analog tracers. Simulated TRU tracers will be loaded into waste forms prior to emplacement into the test trenches. Test trench data acquisition instrumentation will be installed during waste form emplacement. Instrumentation will monitor for moisture movement and tracer detection. Plans for test completion in FY-1986 are also shown. Various buried waste improved-confinement technologies performed by other Department of Energy sites were assessed for applicability to the INEL buried TRU waste. Primary demonstrations were performed at the Hanford site in Washington and at ORNL. This report also includes information on accomplishments of related activities at the INEL such as the program for Environmental Surveillance of the Radioactive Waste Management complex as well as the Subsurface Migration Studies. 18 refs., 11 figs., 12 tabs

Strike-slip pull-apart process and emplacement of Xiangshan uranium-producing volcanic basin

International Nuclear Information System (INIS)

Qiu Aijin; Guo Lingzhi; Shu Liangshu

2001-01-01

Xiangshan volcanic basin is one of the famous uranium-producing volcanic basins in China. Emplacement mechanism of Xiangshan uranium-producing volcanic basin is discussed on the basis of the latest research achievements of deep geology in Xiangshan area and the theory of continental dynamics. The study shows that volcanic activity in Xiangshan volcanic basin may be divided into two cycles, and its emplacement is controlled by strike-ship pull-apart process originated from the deep regional faults. Volcanic apparatus in the first cycle was emplaced in EW-trending structure activated by clockwise strike-slipping of NE-trending deep fault, forming the EW-trending fissure-type volcanic effusion belt. Volcanic apparatus in the second cycle was emplaced at junction points of SN-trending pull-apart structure activated by sinistral strike-slipping of NE-trending deep faults and EW-trending basement faults causing the center-type volcanic magma effusion and extrusion. Moreover, the formation mechanism of large-rich uranium deposits is discussed as well

Mechanical tunnel excavation in welded tuff

International Nuclear Information System (INIS)

Sperry, P.E.

1991-01-01

The Technical Review Board for the US high-level radioactive waste facility at Yucca Mountain has recommended maximum use of open-quotes the most modern mechanical excavation techniques...in order to reduce disturbance to the rock walls and to achieve greater economy of time and cost.close quotes Tunnels for the waste repository at Yucca Mountain can be economically constructed with mechanical excavation equipment. This paper presents the results of mechanical excavation of a tunnel in welded tuff, similar to the tuffs of Yucca Mountain. These results are projected to excavation of emplacement drifts in Yucca Mountain using a current state-of-the-art tunnel boring machine (TBM)

ENGINEERED BARRIER SYSTEM: PHYSICAL AND CHEMICAL ENVIRONMENT

Energy Technology Data Exchange (ETDEWEB)

R. Jarek

2004-11-23

The purpose of this report is to describe the evolution of the physical and chemical environmental conditions within the waste emplacement drifts of the repository, including the drip shield and waste package surfaces. The abstraction model is used in the total system performance assessment for the license application (TSPA LA) to assess the performance of the engineered barrier system and the waste form. This report develops and documents a set of these abstraction-level models that describe the engineered barrier system physical and chemical environment. Where possible, these models use information directly from other reports as input, which promotes integration among process models used for TSPA-LA. Specific tasks and activities of modeling the physical and chemical environment are included in ''Technical Work Plan for: Near-Field Environment and Transport In-Drift Geochemistry Model Report Integration'' (BSC 2004 [DIRS 171156], Section 1.2.2). As described in the technical work plan, the development of this report is coordinated with the development of other engineered barrier system reports.

Thermal loading study for FY 1996. Volume 2

International Nuclear Information System (INIS)

1996-01-01

The primary objective of this study was to provide recommendations for Mined Geologic Disposal System requirements affected by thermal loading that will provide sufficient definition to facilitate development of design concepts and support life cycle cost determinations. The study reevaluated and/or redefined selected thermal goals used for design and are currently contained in the requirements documents or the Controlled Design Assumption Document. The study provided recommendations as to what, if any, actions (such as edge loading and limiting of the heat variability between waste packages) are needed and must be accommodated in the design. Additionally, the study provided recommendations as to what alternative thermal loads should be maintained for continued flexibility. This report contains seven appendices: Technical basis for evaluation of thermal goals below the potential nuclear was repository at Yucca Mountain; Thermal-mechanical evaluation of the 200 C drift-wall temperature goal; Evaluation of ground stability and support; Coupled ventilation and hydrothermal evaluations; Heat flow and temperature calculations for continuously ventilated emplacement drifts; Thermal management using aging and/or waste package selection; and Waste stream evaluations

