Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “DISPOSAL”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 595 records · Page 33
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Methods of supporting a graphene sheet disposed on a frame support

This disclosure provides systems, methods, and apparatus related to graphene. In one aspect, a method includes submerging a frame support in an etching solution that is contained in a container. A growth substrate, a graphene sheet disposed on the growth substrate, and a primary support disposed on the graphene sheet is placed on a surface of the etching solution. The growth substrate is dissolved in the etching solution to leave the graphene sheet and the primary support floating on a surface of the etching solution. The etching solution in the container is replaced with a washing solution. The washing solution is removed from the container so that the graphene sheet becomes disposed on the frame support.

Han, Bong-Gyoon↗
DOE OSTI2012

A Preliminary Performance Assessment for Salt Disposal of High-Level Nuclear Waste - 12173

A salt repository is one of the four geologic media currently under study by the U.S. DOE Office of Nuclear Energy to support the development of a long-term strategy for geologic disposal of commercial used nuclear fuel (UNF) and high-level radioactive waste (HLW). The immediate goal of the generic salt repository study is to develop the necessary modeling tools to evaluate and improve the understanding of the repository system response and processes relevant to long-term disposal of UNF and HLW in a salt formation. The current phase of this study considers representative geologic settings and features adopted from previous studies for salt repository sites. For the reference scenario, the brine flow rates in the repository and underlying interbeds are very low, and transport of radionuclides in the transport pathways is dominated by diffusion and greatly retarded by sorption on the interbed filling materials. I-129 is the dominant annual dose contributor at the hypothetical accessible environment, but the calculated mean annual dose is negligibly small. For the human intrusion (or disturbed) scenario, the mean mass release rate and mean annual dose histories are very different from those for the reference scenario. Actinides including Pu-239, Pu-242 and Np-237 are major annual dose contributors, and the calculated peak mean annual dose is acceptably low. A performance assessment model for a generic salt repository has been developed incorporating, where applicable, representative geologic settings and features adopted from literature data for salt repository sites. The conceptual model and scenario for radionuclide release and transport from a salt repository were developed utilizing literature data. The salt GDS model was developed in a probabilistic analysis framework. The preliminary performance analysis for demonstration of model capability is for an isothermal condition at the ambient temperature for the near field. The capability demonstration emphasizes key attributes of a salt repository that are potentially important to the long-term safe disposal of UNF and HLW. The analysis presents and discusses the results showing repository responses to different radionuclide release scenarios (undisturbed and human intrusion). For the reference (or nominal or undisturbed) scenario, the brine flow rates in the repository and underlying interbeds are very low, and transport of radionuclides in the transport pathways is dominated by diffusion and greatly retarded by sorption on the interbed filling materials. I-129 (non-sorbing and unlimited solubility with a very long half-life) is the dominant annual dose contributor at the hypothetical accessible environment, but the calculated mean annual dose is negligibly small that there is no meaningful consequence for the repository performance. For the human intrusion (or disturbed) scenario analysis, the mean mass release rate and mean annual dose histories are very different from those for the reference scenario analysis. Compared to the reference scenario, the relative annual dose contributions by soluble, non-sorbing fission products, particularly I-129, are much lower than by actinides including Pu-239, Pu-242 and Np-237. The lower relative mean annual dose contributions by the fission product radionuclides are due to their lower total inventory available for release (i.e., up to five affected waste packages), and the higher mean annual doses by the actinides are the outcome of the direct release of the radionuclides into the overlying aquifer having high water flow rates, thereby resulting in an early arrival of higher concentrations of the radionuclides at the biosphere drinking water well prior to their significant decay. The salt GDS model analysis has also identified the following future recommendations and/or knowledge gaps to improve and enhance the confidence of the future repository performance analysis. - Repository thermal loading by UNF and HLW, and the effect on the engineered barrier and near-field performance. - Closure and consolidation of salt rocks by creep deformation under the influence of thermal perturbation, and the effect on the engineered barrier and near-field performance. - Brine migration and radionuclide transport under the influence of thermal perturbation in generic salt repository environment, and the effect on the engineered barrier and near-field performance and far-field performance. - Near-field geochemistry and radionuclide mobility in generic salt repository environment (high ionic strength brines, elevated temperatures and chemically reducing condition). - Degradation of engineer barrier components (waste package, waste canister, waste forms, etc.) in a generic salt repository environment (high ionic strength brines, elevated temperatures and chemically reducing condition). - Waste stream types and inventory estimates, particularly for reprocessing high-level waste. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

