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At least 811 records · Page 45
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Corrective Action Decision Document/Closure Report for Corrective Action Unit 542: Disposal Holes, Nevada Test Site, Nevada, Revision 0, with ROTC 1-2

The purpose of this Corrective Action Decision Document/Closure Report is to provide justification and documentation supporting the recommendation for closure of CAU 542 with no further corrective action. To achieve this, corrective action investigation (CAI) activities were performed from July 19 through August 25, 2006, as set forth in the Corrective Action Investigation Plan for Corrective Action Unit 542: Disposal Holes. The purpose of the CAI was to fulfill the following data needs as defined during the data quality objective (DQO) process: Determine whether contaminants of concern (COCs) are present. If COCs are present, determine the nature and extent. Provide sufficient information and data to complete appropriate corrective actions.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2025

Past Approaches for Spent Nuclear Fuel, High-Level, and Transuranic Waste Disposal in the United States—Part 1: Safety Criteria and Treatment of Uncertainty

The United States (US), with its 50-year experience in developing deep geologic disposal for transuranic waste, spent nuclear fuel (SNF), and high-level radioactive waste (HLW), has much to share with other countries. Yet, other countries are better able to translate the US experience and corresponding policy decisions into solutions sensible for their country when they understand the compliance requirements in US laws and regulations. This paper presents past approaches in the generic and site-specific standards of the US Environmental Protection Agency (EPA) and implementing regulations of the US Nuclear Regulatory Commission (NRC) using the framework provided by (1) key questions of the Blue Ribbon Commission on America’s Nuclear Future, and (2) international consensus standards. Both the 1985 EPA generic standards, as updated in 1993 and applied at the operating Waste Isolation Pilot Plant in southern New Mexico for transuranic waste from atomic energy defense activities, and the EPA 2008 site-specific standards and implementing regulations, as applied at the proposed Yucca Mountain repository in southern Nevada for SNF and HLW, adopt the strategy of using quantitative, probabilistic analysis to assess performance and compliance.

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DOE OSTI2025

Sulfate Conversion of Reillex HPQ Anion Exchange Resin for Disposal (Interim Report)

This report describes preliminary data to validate the Savannah River Plutonium Processing Facility’s (SRPPF) flowsheet for conversion of used Reillex HPQ anion exchange resin from the nitrate form to the sulfate form. The nitrate form is an oxidizer and therefore does not meet acceptance criteria for disposal at the Waste Isolation Pilot Plant (WIPP). The purpose of this study is to develop data to support acceptance for this disposition pathway. Due to the challenges characterizing the nitrate concentration on solid resin, the data developed to date are based upon indirect analysis of the ion exchange column effluent by ion chromatography. These challenges are discussed and two methods for quantification of nitrate directly on the resin are recommended for further development: TGA-MS and permanganate digestion followed by IC. The resin used for this work was provided in the chloride form; this is the form in which resin is supplied by the manufacturer. However, it had to be converted to the nitrate form, which is the form that will be used in SRPPF’s ion exchange process, prior to use in the sulfate conversion experiments. The chloride-form resin was characterized. A lab-scale procedure for the conversion of Reillex HPQ resin from the chloride to nitrate form was validated. The nitrate-form resin was assessed for particle size and chloride concentration to ensure it met SRPPF’s facility specifications. The baseline sulfate conversion flowsheet was tested. However, nitrate was still detectable in the effluent after approximately 10 bed volumes of 1 M sodium sulfate had been passed through the resin bed. Additional experiments were performed to assess the effect of increasing the feed volume, reducing the flowrate, the use of 2 M sulfuric acid instead of sodium sulfate, and the use of irradiated resin. The sulfuric acid test was the only one which provided a nondetectable nitrate concentration (<0.002 M) in the effluent. Detectable nitrate in the column effluent suggests that nitrate is still present on the resin itself.

