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At least 1,135 records · Page 63
All sourcesNASA NTRSDOE OSTI
DOE OSTI2012

Groundwater Impacts of Radioactive Wastes and Associated Environmental Modeling Assessment

This article provides a review of the major sources of radioactive wastes and their impacts on groundwater contamination. The review discusses the major biogeochemical processes that control the transport and fate of radionuclide contaminants in groundwater, and describe the evolution of mathematical models designed to simulate and assess the transport and transformation of radionuclides in groundwater.

radionuclide wastes, fate and transport, modeling,↗
DOE OSTI2024

Synthesizing, Compounding, and Characterizing a Heat Labile Polyurethane Foam

ABSTRACT A need exists for a packaging foam material that can be converted from solid to gaseous degradation products at reasonably low energy levels or temperatures, such as 100O C. This paper will primarily discuss the approaches currently being used to synthesize and characterize such a material. These approaches include the incorporation of novel polyols such as azo containing diols, polycarbonate diols, and polypropylene carbonate polyols into polyurethane foams. Characterization methods include NMR, FTIR, TGA, finite element analysis, impact strength, and others. This project will be funded for a duration of three years. Year 1 focused on developing the proposed test methods and producing an initial rigid polyurethane foam. Year 2 focuses on refining the materials and test methods. If appropriate, design of experiment (doe) techniques will be used to optimize components, component levels, density and other variables to attain required final material properties (TGA weight loss, impact strength, etc.). Year 3 focuses on scaling up to larger engineering quantities. The application for this material is in load securement for transportation of low level radioactive waste materials within the US Department of Energy (DOE) complex. Foam in place process equipment and operators will be shielded using this novel method over current practice. Current practice can involve time consuming methods of load securement in low level radiation environments. This new technique would eliminate exposure time securing the load and greatly improve the As Low As Reasonably Attainable (ALARA) conditions. The objective is to progress to higher Technical Readiness Levels (TRL) and larger pilot scale quantities. This paper discusses methodologies and presents current results to date.

Kranjc, Mark D.↗
DOE OSTI2025

Plutonium Solubility and Supernate Concentration for Neutralized Fast Critical Assembly Discards to Savannah River Site Tank Waste

The Savannah River Site (SRS) plans to dissolve non-irradiated stainless steel (SS)-clad bundles of Fast Critical Assembly (FCA) materials in eighteen batches.1 FCA dissolution is currently underway in the 6.3D dissolver by simultaneous chemical and electrolytic dissolution, which is required to generate the harsh conditions necessary for dissolution of metal-oxide (MOX) and non-aluminum spent nuclear fuels (NASNFs).2 Nitric acid and potassium fluoride are used to promote chemical dissolution.2 Gadolinium will be added during processing as a thermal neutron poison for criticality control. There are no plans for recovering plutonium from this waste stream. After FCA dissolution, the acidic (HNO3/KF) “discards” containing the dissolved metals will be neutralized by addition of 50 wt% sodium hydroxide to a final free hydroxide concentration of 1.2 M.1 Neutralization will precipitate a slurry of insoluble solids, predominantly metal oxides/hydroxides of plutonium, uranium, and SS components. Small fractions of the SS components, Pu, U, and Gd will remain dissolved in the supernate. The neutralized slurry will be composited to existing radioactive waste storage tanks within the SRS Concentration, Storage, and Transfer Facilities (CSTF) containing other similar sludge batch (SB) materials.1 The fate of soluble plutonium and freshly-precipitated, colloidal plutonium from this process are of concern since the total Pu can challenge the waste acceptance criteria (WAC) at the downstream SRS Liquid Waste (LW) facility. Supernate decants including the neutralized FCA discards (nFCAd) within the CSTF will be composited with salt batch (StB) materials and transferred to the SRS Salt Waste Processing Facility (SWPF), where total plutonium is also of concern.

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DOE OSTI2013

NUHOWS - Storage and Transportation of Irradiated Reactor Components in Large Packages - 13439