An Updated Performance Assessment For A New Low-Level Radioactive Waste Disposal Facility In West Texas - 12192

This Performance Assessment (PA) submittal is an update to the original PA that was developed to support the licensing of the Waste Control Specialists LLC Low-Level Radioactive Waste (LLRW) disposal facility. This update includes both the Compact Waste Facility (CWF) and the Federal Waste Facility (FWF), in accordance with Radioactive Material License (RML) No. R04100, License Condition (LC) 87. While many of the baseline assumptions supporting the initial license application PA were incorporated in this update, a new transport code, GoldSim, and new deterministic groundwater flow codes, including HYDRUS and MODFLOWSURFACT{sup TM}, were employed to demonstrate compliance with the performance objectives codified in the regulations and RML No. R04100, LC 87. A revised source term, provided by the Texas Commission on Environmental Quality staff, was used to match the initial 15 year license term. This updated PA clearly confirms and demonstrates the robustness of the characteristics of the site's geology and the advanced engineering design of the disposal units. Based on the simulations from fate and transport models, the radiation doses to members of the general public and site workers predicted in the initial and updated PA were a small fraction of the criterion doses of 0.25 mSv and 50 mSv, respectively. In a comparison between the results of the updated PA against the one developed in support of the initial license, both clearly demonstrated the robustness of the characteristics of the site's geology and engineering design of the disposal units. Based on the simulations from fate and transport models, the radiation doses to members of the general public predicted in the initial and updated PA were a fraction of the allowable 25 mrem/yr (0.25 m sievert/yr) dose standard for tens-of-thousands of years into the future. Draft Texas guidance on performance assessment (TCEQ, 2004) recommends a period of analysis equal to 1,000 years or until peak doses from the more mobile radionuclides occur. The EPA National Emissions Standards for Hazardous Air Pollutants limits radionuclide doses through the air pathway to 10 mrem/yr. Gaseous radionuclide doses from the CWF and the FWF, due to decomposition gases, are a small fraction of the dose limit. The radon flux from the CWF and FWF were compared to the flux limit of 20 pCi/m{sup 2}-s from 40 CFR 192. Because of the thick cover system, the calculated radon flux was a very small fraction of the limit. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Current Efforts in Immobilized Low-Activity Waste Glass Testing to Support the Hanford Site Integrated Disposal Facility - 17119

Cleanup plans for Hanford site high-level radioactive mixed tank waste calls for separating the waste into high-level waste (HLW) and low-activity waste (LAW) fractions. The fractions will be vitrified separately at the Hanford Tank Waste Treatment and Immobilization Plant (WTP). The HLW glass will be sent to an offsite geological disposal facility while the immobilized low-activity waste (ILAW) glass will be disposed at the onsite Integrated Disposal Facility (IDF). In order to inform models used to predict the performance of ILAW glass at IDF, it is necessary to understand the corrosion behavior of the glass. In this paper, we discuss tests that account for the effects of 1) solution chemistry, 2) the ion exchange reaction, and 3) secondary phase formation on the glass dissolution rate. To quantify the effects of solution chemistry, specifically changes in pH, T, and Si concentration, the single-pass flow-through (SPFT) test method is applied. A designed set of these tests results in a set of kinetic rate law parameters that are used in a glass dissolution rate law based on transition-state theory. Furthermore, for SPFT tests with relatively high concentrations of Si, an ion exchange rate is provided. Additionally, through tests such as the pressurized unsaturated flow (PUF) test method, a suite of secondary phases can be compiled. The collection of kinetic rate law parameters, the ion exchange term, and the suite of secondary phases are provided to IDF performance assessment (PA) modelers. Lastly, we present a description of the efforts made to verify and validate the use of these parameters and secondary phases through the use of a field lysimeter present at the Hanford site. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Salt Rock - the 60-Year-Old Prodigy Host Rock for Consent-Based Disposal of Long-Lived Radioactive Waste - 16010

