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At least 775 records · Page 43
All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

Technical Basis and Implementation of Groundwater Pathway Models Used for INL's RH LLW Disposal Facility Performance Assessment - 19440

The Idaho National Laboratory's (INL's) Remote-Handled Low-Level Waste (RH LLW) Disposal Facility has been designed and constructed to receive waste generated by INL in support of the Office of Nuclear Energy (NE) and Naval Reactors (NR) missions. RH LLW is generated on the INL site from spent nuclear fuel handling and operations at the Naval Reactors Facility (NRF), the Advanced Test Reactor (ATR), and at the Material and Fuels Complex (MFC) in addition to other miscellaneous research activities. The facility has been designed to receive RH LLW currently in storage at INL facilities and new waste generated by nuclear research programs. RH LLW received at the facility is solid waste consisting of surface contaminated materials, ion exchange resins used to purify cooling water, and activated metals generated during reactor core change outs, Naval spent fuel management, and nuclear R and D. The facility has been designed to meet the requirements of DOE Order 435.1, Radioactive Waste Management for all-pathways including groundwater and air, radon emissions limits, and inadvertent intrusion limits. The facility design includes a vault system incorporating robust concrete vaults oriented vertically in close-packed arrays, a hydraulic drainage system designed to limit water contact with the stainless steel waste canisters and concrete vaults, and a robust monitoring system. Per the environmental assessment (EA), all waste will be disposed of in stainless steel waste canisters. During facility design and construction, site-specific data was collected to provide laboratory and field scale performance characteristics for the concrete vault and hydraulic drainage systems. Field-scale infiltration tests were conducted and monitoring data was interpreted using numerical simulation. Based on interpretation of the observed laboratory and field test data, the facility performance assessment (PA) was developed. The PA groundwater pathway source release model differentially credits release from surface contaminated waste, ion exchange resins, and activated metals. It also credits steel waste canisters with failure rates based on actual corrosion data. Once released, radionuclide transport through the vadose zone and aquifer to downgradient receptor locations was simulated in the PA using a one-dimensional modeling approach based on three-dimensional modeling of historically-contaminated INL sites. This paper provides an overview of the groundwater pathway source release models and their technical implementation in the PA. It also describes the one-dimensional transport model used for the vadose zone and multi-dimensional model used for the aquifer. (authors)

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DOE OSTI2019

Radioactive Waste Minimization - Clearance for Reuse, Recycling and Disposal - 19598

Virtually all activities related to radioactive materials and to the utilization of nuclear energy generate radioactive waste. Due to limited storage facilities and waning public acceptance for long-term disposal of radioactive waste as well as the associated cost, implementation of strategies and methods for its effective reduction are mandatory. Clearance of materials is one possible concept to reach this goal. The International Atomic Energy Agency (IAEA) defines the process of clearance as 'the removal of radioactive materials or radioactive objects within authorized practices from any further regulatory control by the regulatory body.' Thus, these materials are no further treated as radioactive waste and may be recycled or disposed of conventionally. In quantitative terms, clearance is generally related to the '10 μSv-Concept' This means that the effective dose expected to be incurred by any member of the public due to cleared materials is of the order of 10 μSv or less in a year. A combination of appropriate clearance concepts, methods and measurement techniques, following an integrated clearance strategy for a project, allows the return of up to 97 % of the mass of a decommissioned NPP to the conventional waste cycle. This includes the concept of clearance for a specific purpose, often referred to as 'conditional clearance', i.e. the case where a certain condition has to be fulfilled before the act of clearance has been completed. As a technical expert organization, TUV NORD EnSys accompanied the development, implementation and execution of clearance procedures in more than 13 nuclear power-plants and installations, either operational or in various states of decommissioning. Thus, TUV Nord EnSys has abundant experience in and a comprehensive overview on the implementation of different clearance strategies and procedures. Considering the increased waste-volumes and additional waste-streams to be handled during decommissioning, the transition to a decommission-focused clearance system is of vital importance for successful waste-minimization. This paper provides a view on the perspective of a technical support organisation (TSO) towards implementation of a holistic clearance strategy into the decommissioning concept for radioactive waste minimization. (authors)

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DOE OSTI2020

An Improved Model For Determining Salinity Recharge Time for Deep Borehole Disposal - 20405