Most irradiated reactor components (hardware such as Control Rod Blades, Fuel Channels, Poison Curtains, etc.) generated at reactors previously required significant processing for size reduction due to the available transportation casks not being physically capable of containing unprocessed material. As of July 1, 2008, disposal for this typical waste class (B and C) became inaccessible (for the major part of the nation) due to the Barnwell, SC disposal facility being closed to all but its three compact states (CT, NJ and SC). Currently in the United States, most facilities are storing their irradiated hardware on-site in the spent fuel pools. Until recently with the opening of the Waste Control Specialists' Texas disposal facility, utilities faced the challenges of spent fuel pool space and capacity management. However, even with WCS's disposal availability, the site currently has annual Curie limitations for disposal, which will continue to promote interim on-site storage until such time as disposal is available. In response, Transnuclear Inc., (TN) an AREVA company, proceeded with designing a new large Radioactive Waste Container (RWC) that can be used to package irradiated hardware without the need for significant processing. The design features of the RWC allows for intermittent loadings of the hardware for better packaging efficiency, higher packaging density, space savings and reduced cost. This RWC is also compatible with TN's on-site modular vault storage system. Once completely loaded, the RWC can be transported to an on-site storage facility, an off-site storage facility and/or an available disposal facility. To accommodate the transportation, TN has designed a large transportation cask, the MP197HB. As the original design was for transporting fuel, it contains the necessary shielding to allow for the transport of unprocessed irradiated reactor components, while significantly reducing the amount of irradiated hardware shipments required with the use of the new RWC. This paper provides information on the unique design features of the RWC, storage module vaults, MP197HB Transportation Cask and cost saving benefits of using the large RWC for packaging, storage, transport and disposal. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2014

99 Tc Process Monitoring System In-Lab Performance Characterization

A 99 Tc Process Monitoring (Tc-Mon) System has been designed and built for deployment at the recently constructed 200 West Pump & Treat (200W P&T) Plant in the 200 West Area ZP-1 Operable Unit of the Hanford Site. The plant is operated by CH2M Hill Plateau Remediation Company (CHPRC). The Tc-Mon system was created through collaboration between Pacific Northwest National Laboratory (PNNL) and Burge Environmental, Inc. The design of the new system has been optimized based on experience from an earlier field test (2011) of a prototype system at the 200W-ZP-1 Interim Pump & Treat Plant.

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DOE OSTI2017

Low Temperature Waste Form for Supplemental Immobilization of Hanford Low Activity Waste - 17485

The Hanford site has approximately 56 million gallons of radioactive waste stored in 177 underground storage tanks. The Hanford Waste Treatment and Immobilization Plant (WTP) is being constructed to treat the waste but will only have sufficient capacity to treat about one-third of the Low Activity Waste (LAW) portion within the anticipated time-frame for completing the waste treatment mission. The LAW vitrification facility will need to be supplemented with additional LAW vitrification capacity or an alternate LAW immobilization technology. The LAW immobilization technologies considered must be capable of treating approximately 65% of Hanford's LAW and will be evaluated with respect to facilities required, quantity of final waste forms and secondary wastes produced, waste form performance data, technical viability, and life cycle cost and schedule estimates. A low temperature (i.e. non-thermal) waste form and treatment process is being evaluated to provide the additional LAW immobilization capacity. Low temperature, cementitious waste forms have been evaluated previously for immobilizing Hanford tank waste but were not further developed for various reasons. Over the past few years the Hanford Tank Operations Contractor has conducted a technology development program to evaluate a low temperature waste form for immobilization of Hanford LAW including formulation development and testing with a range of simulant compositions spiked with radioactive and hazardous contaminants of concern (COCs). The objectives of this program include; developing a low temperature waste form formulation that is robust with respect to both waste form performance and processing properties over a range of LAW compositions, demonstrating waste form performance with real waste, and conducting Engineering Scale testing to advance the technical maturity of the low temperature immobilization process. The purpose of this paper is to report progress toward these objectives, in particular testing with real waste including additives to enhance technetium and iodine retention. (authors)

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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 OSTI2015

Extending the lifetime of mixer paddles used in the production of a low-level radioactive cementitious waste form - 15624

The salt-stone Production Facility (SPF) at the Savannah River Site (SRS) immobilizes low-level radioactive salt solution as a cementitious waste form known as salt-stone. The mixer used in the SPF is a '10-inch' Readco-Kurimoto continuous mixer with paddles fabricated from a high toughness steel. salt-stone is processed daily at a rate of 9.46 E-03 m{sup 3}/s (150 gallons per minute) in a caustic and radioactive environment. The paddles have degraded much quicker than expected which limits production rates and increases outage times. Due to the high cost of the mixer and the limited maintenance due to the radioactive environment, a combination of material properties and paddle configuration must be employed to extend the lifetime of the paddles. This paper will discuss two wear tests conducted by the Savannah River National Laboratory (SRNL) to determine the optimum paddle configuration to reduce wear on the mixer paddles. The first wear test used the mixer design operated up to December 2011 as the baseline and the second test used the mixer design installed after December 2011 as its baseline. As a result of these tests, the SPF will be implementing a new paddle configuration in the mixer currently installed in the SPF process room. (authors)

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DOE OSTI2018

In Backyards and Bedrooms: Challenges and Successes in Gaining Access to Homes for Radiological Testing - 18204