The USA's HLW-disposal program has been on hold since 2010 pending enabling legislation for its only candidate HLW-repository since 1987 at the Yucca Mountain site in Nevada and/or other, 'consent-based', HLW-disposition solutions. Related projections in 2008 and 2012 suggested they might open 9-12 years and 35 years after being adequately enabled, respectively. In 2014, domestic HLW arisings exceeded the capacity of the candidate repository by >4,000 metric tons (MT) and they were projected to grow by 2,000-2,300 MT/year. One or more new HLW repositories are thus highly-likely in the future. In the meantime, the DOE should make every effort to promptly take title to civilian-generated HLW by other means. Historical evidence in the USA and abroad during the past 30+ years shows that all HLW-repository siting and development schedules and costs are governed by the inherent, intricate, relationships and related domino effects of the: 1. Maturity/robustness of the disposal concept; 2. Related levels of relevant domestic repository-sciences/engineering expertise, data, institutional knowledge, and applicable regulations; 3. Abundance and 'ease' of physical access to potentially-suitable host-rock sites; 4. 'Trust' in the implementing and regulating organizations; 5. Timely resolution of inevitable contentions and lawsuits; 6. Facility-host-acceptance level; 7. Timely and adequate enabling legislation, and therefore, ultimately, by 8. Prevailing Political Will. Based upon their respective status at the end of 2015, salt rock still offers the most, but not the only, promising path forward for new, mined, HLW repositories in the USA, and ∼5-km-deep boreholes offer the most-promising disposal-solution for small HLW containers/packages/pellets. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Directional Drilling Technology for HLW Disposal - Outline of System and its Application - 16078

During high-level waste (HLW) disposal site selection, borehole measurement and logging to investigate the hydro-geological and geo-mechanical conditions of the host rocks is a very important way to examine the potential of the disposal candidates. In Japan, attention is being given to Neogene soft sedimentary rock as the host rock for HLW disposal. In particular, the soft sedimentary rock in coastal areas is thought to be one of the best candidates, because there is little driving force from underground water. Directional drilling is supposed to be efficient under limited topographic and geological conditions, and the Central Research Institute of Electric Power Industry (CRIEPI) has been conducting a project on directional drilling and logging/measurement technologies since 2000. Basic directional drilling system was developed and the system was applied to the Neogene normal sedimentary rock at the Hokushin area of the Horonobe town in Hokkaido. The borehole was drilled to the 700 m long and the applicability of the system was confirmed until 2004. A basic directional drilling system was developed, and the system was applied to the Neogene normal sedimentary rock in the Hokushin area, in the town of Horonobe in Hokkaido. A 700-m borehole was drilled, and the applicability of the system was confirmed in 2004. After conducting a seismic reflection survey for the Omagari fault distribution at the Kami-Horonobe area in the town of Horonobe, a drilling site was selected, and a borehole trace was planned to perpendicularly intersect the fault zone in 2005. Considering the planned trace, a 1000-m-long borehole was drilled to a depth of 450 m. From 750 m to the bottom (1000 m), the borehole was horizontal. The total core recovery was 99.8%, even though it was drilled in the fault zone. The geological, hydrological, geo-mechanical, geophysical, and geochemical data were collected using borehole logging/measurement/survey and core logging/measurement/analysis, and the Omagari fault was characterized. After conducting all the work in the borehole, a steel pipe was inserted into the borehole to case the wall. Open hole sections for monitoring were constructed considering the hydro-geological conditions throughout the hole. Then, the monitoring system was inserted and set up to obtain the initial conditions of the groundwater pressure and chemistry. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Lessons Learned Concerning the Onsite Disposal Facility at the Fernald Preserve, Harrison, Ohio - 16176