In a previous publication (K. P. Travis, D. Burley and F. G. F. Gibb, WM2017 Conference, Phoenix Arizona paper no. 17480) we introduced a numerical model to estimate the time it would take for a pressure perturbation to subside following its creation from the construction of a deep geological borehole in water-saturated rock. The model was based on the notion of a point source of momentum - a solution of the time-dependent pressure-diffusion equation. Using the model, we estimated that it would take on the order of 10 k years for physical equilibrium to become re-established following the sinking of a 5 km borehole in granite. The significance of such a model is that it places an upper bound on the lifetime required of a borehole sealing system - a necessary input to a borehole post-closure safety case assessment. No engineered sealing system has ever been devised which is capable of retaining its sealing properties for 10 half-lives of long-lived radioisotopes in spent fuel or high-level waste. One of the key advantages of disposing of nuclear waste via Deep Borehole Disposal (DBD) is the natural sealing provided by density stratified groundwater. Upon sinking of a borehole, and subsequent filling with fresh water, brine or drilling mud, this natural barrier may be temporarily damaged. Over time, fresh brine from the far-field will flow towards or away from the hole (driven by a pressure gradient) and will eventually re-establish the original salinity gradient. It follows that the engineered seals need last only as long as the time required for this salinity gradient to reset itself. One of the limitations of our previous model was the use of a static boundary condition on the borehole wall. The model used a boundary pressure which varied quadratically with depth (arising from differences between the pressure of a column of fresh water in the borehole and that of a column of brine in the host rock). However as brine replaces fresh water in the borehole (driven by a pressure gradient), the boundary function must change with time. We now introduce an improved model which takes this time dependent boundary condition into account. The model also takes into account the time taken for a mixture of brine and fresh water to re-establish chemical equilibrium through the process of diffusion. The paper contains the mathematical details of our new iterative model as well as results showing the time taken to reach steady state, and concentration profiles for the components of the brine-filled borehole as a function of time together with a discussion on the implications of the results for developing a post-closure safety assessment for DBD. (authors)

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DOE OSTI2020

Hydrologic Modeling of Subsurface Brine Density Stratification for Deep Borehole Disposal - 20213

Density-dependent and topographically-driven subsurface brine flow was modeled in regional-scale and Deep Borehole Disposal-scale domains under varying conditions. These conditions included variable basement permeability, brine density, and hydraulic head. Findings revealed the possibility of hydraulic isolation of deep basement for at least 1 million years under appropriate conditions, which supports the viability of the Deep Borehole Disposal concept. (authors)

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DOE OSTI2020

Criticality Consideration for Geologic Disposal of DOE-Owned Spent Nuclear Fuels - 20255

Decisions must be made relative to packaging spent nuclear fuel (SNF) for storage, transportation, and eventual disposal. This paper evaluates potential criticality-related loading and packaging considerations for Department of Energy (DOE)-owned SNF that could be impacted favorably or unfavorably by disposal uncertainties and proposes approaches to minimize such impacts. Using Al-Based SNF as an example and drawing on analyses performed in support of the Yucca Mountain Repository License Application (LA), this paper explores the role of natural barriers, engineered barriers, and waste form characteristics relative to reducing the potential for postclosure criticality. Several parameters important to postclosure criticality are considered in order to determine which are useful discriminators to support packaging decisions. Using these parameters, a qualitative assessment is performed relative to three options for managing DOE-owned SNFs. (authors)

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DOE OSTI2020

Study on the Strategy for Final Disposal and Reuse of Decontamination Wastes after the Fukushima Daiichi NPP Accident - 20426

After the Fukushima Daiichi NPP accident, a large area of East Japan was contaminated by radio-Cs. To discuss the final disposal of the contaminated wastes off-site and the required development of related technologies, it is important to understand the volume and radiation concentration of the wastes after various treatments. For a scientific investigation purpose, a research committee has been formed by the Society for the Remediation of Radioactive Contamination in the Environment, which is independent from the Ministry of Environment. This committee's work is focused on discussing various scenarios for the final disposal considering a mass balance evaluation of the entire process, economic evaluations, and stakeholder consultation for final consensus. The current discussions of this committee are herein reported. (authors)

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DOE OSTI2020

Generic Post-Closure Safety Assessment for Deep Borehole Disposal of Heat-Generating Waste - 20451