ENGINEERED BARRIER SYSTEM: PHYSICAL AND CHEMICAL ENVIRONMENT

Energy Technology Data Exchange (ETDEWEB)

G.H. Nieder-Westermann

2005-04-07

The purpose of this report is to describe the evolution of the physical and chemical environmental conditions within the waste emplacement drifts of the repository, including the drip shield and waste package surfaces. The abstraction model is used in the total system performance assessment for the license application (TSPA LA) to assess the performance of the engineered barrier system and the waste form. This report develops and documents a set of these abstraction-level models that describe the engineered barrier system physical and chemical environment. Where possible, these models use information directly from other reports as input, which promotes integration among process models used for TSPA-LA. Specific tasks and activities of modeling the physical and chemical environment are included in ''Technical Work Plan for: Near-Field Environment and Transport In-Drift Geochemistry Model Report Integration'' (BSC 2004 [DIRS 171156], Section 1.2.2). As described in the technical work plan, the development of this report is coordinated with the development of other engineered barrier system reports.

Coupled In-Rock and In-Drift Hydrothermal Model Study For Yucca Mountain

International Nuclear Information System (INIS)

G. Danko; J. Birkholzer; D. Bahrami

2006-01-01

A thermal-hydrologic-natural-ventilation model is configured for simulating temperature, humidity, and condensate distributions in the coupled domains of the in-drift airspace and the near-field rockmass in the proposed Yucca Mountain repository. The multi-physics problem is solved with MULTIFLUX in which a lumped-parameter computational fluid dynamics model is iterated with TOUGH2. The solution includes natural convection, conduction, and radiation for heat as well as moisture convection and diffusion for moisture transport with half waste package scale details in the drift, and mountain-scale heat and moisture transport in the porous and fractured rock-mass. The method provides fast convergence on a personal computer computational platform. Numerical examples and comparison with a TOUGH2 based, integrated model are presented

Drift chamber

International Nuclear Information System (INIS)

Inagaki, Yosuke

1977-01-01

Drift chamber is becoming an important detector in high energy physics as a precision and fast position detector because of its high spatial resolution and count-rate. The basic principle is that it utilizes the drift at constant speed of electrons ionized along the tracks of charged particles towards the anode wire in the nearly uniform electric field. The method of measuring drift time includes the analog and digital ones. This report describes about the construction of and the application of electric field to the drift chamber, mathematical analysis on the electric field and equipotential curve, derivation of spatial resolution and the factor for its determination, and selection of gas to be used. The performance test of the chamber was carried out using a small test chamber, the collimated β source of Sr-90, and 500 MeV/C electron beam from the 1.3 GeV electron synchrotron in the Institute of Nuclear Study, University of Tokyo. Most chambers to date adopted one dimensional read-out, but it is very advantageous if the two dimensional read-out is feasible with one chamber when the resolution in that direction is low. The typical methods of delay line and charge division for two dimensional read-out are described. The development of digital read-out system is underway, which can process the signal of a large scale drift chamber at high speed. (Wakatsuki, Y.)

Grouting of nuclear waste vault shafts

International Nuclear Information System (INIS)

Gyenge, M.

1980-01-01

A nuclear waste vault must be designed and built to ensure adequate isolation of the nuclear wastes from human contact. Consequently, after a vault has been fully loaded it must be adequately sealed off to prevent radionuclide migration which may be provided by circulating ground water. Of particular concern in vault sealing are the physical and chemical properties of the sealing materials its long-term durability and stability and the techniques used for its emplacement. Present grouting technology and grout material are reviewed in terms of the particular needs of shaft grouting. Areas requiring research and development are indicated

Design of the human computer interface on the telerobotic small emplacement excavator

International Nuclear Information System (INIS)

Thompson, D.H.; Killough, S.M.; Burks, B.L.; Draper, J.V.

1995-01-01

The small emplacement excavator (SEE) is a ruggedized military vehicle with backhoe and front loader used by the U.S. Army for explosive ordinance disposal (EOD) and general utility excavation activities. This project resulted from a joint need in the U.S. Department of Energy (DOE) for a remote controlled excavator for buried waste operations and the U.S. Department of Defense for remote EOD operations. To evaluate the feasibility of removing personnel from the SEE vehicle during high-risk excavation tasks, a development and demonstration project was initiated. Development of a telerobotic SEE (TSEE) was performed by the Oak Ridge National Laboratory in a project funded jointly by the U.S. Army and the DOE. The TSEE features teleoperated driving, a telerobotic backhoe with four degrees of freedom, and a teleoperated front loader with two degrees of freedom on the bucket. Remote capabilities include driving (forward, reverse, brake, steering), power takeoff shifting to enable digging modes, deploying stabilizers, excavation, and computer system booting

Nuclear waste vault sealing

International Nuclear Information System (INIS)

Gyenge, M.