An onsite disposal facility (OSDF) was constructed at the Fernald Preserve, Ohio, for the disposal of contaminated soil and debris generated during a Comprehensive Environmental Response, Compensation, and Liability Act site cleanup. Completed in October of 2006, the 30 ha facility contains 2.26 million m{sup 3} of waste materials, has a maximum height of approximately 19.8 m, and consists of eight individual cells. The dimensions of cells 1-7 are approximately 213 m x 122 m each, or 26,000 m{sup 2} (2.60 ha). The dimensions of cell 8 (the last cell constructed) are larger than those of the other cells (approximately 3.8 ha). Each cell was constructed with a leachate collection system (LCS) that collected rainwater and storm-water runoff during waste placement and prevented it from entering the underlying environment. Other engineered features include a multilayer composite liner system, a leak detection system (LDS) positioned beneath the primary liner, and a multilayer composite cover that was placed over each cell following the completion of waste-placement activities. The LCS and LDS layers are designed to convey (by gravity) any leachate/fluid that enters the system through pipes to the west side of the facility and into tanks located in valve houses. From there, leachate can be sampled and conveyed to treatment. Groundwater is monitored beneath each cell via monitoring wells. In November 2006, the US DOE Office of Legacy Management (LM) assumed responsibility for the long-term care and maintenance of the OSDF. LM has nine years of monitoring and maintenance experience with the Fernald OSDF. Monitoring data is reported each year in the Fernald Annual Site Environmental Report. The data collected and inspections conducted indicate that the facility is operating as designed. The past nine years has provided LM with the opportunity to evaluate how well many of the design and as-built features of the facility have functioned over time as they pertain to environmental monitoring, valve-house operation and maintenance, and leachate transmission. Lessons learned concerning the long-term care and maintenance of the OSDF at the Fernald Preserve are presented. The main lessons learned include the following: How bivariate plots can be used to address environmental monitoring challenges created by installing an engineered disposal facility in a location where contamination is above the background level but below cleanup standards. How valve-house design decisions can help to alleviate long-term monitoring challenges. How scale buildup can lead to leachate transmission issues. Some additional minor lessons learned are also presented. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Update on implementing geological disposal in the UK - 16266

For more than 60 years, Britain has been accumulating radioactive waste which is currently stored safely at over 20 sites around the country. Radioactive Waste Management (RWM) is responsible for making this waste even safer by disposing of it securely for the long term in a GDF. This paper provides an update on the implementation of the process for siting a Geological Disposal Facility (GDF) in the United Kingdom (UK) and the role of RWM as the developer in its implementation. The UK Government favours an approach for selecting a GDF site that is based on working in partnership with communities. In early 2013 when there were no longer any communities participating in the site selection process, the UK Government embarked upon a review to consider what lessons could be learned. Informed by this review the UK Government published in July 2014 a renewed approach to implementing geological disposal based on working with interested communities. The first two years of the programme comprises of initial actions overseen by Government recognising the need to provide additional information before the siting process is formally launched in 2017. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Low Activity Waste Glass Waste Form Data Package for Hanford's Integrated Disposal Facility Performance Assessment - 16480

A Direct Feed Low Activity Waste (DFLAW) flowsheet has been initiated to begin immobilization of Hanford tank wastes. In the DFLAW configuration, a pretreatment system will segregate the high level and the low activity wastes (LAW). The LAW will be immobilized in the Waste Treatment Plant (WTP) and disposed of in Hanford's Integrated Disposal Facility (IDF). To obtain authorization to dispose of wastes in the IDF, a Performance Assessment (PA) must be completed in accordance with DOE Order 435.1. The PA shall include calculations of potential doses to representative future members of the public and potential releases from the facility to provide a reasonable expectation that performance objectives will not be exceeded as a result of operation and closure of the facility. To support the development of the PA, data from ILAW glass testing are needed to model the long term release of contaminants over the compliance period and verify that impacts to groundwater quality will be within performance objective limits. Washington River Protection Solutions is conducting an ILAW glass testing program to produce the required data. This program is a combined effort involving National Laboratories with known expertise in glass waste form research. The work involves accelerated testing to develop parameters that can be used in computer simulations of glass corrosion and contaminant release in the IDF environment. The results of this work will be compiled in an ILAW glass waste form performance data package for the IDF PA. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Spent Nuclear Fuel Management and Disposal in Finland - Status and Lessons Learned at the End of 2015 - 16507