The deep borehole disposal (DBD) concept for high-activity waste consists of drilling a large-diameter borehole into crystalline basement rock to a depth of about 5,000 m, placing waste packages in the lower portion of the borehole - the waste emplacement zone - and sealing and plugging the upper portion of the borehole with a combination of bentonite, cement plugs, and sand or crushed rock ballast. The waste emplacement zone is several times deeper than typical mined repositories and is well below the typical maximum depth of fresh groundwater resources. Recent work describes a generic reference case for deep borehole disposal of Cs and Sr capsules stored at the Hanford Site. Two scenarios are described: an undisturbed (nominal) scenario, and disturbed scenario, in which a waste package is stuck adjacent to a transmissive fracture zone in crystalline basement. Simulations of the processes affecting radionuclide release and transport in the two scenarios provide the basis for quantitative post-closure performance assessment (PA) of the reference case. This paper presents a probabilistic PA of the stuck waste package scenario, i.e. uncertainties affecting radionuclide mobilization and transport are propagated through 200 realizations of a predictive model to quantify uncertainties in predicted radionuclide concentrations within the borehole and the fracture zone. Simulations predict no radionuclide transport in the borehole to distances greater than 100 m above the stuck waste package. An assumed lateral pressure gradient results in predictions of limited radionuclide transport in the fracture zone at a lateral distance of 200 m from the stuck waste package. These results complement previous results for the nominal scenario and contribute to the development of a generic safety case for DBD. (authors)

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DOE OSTI2024

Preliminary Analysis of Advanced Reactor Spent Nuclear Fuel Storage, Transportation, and Disposal

Due to increased interest in advanced reactor deployment and their associated potential new fuel cycles, the U.S. Department of Energy (DOE) Spent Fuel and Waste Science and Technology (SFWST) program has begun to evaluate the possible implications of long term management and final disposition of the spent nuclear fuel (SNF) generated. Safely managing and dispositioning this SNF, along with any other associated radioactive waste, is the primary focus of this initial preliminary assessment. This paper summarizes efforts to evaluate the characteristics and packaging options for three types of advanced reactor SNF forms: (1) tristructural isotropic (TRISO), (2) metallic, and (3) irradiated fuel salt presented in the report titled “Storage, Transportation, and Disposal of Advanced Reactor Spent Nuclear Fuel and High-Level Waste”. TRISO and metallic SNF and their associated waste streams were emphasized because of the near-term anticipated demonstrations of X-energy’s Xe-100 and TerraPower and GE Hitachi’s Natrium advanced reactors. Preliminary information on spent fuel salts discharged from molten-salt reactors (MSRs) was also examined to provide a baseline for future efforts. All calculations and assumptions were based on publicly available information. This paper identifies several different reactors that produce either TRISO or metallic SNF as well as a few of the reactor and fuel characteristics used for the assessments. Based on these characteristics, calculations were performed to determine the applicability of packaging SNF into existing canister designs. The evaluations included geometric (e.g., dimension, volume) and mass/weight considerations, known operational approaches and loading procedures, physical and chemical considerations/conditions for storage environments, and as-loaded radiation, thermal, and criticality analyses to identify constraints on storage, transportation, and disposal. Gaps in publicly available data pertaining to reactor operation and/or fuel composition provide increased uncertainty in some evaluations. Additionally, uncertainty in packaging and SNF management operations provide additional uncertainty. However, preliminary conclusions can still be assessed through this work and are presented in this paper.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Waste Form Release Calculations for the 2005 Integrated Disposal Facility Performance Assessment. Erratum

This report refers to or contains K g values for glasses LAWA44, LAWB45 and LAWC22 affected by calculations errors as identified by Papathanassiu et al. (2011). The corrected K g values are reported in an erratum included in the revised version of the original report. The revised report can be referenced as follows: Pierce E. M. et al. (2004) Waste Form Release Data Package for the 2005 Integrated Disposal Facility Performance Assessment. PNNL-14805 Rev. 0 Erratum. Pacific Northwest National Laboratory, Richland, WA, USA.