1980-01-01

A nuclear waste vault must be designed and built to ensure adequate isolation of the nuclear wastes from human contact. Consequently, after a vault has been fully loaded, it must be adequately sealed off to prevent radionuclide migration which may be provided by circulating groundwater. Vault sealing entails four major aspects, i.e.: (a) vault grouting; (b) borehole sealing; (c) buffer packing; and (d) backfilling. Of particular concern in vault sealing are the physical and chemical properties of the sealing material, its long-term durability and stability, and the techniques used for its emplacement. Present sealing technology and sealing materials are reviewed in terms of the particular needs of vault sealing. Areas requiring research and development are indicated

Project Guarantee 1985. Repository for high-level radioactive waste: construction and operation

International Nuclear Information System (INIS)

Anon.

1985-01-01

An engineering project study aimed at demonstrating the feasibility of constructing a deep repository for high-level waste (Type C repository) has been carried out; the study is based on a model data-set representing typical geological and rock mechanical conditions as found outside the so-called Permocarboniferous basin in the regions under investigation by Nagra in Cantons Aargau, Schaffhausen, Solothurn and Zuerich. The repository is intended for disposal of high-level waste and any intermediate-level waste from re-processing in which the concentration of long-lived alpha-emitters exceeds the permissible limits set for a Type B repository. Final disposal of high-level waste is in subterranean, horizontally mined tunnels and of intermediate-level waste in underground vertical silos. The repository is intended to accomodate a total of around 6'000 HWL-cylinders (gross volume of around 1'200 m3) and around 10'000 m3 of intermediate-level waste. The total excavated volume is around 1'100'000 m3 and a construction time for the whole repository (up to the beginning of emplacement) of around 15 years is expected. For the estimated 50-year emplacement operations, a working team of around 60 people will be needed and a team of around 160 for the simultaneous tunnelling operations and auxiliary work. The project described in the present report permits the conclusion that construction of a repository for high-level radioactive waste and, if necessary, spent fuel-rods is feasible with present-day technology

Gas and water flow in an excavation-induced fracture network around an underground drift: A case study for a radioactive waste repository in clay rock

Science.gov (United States)

de La Vaissière, Rémi; Armand, Gilles; Talandier, Jean

2015-02-01

The Excavation Damaged Zone (EDZ) surrounding a drift, and in particular its evolution, is being studied for the performance assessment of a radioactive waste underground repository. A specific experiment (called CDZ) was designed and implemented in the Meuse/Haute-Marne Underground Research Laboratory (URL) in France to investigate the EDZ. This experiment is dedicated to study the evolution of the EDZ hydrogeological properties (conductivity and specific storage) of the Callovo-Oxfordian claystone under mechanical compression and artificial hydration. Firstly, a loading cycle applied on a drift wall was performed to simulate the compression effect from bentonite swelling in a repository drift (bentonite is a clay material to be used to seal drifts and shafts for repository closure purpose). Gas tests (permeability tests with nitrogen and tracer tests with helium) were conducted during the first phase of the experiment. The results showed that the fracture network within the EDZ was initially interconnected and opened for gas flow (particularly along the drift) and then progressively closed with the increasing mechanical stress applied on the drift wall. Moreover, the evolution of the EDZ after unloading indicated a self-sealing process. Secondly, the remaining fracture network was resaturated to demonstrate the ability to self-seal of the COx claystone without mechanical loading by conducting from 11 to 15 repetitive hydraulic tests with monitoring of the hydraulic parameters. During this hydration process, the EDZ effective transmissivity dropped due to the swelling of the clay materials near the fracture network. The hydraulic conductivity evolution was relatively fast during the first few days. Low conductivities ranging at 10-10 m/s were observed after four months. Conversely, the specific storage showed an erratic evolution during the first phase of hydration (up to 60 days). Some uncertainty remains on this parameter due to volumetric strain during the

Consistent guiding center drift theories

International Nuclear Information System (INIS)

Wimmel, H.K.

1982-04-01

Various guiding-center drift theories are presented that are optimized in respect of consistency. They satisfy exact energy conservation theorems (in time-independent fields), Liouville's theorems, and appropriate power balance equations. A theoretical framework is given that allows direct and exact derivation of associated drift-kinetic equations from the respective guiding-center drift-orbit theories. These drift-kinetic equations are listed. Northrop's non-optimized theory is discussed for reference, and internal consistency relations of G.C. drift theories are presented. (orig.)