In November 2015 the Finnish Government approved and the Parliament ratified Posiva's 2012 License Application for Construction of a mined, deep geological repository for the spent nuclear fuel (SNF) at the Olkiluoto site in the Eurajoki municipality based upon recommendations provided by the Finnish Radiation Safety Authority (STUK), international expert groups and long-standing local support. The preceding siting process began in 1983 and it included disposal-concept selection (KBS-3V), 'consent-based' site selection and large-scale in-situ tests conducted in an underground research laboratory, the ONKALO, situated ∼440 m below the mean Sea level immediately adjacent to the planned repository at the Olkiluoto site. At the end of 2015, the Finnish SNF repository is projected to open in 2024 as the first of its kind. Four related lessons learned in Finland during the past 30 years are: 1. The selection of potentially-suitable sites for geological disposal of SNF proved to be more difficult and more time consuming than initially envisioned. 2. The siting of nuclear facilities has evolved from being mainly a radionuclide containment and isolation challenge to becoming a critical moving-target, socio-political challenge governing progress with majority local acceptance being a fundamental imperative. 3. Public and political acceptance and support are strongly linked to the competency, credibility and transparency of the involved organisations and the maturity/pedigree of the proposed disposal concept. 4. Timely sharing lessons learned and conducting selective collaborations with other countries proved to be successful keys to timely and cost-effectively obtain and maintain public and political acceptance and support. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Design of an Evapotranspiration Surface Barrier for the Subsurface Disposal Area at INL - 19220

At the INL Site, the US DOE is planning to construct one of its largest CERCLA landfill covers. An evapotranspiration surface barrier that will be approximately 65 ha (160 acres) will cover the 39-ha (97-acre) Subsurface Disposal Area at the INL to isolate legacy mixed waste disposed of prior to 1988, low-level radioactive waste disposed of after 1988, and debris from demolition of structures being used to process some of the legacy mixed waste. The evapotranspiration surface barrier will limit percolation to less than 1 cm (0.39 in.) per year under typical precipitation conditions; provide a minimum 3-m (10-ft)-thick cover over all waste to minimize the potential for plant, animal, and inadvertent human intrusion; support growth of native vegetation; and visually blend in with the surrounding desert landscape. A robust design will allow the evapotranspiration surface barrier to resist erosion and perform for a minimum of 1,000 years. The 1-cm (0.39-in.)-per-year and 1,000-year performance objectives were established by a CERCLA record of decision and its implementing remedial design work plan, and these performance objectives are consistent with a performance assessment conducted under US DOE Order 435.1, 'Radioactive Waste Management.' (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Implementation of a Risk-Informed Monitoring Program for the Saltstone Disposal Facility - 19374

Section 3116(b) of the Ronald W. Reagan National Defense Authorization Act for Fiscal Year 2005 (NDAA) requires the US NRC, in coordination with the appropriate State agency, to monitor certain disposal actions of the US DOE to assess compliance with the performance objectives in Subpart C of Part 61 of Title 10 of the Code of Federal Regulations (CFR). Under Section 3116(b), the US NRC monitors the US DOE disposal of salt waste at the Saltstone Disposal Facility (SDF) at the Savannah River Site (SRS) in South Carolina. Since the first US NRC monitoring plan for the SDF was issued in 2007, the US NRC staff has made the risk-significant areas of interest more specific. Following the US DOE issuance of the 2009 performance assessment for the SDF and the US NRC review of that document, the US NRC issued its second monitoring plan for the SDF in 2013. The 2013 US NRC monitoring plan included 11 monitoring areas, which were broken down into 40 specific monitoring factors. The US NRC prioritized those monitoring factors (i.e., low, medium, high) based on the sensitivity of model results to the assumptions or processes covered by the monitoring factor and the uncertainty in those assumptions or processes. Those monitoring factors can be opened, closed, revised, and re-prioritized based on new information or changes in the US DOE activities and plans. For example, the list of US NRC high-priority monitoring factors has changed in response to changing information. The flexibility of this approach allows the US NRC to remain focused on risk-significant areas of interest. The US NRC staff has made changes to the NDAA monitoring program at the SDF since the issuance of the 2007 monitoring plan. First, the development of more specific monitoring factors and the prioritization of those monitoring factors has been successful in focusing the US NRC and the US DOE resources on risk-significant issues and providing transparency to stakeholders. Second, the US NRC reviews have become more focused on addressing specific monitoring factors, which has made the process for updating the monitoring plan more transparent. Finally, periodic management meetings and monthly telephone calls maintain open communications between the US NRC, US DOE, South Carolina Department of Health and Environmental Control (SCDHEC), and US EPA. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