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DOE OSTI2025

Guiding Principles for Geochemical/Thermodynamic Model Development and Validation in Nuclear Waste Disposal: A Close Examination of Recent Thermodynamic Models for H + —Nd 3+ —NO 3 - (—Oxalate) Systems

Development of a defensible source-term model (STM), usually a thermodynamical model for radionuclide solubility calculations, is critical to a performance assessment (PA) of a geologic repository for nuclear waste disposal. Such a model is generally subjected to rigorous regulatory scrutiny. In this article, we highlight key guiding principles for STM model development and validation in nuclear waste management. We illustrate these principles by closely examining three recently developed thermodynamic models with the Pitzer formulism for aqueous H + —Nd 3+ —NO 3 - (—oxalate) systems in a reverse alphabetical order of the authors: the XW model developed by Xiong and Wang, the OWC model developed by Oakes et al., and the GLC model developed by Guignot et al., among which the XW model deals with trace activity coefficients for Nd(III), while the OWC and GLC models are for concentrated Nd(NO 3 ) 3 electrolyte solutions. The principles highlighted include the following: (1) Principle 1. Validation against independent experimental data: A model should be validated against experimental data or field observations that have not been used in the original model parameterization. We tested the XW model against multiple independent experimental data sets including electromotive force (EMF), solubility, water vapor, and water activity measurements. The results show that the XW model is accurate and valid for its intended use for predicting trace activity coefficients and therefore Nd solubility in repository environments. (2) Principle 2. Testing for relevant and sensitive variables: Solution pH is such a variable for an STM and easily acquirable. All three models are checked for their ability to predict pH conditions in Nd(NO 3 ) 3 electrolyte solutions. The OWC model fails to provide a reasonable estimate for solution pH conditions, thus casting serious doubt on its validity for a source-term calculation. In contrast, both the XW and GLC models predict close-to-neutral pH values, in agreement with experimental measurements. (3) Principle 3. Honoring physical constraints: Upon close examination, it is found that the Nd(III)-NO 3 association schema in the OWC model suffers from two shortcomings. Firstly, its second stepwise stability constant for Nd(NO 3 ) 2+ (log K 2 ) is much higher than the first stepwise stability constant for NdNO 3 2+ (log K 1 ), thus violating the general rule of (log K 2 –log K 1 ) < 0, or $\frac{K1}{K2}$>1. Secondly, the OWC model predicts abnormally high activity coefficients for Nd(NO 3 ) 2 + (up to ~900) as the concentration increases. (4) Principle 4. Minimizing degrees of freedom for model fitting: The OWC model with nine fitted parameters is compared with the GLC model with five fitted parameters, as both models apply to the concentrated region for Nd(NO 3 ) 3 electrolyte solutions. The latter appears superior to the former because the latter can fit osmotic coefficient data equally well with fewer model parameters. The work presented here thus illustrates the salient points of geochemical model development, selection, and validation in nuclear waste management.

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NASA NTRS2022

Un-Numbered Operational Areas PRL 229 Railroad Tie Disposal Area Pond Confirmation Sampling Report