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DOE OSTI2013

A Proposal for Geologic Radioactive Waste Disposal Environmental Zero-State and Subsequent Monitoring Definition - First Lessons Learned from the French Environment Observatory - 13188

Obtaining a reference state of the environment before the beginning of construction work for a geological repository is essential as it will be useful for further monitoring during operations and beyond, thus keeping a memory of the original environmental state. The area and the compartments of the biosphere to be observed and monitored as well as the choice of the markers (e.g. bio-markers, biodiversity, quality of the environment, etc.) to be followed must be carefully selected. In parallel, the choice and selection of the environmental monitoring systems (i.e. scientific and technical criteria, social requirements) will be of paramount importance for the evaluation of the perturbations that could be induced during the operational phase of the repository exploitation. This paper presents learning points of the French environment observatory located in the Meuse/Haute-Marne that has been selected for studying the feasibility of the underground disposal of high level wastes in France. (authors)

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DOE OSTI2013

Safety Assessment for a Surface Repository in the Chernobyl Exclusion Zone - Methodology for Assessing Disposal under Intervention Conditions - 13476

The Radioactive Waste Disposal Facility (RWDF) Buryakovka was constructed in 1986 as part of the intervention measures after the accident at Chernobyl NPP (ChNPP). Today, RWDF Buryakovka is still being operated but its maximum capacity is nearly reached. Plans for enlargement of the facility exist since more than 10 years but have not been implemented yet. In the framework of an European Commission Project DBE Technology GmbH prepared a safety analysis report of the facility in its current state (SAR) and a preliminary safety analysis report (PSAR) based on the planned enlargement. Due to its history RWDF Buryakovka does not fully comply with today's best international practices and the latest Ukrainian regulations in this area. The most critical aspects are its inventory of long-lived radionuclides, and the non-existent multi-barrier waste confinement system. A significant part of the project was dedicated, therefore, to the development of a methodology for the safety assessment taking into consideration the facility's special situation and to reach an agreement with all stakeholders involved in the later review and approval procedure of the safety analysis reports. Main aspect of the agreed methodology was to analyze the safety, not strictly based on regulatory requirements but on the assessment of the actual situation of the facility including its location within the Exclusion Zone. For both safety analysis reports, SAR and PSAR, the assessment of the long-term safety led to results that were either within regulatory limits or within the limits allowing for a specific situational evaluation by the regulator. (authors)

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DOE OSTI2013

Safety Assessment for the Kozloduy National Disposal Facility in Bulgaria - 13507

Due to the early decommissioning of four Water-Water Energy Reactors (WWER) 440-V230 reactors at the Nuclear Power Plant (NPP) near the city of Kozloduy in Bulgaria, large amounts of low and intermediate radioactive waste will arise much earlier than initially scheduled. In or-der to manage the radioactive waste from the early decommissioning, Bulgaria has intensified its efforts to provide a near surface disposal facility at Radiana with the required capacity. To this end, a project was launched and assigned in international competition to a German-Spanish consortium to provide the complete technical planning including the preparation of the Intermediate Safety Assessment Report. Preliminary results of operational and long-term safety show compliance with the Bulgarian regulatory requirements. The long-term calculations carried out for the Radiana site are also a good example of how analysis of safety assessment results can be used for iterative improvements of the assessment by pointing out uncertainties and areas of future investigations to reduce such uncertainties in regard to the potential radiological impact. The computer model used to estimate the long-term evolution of the future repository at Radiana predicted a maximum total annual dose for members of the critical group, which is carried to approximately 80 % by C-14 for a specific ingestion pathway. Based on this result and the outcome of the sensitivity analysis, existing uncertainties were evaluated and areas for reasonable future investigations to reduce these uncertainties were identified. (authors)

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DOE OSTI2018

Numerical Modelling of Radionuclide Migration in the Context of a Near-Surface LILW Disposal Facility - 18618

The reference solution for the management of low and intermediate level short-lived waste (LILW) in France consists in their emplacement in concrete vaults built at the near-surface. The corresponding disposal facility under operation by the French waste management organization (Andra), is located in North-eastern France and planned to be closed in 2060. The safety case developed by Andra has to show the evidence that, amongst other aspects, the radionuclide transfer through the confinement barriers will occur in such a way that the human and environmental exposure will be as low as reasonably achievable (ALARA) and in accordance with the dose constraints. For this purpose, modelling is used to integrate uncertainties about the characteristics of concrete components, waste and water table and their evolution in time and space. At last, modelling is finally used to investigate the radiological impact of specific scenarios in which the failure of safety barriers occurs. In the frame of its technical review of the safety assessment, the French Institute for Radiological Protection and Nuclear Safety (IRSN) also performs such kind of numerical modelling of radionuclide fate from the waste to the hydrogeological outlets, in order to check whether the assessment performed by Andra takes into account sufficiently conservative assumptions. The present work aims to evaluate the role of components of the vaults on radionuclide transport during the post-closure period over a few thousand years. For this purpose, IRSN carries out numerical calculations with MELODIE software, which simulates water flow and transport of species, in order to simulate radionuclide migration from waste packages emplaced in a concrete vault to the surrounding groundwater. The developed transport model focuses on 3-H radionuclide and integrates Andra's main assumptions (release model from the waste packages, chemical and hydraulic properties of the vault components, shrinkage and cracks in concrete components). An adaptive mesh method, based on an a posteriori error estimator, was used to model the cracks and the consideration of a strong heterogeneity in the hydraulic and transport parameters values in the different components of the model. This method shows an improvement of results of about 40 times, whereas the mesh size was increased by only 2%. (authors)