No-migration variance petition for the Waste Isolation Pilot Plant

International Nuclear Information System (INIS)

Duff, M.; Carnes, R.; Hart, J.; Hansen, R.

1991-01-01

The US Department of Energy (DOE) is petitioning the US Environmental Protection Agency (EPA) to allow the emplacement of hazardous wastes subject to the Resource Conservation and Recovery Act (RCRA) land disposal restrictions in the Waste Isolation Pilot Plant (WIPP). The basis of the petition is that there will be no migration of hazardous constituents from the repository for as long as the wastes remain hazardous. The EPA regulations in 40 CFR Section 268.6 identify specific criteria that must be addressed in making a demonstration of no migration. EPA's approval of this petition will allow the WIPP facility to accept wastes otherwise prohibited or restricted from land disposal. 5 refs

Peak Ground Velocities for Seismic Events at Yucca Mountain, Nevada

International Nuclear Information System (INIS)

K. Coppersmith; R. Quittmeyer

2005-01-01

horizontal PGV hazard curve for the waste emplacement level. The relation of this analysis to other work feeding the seismic consequence abstraction and the TSPA is shown on Figure 1-1. The ground motion hazard results from the PSHA provide the basis for inputs to a site-response model that determines the effect of site materials on the ground motion at a location of interest (e.g., the waste emplacement level). Peak ground velocity values determined from the site-response model for the waste emplacement level are then used to develop time histories (seismograms) that form input to a model of drift degradation under seismic loads potentially producing rockfall. The time histories are also used to carry out dynamic seismic structural response calculations of the drip shield and waste package system. For the drip shield, damage from seismically induced rockfall also is considered. In the seismic consequence abstraction, residual stress results from the structural response calculations are interpreted in terms of the percentage of the component (drip shield, waste package) damaged as a function of horizontal PGV. The composite hazard curve developed in this analysis, which reflects the results of site-response modeling and the bound to credible horizontal PGV at the waste emplacement level, also feeds the seismic consequence abstraction. The composite hazard curve is incorporated into the TSPA sampling process to bound horizontal PGV and related seismic consequences to values that are credible

Buried waste containment system materials. Final Report

International Nuclear Information System (INIS)

Weidner, J.R.; Shaw, P.G.

1997-10-01

This report describes the results of a test program to validate the application of a latex-modified cement formulation for use with the Buried Waste Containment System (BWCS) process during a proof of principle (POP) demonstration. The test program included three objectives. One objective was to validate the barrier material mix formulation to be used with the BWCS equipment. A basic mix formula for initial trials was supplied by the cement and latex vendors. The suitability of the material for BWCS application was verified by laboratory testing at the Idaho National Engineering and Environmental Laboratory (INEEL). A second objective was to determine if the POP BWCS material emplacement process adversely affected the barrier material properties. This objective was met by measuring and comparing properties of material prepared in the INEEL Materials Testing Laboratory (MTL) with identical properties of material produced by the BWCS field tests. These measurements included hydraulic conductivity to determine if the material met the US Environmental Protection Agency (EPA) requirements for barriers used for hazardous waste sites, petrographic analysis to allow an assessment of barrier material separation and segregation during emplacement, and a set of mechanical property tests typical of concrete characterization. The third objective was to measure the hydraulic properties of barrier material containing a stop-start joint to determine if such a feature would meet the EPA requirements for hazardous waste site barriers

Deep underground disposal of radioactive wastes: Near field effects

International Nuclear Information System (INIS)

1985-01-01

This report reviews the important near-field effects of the disposal of wastes in deep rock formations. The basic characteristics of waste form, container and package, buffer and backfill materials and potential host-rock types are discussed from the perspective of the performance requirements of the total repository system. Effects of waste emplacement on the separate system components and on the system as a whole are discussed. The effects include interactions between groundwater and brines and the other system components, thermal and thermo-mechanical effects, and chemical and geochemical reactions. Special consideration is given to the radiation field that exists in proximity to the waste containers and also to the coupled effects of different phenomena

Safety principles and technical criteria for the underground disposal of high level radioactive wastes

International Nuclear Information System (INIS)

1989-01-01

The main objective of this book is to set out an internationally agreed set of principles and criteria for the design of deep underground repositories for the disposal of high level radioactive wastes. This book is concerned with the post-closure period. Consideration of the operational requirements which must be met when wastes are being handled, stored and emplaced are not therefore included

Waste Package Project quarterly report, July 1, 1995--September 30, 1995

International Nuclear Information System (INIS)