This Confirmatory Sampling (CS) Report (CSR) was prepared for the National Aeronautics and Space Administration (NASA), Kennedy Space Center (KSC), Florida in accordance with the CS Work Plan (CSWP) (HGL, 2022a). The CSR summarizes investigation activities performed on the Railroad Tie Disposal Area (RTDA) of the Un-Numbered Operational Areas (UNOA), Potential Release Location (PRL) 229 located at KSC. HydroGeoLogic, Inc. (HGL) prepared this report under Contract Number 80KSC019D0012. The purpose of the CSR is to provide information to evaluate the level of potential contamination in surface water and sediment samples collected from the RTDA PRL 229 Pond and determine if contaminants pose unacceptable risks to ecological receptors. The CSR provides analytical data of 10 sets of surface water and sediment samples collected in 2021. Surface water samples were analyzed for volatile organic compounds (VOCs), semi-volatile organic compounds (SVOCs), polynuclear aromatic hydrocarbon (PAHs), metals, hardness, and salinity, while sediment samples were analyzed for metals, total organic carbon (TOC), SVOCs, PAHs, and total petroleum hydrocarbons (TPH). The CSR includes a Screening Level Ecological Risk Assessment (SLERA) conducted under KSC’s Resource Conservation and Recovery Act (RCRA) Corrective Action Program to determine if unacceptable risks to ecological receptors exist at PRL 229 Pond. Metals and PAHs were detected in surface water samples, including detections of antimony, arsenic, barium, calcium, copper, lead, magnesium, thallium, zinc, anthracene, benzo(a)anthracene, benzo(a)pyrene, benzo(b)fluoranthene, benzo(g,h,i)perylene, benzo(k)fluoranthene, chrysene, dibenzo(a,h)anthracene, fluoranthene, indeno(1,2,3-c,d)pyrene and pyrene. Based on measured salinity (1.03 to 1.21 parts per thousand), the pond was identified as freshwater. Sediment analytical results also contained detections of metals and PAHs, including antimony, arsenic, barium, beryllium, cadmium, chromium, copper, lead, mercury, nickel, selenium, thallium, zinc, anthracene benzo(a)anthracene, benzo(a)pyrene, benzo(b)fluoranthene, benzo(g,h,i)perylene, benzo(k)fluoranthene, chrysene, dibenzo(a,h)anthracene, fluoranthene, indeno(1,2,3-c,d)pyrene and pyrene. Potential risks to ecological receptors were evaluated in accordance with the SLERA Work Plan and the Decision Process Document for the Resource Conservation and Recovery Act (RCRA) Corrective Action Program at KSC (DPD) (Geosyntec, 2019), and using the 2021 sediment and surface water analytical results. Potential ecological receptors evaluated in the SLERA include the aquatic community, benthic invertebrates, aquatic birds, and aquatic mammals. The SLERA included Steps 2a through 2c of the Ecological Risk Assessment (ERA) process in accordance with the DPD. Using multiple lines of evidence, the SLERA concluded that many of the detected analytes are present at concentrations similar to those in background or reference locations at KSC, and/or at concentrations that would not pose a significant risk to ecological communities. There are no screening values available for the avian community, and PAHs are bioaccumulative compounds. For this reason, food web modeling was completed as a supplement to Step 2c to provide a quantitative evaluation of this exposure pathway. Food web modeling indicates that PAHs are unlikely to pose a risk to birds exposed to site sediment and surface water. Based on the results of the SLERA, the contaminants in sediment and surface water pose no to minimal risk to the ecological communities that could live or forage at the site. Based on the results of the SLERA, no further study is recommended for the RTDA Pond at PRL 229.

Confirmation Sampling↗
DOE OSTI2015

RCRA PART B Permit Application Nevada National Security Site (NNSS) for Waste Management Activities at the NNSS Explosive Ordnance Disposal Unit (EODU)

This volume of the Resource Conservation and Recovery Act (RCRA) permit application by the Department of Energy, National Nuclear Security Administration Nevada Field Office (NNSA/NFO) seeks continued operation of the Explosive Ordnance Disposal Unit (EODU) at the Nevada National Security Site (NNSS). Explosives, explosive materials, and unexploded ordnance are declared waste explosives if they are damaged, are defective, have exceeded their shelf life, are the result of environmental restoration (legacy), or are discarded explosive material from research or testing projects. This document provides a description of the regulated waste codes accepted for treatment (controlled detonations), the general design of the unit, and the processes used to safely detonate onsite-generated waste explosives and unexploded ordnance, which are defined as hazardous waste under specific regulations. The Area 11 EODU consists of a firing point (controlled access area where firing of explosives is conducted), firing pad (prepares site where explosives hazardous waste is fired), and ancillary equipment that includes an explosives magazine and firing point area. Additionally, live explosives training is conducted to ensure proficiency of the EODU operations personnel, explosives handlers, and the firing system components. Training detonations do not include waste explosives and are limited in respect to explosives volume and frequency personnel, explosives handlers, and the firing system components. Training detonations do not include waste explosives and are limited in respect to explosives volume and frequency. For safety and security reasons, some of the contents of the full application are not available for public viewing. This public version of the application contains most, but not all, of the contents as submitted to the regulatory agencies.