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DOE OSTI2020

Waste Shielded Casks Transfer by Funicular from the Surface to Disposal Level in Cigeo - 20007

Andra is completing the detailed engineering phase of Cigeo - the French HLW and ILW-LL DGR. A funicular system is dedicated to the transfer of the waste disposal containers from the surface to the underground infrastructures down the nuclear access ramp. This paper explains the rationale having led to the choice of a funicular and presents the advantages of this system. Main technical components and mechanical performance of the funicular system are then detailed. Lastly, the ongoing two stages qualification testing program managed by Andra (in order to reach a level 6 on the international Technology Readiness Level scale) is discussed. (authors)

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DOE OSTI2012

A criticism of applications with multi-criteria decision analysis that are used for the site selection for the disposal of municipal solid wastes

Highlights: Black-Right-Pointing-Pointer The existing structure of the multi-criteria decision analysis for site selection is criticized. Black-Right-Pointing-Pointer Fundamental problematic points based on the critics are defined. Black-Right-Pointing-Pointer Some modifications are suggested in order to provide solutions to these problematical points. Black-Right-Pointing-Pointer A new structure for the decision making mechanism is proposed. Black-Right-Pointing-Pointer The feasibility of the new method is subjected to an evaluation process. - Abstract: The main aim of this study is to criticize the process of selecting the most appropriate site for the disposal of municipal solid wastes which is one of the problematic issues of waste management operations. These kinds of problems are pathological symptoms of existing problematical human-nature relationship which is related to the syndrome called ecological crisis. In this regard, solving the site selection problem, which is just a small part of a larger entity, for the good of ecological rationality and social justice is only possible by founding a new and extensive type of human-nature relationship. In this study, as a problematic point regarding the discussions on ecological problems, the existing structure of the applications using multi-criteria decision analysis in the process of site selection with three main criteria is criticized. Based on this critique, fundamental problematic points (to which applications are insufficient to find solutions) will be defined. Later, some modifications will be suggested in order to provide solutions to these problematical points. Finally, the criticism addressed to the structure of the method with three main criteria and the feasibility of the new method with four main criteria is subjected to an evaluation process. As a result, it is emphasized that the new structure with four main criteria may be effective in solution of the fundamental problematic points.

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DOE OSTI2013

A material flow analysis on current electrical and electronic waste disposal from Hong Kong households

Highlights: ► Most household TWARC waste is sold directly to private e-waste collectors in HK. ► The current e-waste recycling network is popular with HK households. ► About 80% of household generated TWARC is exported overseas each year. ► Over 7000 tonnes/yr of household generated TWARC reach landfills. ► It is necessary to upgrade safety and awareness in HK’s e-waste recycling industry. - Abstract: A material flow study on five types of household electrical and electronic equipment, namely television, washing machine, air conditioner, refrigerator and personal computer (TWARC) was conducted to assist the Government of Hong Kong to establish an e-waste take-back system. This study is the first systematic attempt on identifying key TWARC waste disposal outlets and trade practices of key parties involved in Hong Kong. Results from two questionnaire surveys, on local households and private e-waste traders, were used to establish the material flow of household TWARC waste. The study revealed that the majority of obsolete TWARC were sold by households to private e-waste collectors and that the current e-waste collection network is efficient and popular with local households. However, about 65,000 tonnes/yr or 80% of household generated TWARC waste are being exported overseas by private e-waste traders, with some believed to be imported into developing countries where crude recycling methods are practiced. Should Hong Kong establish a formal recycling network with tight regulatory control on imports and exports, the potential risks of current e-waste recycling practices on e-waste recycling workers, local residents and the environment can be greatly reduced.

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