The Durability of Containers for the Disposal of Radioactive Waste - Swiss and Belgian R and D Studies - 19392

In this paper, we present a summary of the results from research and development (R and D) studies associated with the durability of steel-based disposal canisters for high-level waste (HLW) and spent nuclear fuel (SNF), focusing on experimental studies carried out in support of the Swiss and Belgian disposal programmes. We present the results of laboratory studies carried out for periods of the order of ten to twenty years, and compare them with more recent observations from in situ experiments carried out at the Mont Terri and Grimsel underground laboratories. For the Belgian programme, which is currently envisaging relatively high gamma dose rates outside the steel overpack, we also show the results of studies focusing on gamma radiation at dose rates compatible with those expected in the disposal system. These studies show how, in bentonite systems, corrosion rates decrease to values lower than 1 μm year{sup -1} after an initial period. In the case of cement-based systems, corrosion rates lower than 0.1 μm year{sup -1} are typically observed, with modest effects of gamma irradiation. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Thermal Analysis for SRS Saltstone Disposal Unit Facility

Saltstone Disposal Units (SDU) are long term disposal units for low activity grout produced from the solidification of decontaminated non-hazardous salt waste at SRS. The unit is a cylindrical disposal vault of 150 ft. in diameter and 23 ft. in height. The total volume the SDU can hold is 3 million gallons of grout. The hydration process of the grout material produces heat from an exothermic chemical reaction. If the grout reaches temperatures higher than the maximum safety temperature of 95 deg. C, cracks can start forming leading to contamination of the environment. The purpose of this study is to develop the optimal baseline domain boundary conditions and to perform a sensitivity analysis with respect to the baseline modeling conditions. The following assumptions were made for the modeling calculations: - The system geometry is axisymmetric; - The system was at steady state; - Ambient temperature was set to 100 deg. F; - Air was assumed to be an incompressible ideal gas; - Radiation was neglected for a conservative estimation; - Natural convection for the SDU vapor space is assumed to be turbulent because Rayleigh number based on typical operating conditions is higher than 10{sup 8}; - Water tables were not taken into consideration. The following sensitivity analyses were performed to find the optimal domain boundary: - Mesh size sensitivity; - Soil depth sensitivity - Soil depth varied from 30 ft to 250 ft with a soil thickness of 90 ft; - Soil thickness sensitivity - Soil thickness varied from 30 ft to 270 ft with a soil depth of 180 ft; - Soil conductivity sensitivity - Soil thermal conductivity varied from 0.25 W/m-K to 1 W/m-K; - Ambient Temperature Sensitivity - Ambient temperature varied from 70 deg. F to 110 deg. F. For the thermal calculations, the temperature of the bottom of the soil was set to 17.4 deg. C. The heat flux on the roof of the building and surface of the earth is 163.2 W/m{sup 2}. The soil sides are set to be adiabatic. The sensitivity analysis shows that the mesh size has no effect on the numerical results. This is mainly due to a small convection zone and large conduction zones in the modeling domain. The optimal domain boundary was found to be a soil thickness of 90 ft and a soil depth of 150 ft. The different sensitivity analysis performed here shows the optimal boundary domain to be used for thermal calculations is a soil depth of 150 ft beneath the SDU and a soil thickness of 90 ft. Anything larger would produce similar solutions but at a longer computational time to solve. The maximum temperature was more sensitive to soil properties as shown here. While the ambient temperature affected the max temperature linearly. Most of the heat transfer takes place downward vertically between the SDU and the soil beneath it. Using the optimal domain boundary determined by the sensitivity analyses, the maximum thermal loading to keep the maximum grout temperature below 95 deg. C is 1.646 W/m{sup 3}.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Origins and Contents of the Waste Specifications: the Example of the French LLW-SL Disposal Facility - 20016