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DOE OSTI2015

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

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.

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DOE OSTI2017

Ethical Considerations in Determining the Societal Investment in Warning Future Generations about Latent Long-term Risks from Radioactive Waste Disposal Facilities - 17058

Writers and philosophers have long romanticized and pondered the great challenge of communicating with future generations. The only way that man can travel through time. Never has this notion been more studied and debated than in the nuclear waste disposal arena. Regulatory requirements in the United States and internationally, require that there be some form of notification of future generations in place once a deep geologic nuclear waste repository has been closed. It is considered a generational responsibility. Since the difficulty of this task is so daunting, it has been assumed that the monetary cost to accomplish this millennia-spanning message-in-a-bottle must also be extreme. This paper will discuss both the ethical and equitable implications across generations, as they pertain to the reality of future inadvertent intrusion scenarios and the associated cost of communicating through time to prevent them. (authors)

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DOE OSTI2017

Representation of Global Climate Change in Performance Assessment Models for Disposal of Radioactive Waste - 17183

Evolving concepts of global climate change that include impacts of anthropogenic greenhouse gases require reassessment of conventional conservative approaches in performance assessment studies that assume cold and wet future glacial climates. Global paleo-climate data for the Quaternary (past 2 million years) show saw-tooth patterns (temperature/proxy temperature data versus time) of glacial and interglacial intervals accompanied by cyclical changes in the atmospheric content of CO{sub 2} (low contents during glacial cycles and higher contents during interglacial cycles). These cycles are correlated with changes in the earth's orbital configurations (Milankovitch cycles) but require nonlinear linkages with additional climate-forcing components to reproduce past glacial and interglacial transitions. The earth is currently in an interglacial period and the timing of the inception of the next glacial period is critically important for forecasting future climate states. Climate modeling studies coupling Milankovitch cycles with variable atmospheric CO{sub 2} content demonstrate conclusively that the earth is very unlikely to return to a glacial state when CO{sub 2} concentrations remain above pre-industrial levels (∼ 280 parts per million; current levels exceed 400 parts per million). A return to pre-industrial concentrations will likely require > 50,000 years and may require hundreds of thousands of years even if current emissions were to drop dramatically. The most likely scenario for the future global climate is persistence of the current interglacial period with variable but progressively increasing temperatures. Performance assessments need to revise site-specific evaluations of future climate assuming global warming within a prolonged interglacial period with increased probability of extreme weather events. Four radioactive waste disposal sites are illustrated that currently incorporate or will incorporate revised current climate concepts in modeling studies for performance assessments. (authors)

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DOE OSTI2017

Analysis of the Needs for Long Term Memory and Knowledge Preservation Relating to Radioactive Waste Disposal Facilities - 17213

Proposals for a set of provisions for long-term memory and knowledge to be defined for a radioactive waste disposal facility, near surface or deep underground, address two primary motives, related to two ethical principles. The first motive is to prevent future generations from interfering involuntarily with the repository. This requires maintaining awareness of the repository, and addresses the ethical principle of protection of man and environment. The second motive is to provide future generations all the available relevant information which might help them make informed decisions about intentional actions, and assess the consequences. This requires transmitting detailed knowledge of the repository, and addresses the ethical principle of preservation of freedom of action. The set of provisions to be implemented with respect to each of these motives may not be the same. In order to define and assess the set of provisions, it is also useful to identify the various components of the process of transmission of a given message, or set of messages, to future generations. Three sub-processes have been identified: (i) 'memorization', at the producer stage, where a full set of information to be transmitted is identified, organized and expressed; (ii) 'preservation', where the potential durability of records is extended, the preservation conditions are controlled and where the records may be restored, if their status is degraded; (iii) 'access', at the receiver stage, where the receiver has to be notified of the existence of the information, to find it and to interpret it properly. As a failure of transmission to future generations would result from the failure of any of the subprocesses, a minimal set of provisions may be defined from this decomposition. (authors)

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DOE OSTI2017

Preliminary Study on Recycling and Reuse of Metals from Decommissioning in Construction of Disposal Vaults in Slovakia - 17217

The decommissioning of nuclear installations represents a complex process resulting in the generation of large amounts of various waste materials. Majority of them are non-contaminated, however some of them may be hazardous as well (e.g. chemicals, asbestos, oil products, etc.). During decommissioning of NPP, significant quantities of materials with various levels of radionuclides concentration arise as well. All of these materials should be managed appropriately keeping in mind safety principles, legislative requirements, lessons learned and other relevant factors. Therefore, selection of an appropriate strategy for management of the materials arising from decommissioning is crucial and strongly influences the overall effectiveness of decommissioning process. In line with international incentives for optimization of radioactive material management, concepts of recycling and reuse of slightly radioactive materials are widely discussed and applications of these concepts are analyzed. Therefore, the paper is devoted to the preliminary study of potential for recycling of re-melted metals from decommissioning in form of reinforcement bars or mesh used in the construction of ferro-concrete disposal vaults within the near surface repository. The paper presents the preliminary results and discusses the technical and non-technical aspects related to the recycling scenario. Moreover, comparison of various scenarios related to the scrap metal clearance and recycling within nuclear industry is included. Key findings and lessons learned from study may be useful for the international community dealing with waste management optimization and in particular with recycling of materials within nuclear industry. (authors)

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DOE OSTI2015

Numerical modeling of groundwater flow in the Navajo sandstone aquifer at the Tuba city, Arizona, disposal site - 15167

A numerical model is being developed to describe groundwater flow in the Navajo Sandstone aquifer that underlies the US DOE Office of Legacy Management Tuba City disposal site that is located near Tuba City, Arizona. The site is a former uranium-ore processing mill that operated from 1956 until 1966. Resulting groundwater contamination extends approximately 457 meters (1,500 feet) off the site. The primary contaminants are nitrate, sulfate, and uranium. Onsite solid waste encapsulation was completed by 1990 and pump-and-treat groundwater remediation was implemented in 2002. A site conceptual model, along with an approach to numerically simulate groundwater flow and advective transport of contaminants at the site, is being developed. Numerical model calibration will address transient periods of active and inactive groundwater withdrawal during the remedial action. The calibrated model will be applied to forecast contaminant capture under the current and hypothetical groundwater extraction scenarios. (authors)

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DOE OSTI2015

Low and intermediate level short lived waste disposal in France: improving confidence - 15254

Thanks to the significant reduction of waste production and according to forecasts derived from the French National Inventory of Radioactive Materials and Waste issued by Andra, Centre de l'Aube disposal facility should be able to accommodate low and intermediate level short lived waste generated by all presently operated or decided nuclear facilities in France. The facility could then be operated till 2050-2060. Andra has to maintain the facility up to date and to adapt it to meet waste generators needs, in particular needs related to decommissioning works. Its safety is assessed by periodical safety review and confirmed by an environmental monitoring program. Safety also relies on an acceptance process that takes benefit from 20 years of experience and must now be adapted to the management of decommissioning waste. Confidence by local communities is maintained through a permanent dialogue. (authors)

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DOE OSTI2015

Mock-Up activities for the extraction of samples from a salt-stone disposal unit - 15502