The waste specifications published by Andra for its surface disposal facilities for Low and Intermediate, Short-Lived waste (CSA) are primarily defined to ensure the safe disposal of waste, both in the short and long term, spanning from disposal operations to hundreds of years after closure, when the man-made safety barriers have long disappeared into dust. The paper presents the methodology used at Andra to derive the CSA waste specifications, also known as Waste Acceptance Criteria, their contents, and the process in place to verify compliance of the waste with the specifications. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Deep Isolation-Development of the Safety Case for Disposal of Radioactive Wastes in Horizontal Boreholes - 20028

Deep Isolation has developed a safe, secure, and permanent geological disposal method for high-level waste, including spent nuclear fuel as well as sealed sources and other highly radioactive materials. The method leverages well established directional drilling technology to create horizontal repositories deep underground. The combination of great depth (1-3 km) and the ability to precisely position repositories in a horizontal orientation provides access to geologic strata that are inaccessible for typical mined and deep vertical borehole repositories. A number of potential direct and indirect safety benefits accrue including:1) greater depth below regional freshwater aquifers; 2) increased flexibility and more geologic options for siting; 3) access to deep formations with persistent and sustained reducing conditions. In addition there are a number of safety elements related specifically to the horizontal repository geometry including: 1) passive direction of thermally driven fluid and radionuclide movement away from the vertical access hole and toward the 'dead end' portion of the repository; 2) mitigation of seismic hazards by orienting repositories parallel to local and regional fault structures. In this paper we explore and discuss some of the key technologic, geologic and hydrologic elements that support the deep horizontal borehole safety case. The stalemate seen across the globe on the disposal of spent nuclear fuel and high-level waste can be broken. Deep Isolation offers a novel option for safe, secure, and permanent deep geological disposal of nuclear waste that can be developed as centralized repositories or adapted to smaller regional or site-specific repositories located near waste sources. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Applications of neutron-absorbing structural amorphous metal (SAM) coatings for criticality safety controls of used fuel storage, transportation, and disposal - 14534

Used nuclear fuel contains fissionable materials ({sup 235}U, {sup 239}Pu, {sup 241}Pu, etc.). To prevent nuclear criticality in used fuel storage, transportation, and disposal, neutron-absorbing materials (or neutron poisons, such as borated stainless steel, Boral{sup TM}, Metamic{sup TM}, Ni-Gd, etc.) would have to be applied. Corrosion-resistant, iron-based structural-amorphous metals (SAMs) have been tested to determine their relative corrosion resistance. Many of these materials can be applied as coatings with advanced High Velocity Oxy-Fuel (HVOF) thermal-spray technology. SAM2X5, SAM1651, and other SAMs are amorphous-metal composite alloys that have been identified as having outstanding corrosion resistance. Because of its high boron content, SAM2X5 can be applied as the neutron-absorbing coatings to the metallic support structure for criticality-safety controls of used fuel in racks in wet storage pool and baskets inside the storage containers, the transportation cask, and eventually the disposal containers in repository disposal. Research and experiment conducted at Lawrence Livermore National Laboratory indicated that the high boron-containing SAM2X5 coating could be an effective criticality control material for used fuel management. The neutron irradiation experiments and the neutron transmission measurements conducted at McClellan Nuclear Radiation Center indicated that extensive fast neutron irradiation did not change the structure of the amorphous SAM2X5 melt-spun ribbons, and SAM2X5 exhibited effective neutron absorbing capability, similar to Boral{sup TM} and Metamic{sup TM}. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2016

Data Validation Package May 2016 Groundwater Sampling at the Bluewater, New Mexico, Disposal Site, September 2016

Groundwater samples were collected from monitoring wells at the Bluewater, New Mexico, Disposal Site to monitor groundwater contaminants as specified in the 1997 Long-Term Surveillance Plan for the DOE Bluewater (UMTRCA Title II) Disposal Site Near Grants, New Mexico (LTSP). Sampling and analyses were conducted as specified in the Sampling and Analysis Plan for U.S. Department of Energy Office of Legacy Management Sites (LMS/PRO/S04351, continually updated). A duplicate sample was collected from location 16(SG).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