Mock-up testing with respect to the core-drilling of simulant salt-stone grout was conducted to support the future retrieval of field-emplaced salt-stone samples from salt-stone Disposal Unit (SDU) Cell 2A. The primary objectives of extracting samples from SDU Cell 2A are to enable measurement of field-emplaced samples with respect to properties that influence long-term performance, such as saturated hydraulic conductivity, and to compare the measured properties for field-emplaced and comparable samples prepared in the laboratory. Initial mock-up activities (Phases 1 and 2) were conducted in Fiscal Years (FY) 2013 and 2014 in order to demonstrate the viability of standard core-drilling techniques that are typically employed for sampling concrete structures. The primary requirement for the core-drilling technique was that it should minimize the potential of physical or chemical sample changes that would artificially degrade key grout properties. Mock-up activities were aimed at optimizing the techniques for core-drilling, and subsequent sample extraction, using a set-up that mimicked, to the extent practical, both the physical characteristics, and operational complexities, of core-drilling salt-stone grout from the roof of the SDU. Samples extracted from the grout monolith were subsequently characterized with respect to saturated hydraulic conductivity and compressive strength, and compared to data for cast samples prepared using the same simulant salt-stone material. The comparison of these properties for the two sample types was expected to highlight potential flaws (e.g., micro-cracking) introduced by the core-drilling process. While the data for saturated hydraulic conductivity were virtually identical (approximately 2 E-09 cm/sec), the compressive strength of the core-drilled samples (approximately 3.5 MPa (500 psi)) was on average half that measured for the cast samples. It was suspected that increased water saturation and surface scoring of the core-drilled samples might account for the disparate compressive Strength data. During Phases 1 and 2 it was determined that cores could be successfully drilled and extracted to a depth of approximately 0.9 meters (3 feet) below the grout surface. Phase 3 (FY 2105) has been initiated and is primarily directed at demonstrating the ability to core-drill through approximately 1.8 meters (6 feet) of overlying grout to retrieve the desired sample from the SDU. The original drill depth of 0.6 - 0.9 meters (2-3 feet) was based on an SDU fill height of 5.6 meters (18.5 feet) and sample location of approximately 4.9 meters (16 feet); recently the allowable fill height in the SDUs has been increased to 6.5 meters (21.25 feet); hence the increased depth of overlying grout above the samples to be retrieved. (authors)

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DOE OSTI2015

Approach for the isotopic characterization of irradiated start-up sources for disposal at WIPP - 15674

The National Nuclear Security Administration, Global Threat Reduction Initiative (GTRI)/ Off-site Source Recovery Program (OSRP) requires an approved final disposition pathway for sealed irradiated startup sources with uncertain irradiation histories. To create this path a reliable method is required to determine yields of transmutation, fission and activation products contained within the source at the time of disposal. Because of the uncertain histories of these sources, a number of likely histories are modeled with the code ORIGEN in order to establish the bounds of the radioisotope yields. The ORIGEN simulations are confirmed with gamma spectroscopy, specifically examining the activities of several fission and transmutation isotopes. These calculations, coupled with other previously established characterization techniques for transuranic neutron sources, will be used to establish a waste profile that is in compliance with Waste Isolation Pilot Plant (WIPP) waste acceptance criteria. (authors)

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DOE OSTI2016

Directional Drilling Technology for HLW Disposal - Long Term Monitoring in the Borehole - 16079

Central Research Institute of Electric Power Industry (CRIEPI) has been conducting a project for the directional drilling and measurement/logging technologies to survey hydro-geological condition at the stage of the preliminary investigation (PI) of site selection for the deep geological disposal of HLW. 1000 m length borehole which intersects the Omagari fault was drilled by directional drilling technology in order to verify the drilling technology in Hokkaido. Long term monitoring to clarify pore water pressure and water chemistry in the drilled borehole was also developed. We set measurement intervals on a hanging wall side of a main fracture and started the measurement of pore water pressure in February 2014. We assume that long term monitoring for several boreholes will be achieved in vast areas where power supply and/or communication networks are not available at the PI stage. Thus, we have been developing an automatic data acquisition system for long term monitoring, which enables us to access data remotely. The monitoring system, which was set in Horonobe-town in Hokkaido, is an automatic data acquisition system that performs long term monitoring of pore water pressure. This paper provides an overview of the development of monitoring technologies of pore water pressure measurement in the borehole as well as the automatic data acquisition system. (authors)

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