The Port Hope Area Initiative (PHAI) is a federal environmental clean-up program. Its mandate is the remediation and local, long-term, safe management of approximately 1.7 million cubic metres of historic low-level radioactive waste in the adjacent communities of Port Hope and Port Granby in Southern Ontario, Canada. The PHAI is being undertaken by Canadian Nuclear Laboratories (CNL), on behalf of Atomic Energy of Canada Limited, a federal Crown corporation. It is currently one of the largest environmental remediation projects in Canada The safety of workers, the public and the environment while delivering the PHAI on time and within budget is CNL's number one priority. The PHAI is being carried out as two projects - the Port Hope Project and the Port Granby Project. The Port Hope Project is the larger of the two and includes the cleanup of approximately 1.2 million cubic metres of historic waste from various locations within the community. The Port Hope Project includes a unique and important component - the Property Radiological Survey. This survey involves the interior and exterior investigation of approximately 4,800 small-scale properties - primarily residential - in the Municipality of Port Hope for the presence or absence of historic low-level radioactive waste. The Property Radiological Survey began in 2012 and is being conducted in five campaigns with between 450 and 1,250 properties in each campaign. The survey begins with radon monitoring, followed by interior and exterior gamma radiation tests, and concludes with exterior borehole drilling/soil sampling. The full suite of tests take approximately 18 - 24 months to complete at each property. This paper will describe the communications strategy implemented to obtain consent from thousands of property owners for CNL's contractor to conduct radiological testing in private living spaces such as bathrooms, bedrooms and closets, as well as front yards, backyards and outbuildings. This strategy incorporates tactics used to secure consent from residents who are reluctant to participate in the survey for reasons including privacy issues, family situations, life events and misapprehensions about the survey and/or the PHAI in general. The paper will include information about how residents' concerns are addressed, the stepped approach to securing access from owners who do not respond to repeated attempts at contact, and the importance of close coordination with the contractor to ensure that all components of the investigations are aligned so that work is conducted in a safe, timely and systematic manner. The authors will also describe the process for keeping the regulator and the municipality informed of survey outreach efforts, challenges and successes, and will outline the process for notifying the approximately 800 owners of properties on which historic low-level radioactive waste is expected to be identified and transitioning those owners from the survey to the remediation phase. (authors)

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

Tank Closure Cesium Removal Project CST Simulant Cesium Batch Contact Kinetics Test Results and Column Performance Predictions

To evaluate the cesium removal capabilities of the caustic-washed R9120-B version of crystalline silicotitanate (CST) ion exchange media from Savannah River Site (SRS) waste during Tank Closure Cesium Removal (TCCR) column processing, batch contact kinetics tests were performed at 23 °C in SRS Average simulant. Cesium loading kinetics data was also collected (in current and previous testing; King, et al., 2018a) using the same simulant for an older CST sample (IE- 911) which was caustic-washed by the vendor many years ago and was recently retrieved from drums stored at SRS. (Note that R9120-B and IE-911 are believed to be identical materials, after the R9120-B was caustic washed by SRNL, and differ only by particle size and manufacturing date, and thus can be considered different batches). Although cesium loading kinetics testing can be conducted more rigorously using dynamic column tests, the batch contact method requires less simulant and considerably less experimental effort. Average percent removal data for batch contact samples for the two CST batches are provided in Table ES-1. Faster cesium uptake was observed with IE-911, although the loading rates for the two media batches were similar. After 5.5 hours of CST contact, 81% cesium removal was observed with R9120-B versus 86% removal with IE-911. Although these differences are small, the trend is consistent across the timed results, with the larger bead material always loading slower. These seemingly small differences in percent removal are believed to have a large impact on the column performance. Cesium loading curves for the two CST batches are provided in Figure ES-1 below. Faster cesium loading with IE-911 CST is presumed to be associated with the smaller average particle diameter (408 μm mean volume diameter) versus R9120-B (572 μm diameter). Cesium loading kinetic data was needed to conduct column performance comparisons (i.e., model predictions) for the two CST batches.

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

Implementing a Low Activity Waste Direct Feed Capability in the Hanford Waste Treatment Plant - 17277

The original design of the Waste Treatment and Immobilization Plant (WTP) requires that all waste be processed through the Pretreatment (PT) Facility in order to separate the low-activity waste and high-level waste streams. The U.S. Department of Energy (DOE) has developed a sequenced approach to completing the WTP Project, beginning with feeding some liquid waste directly to the Low-Activity Waste (LAW) Facility, bypassing the PT Facility. Sending low-activity waste from the tank farms to the LAW Facility is referred to as direct-feed LAW (DFLAW). Using this approach, low-activity waste will be pretreated in a Low-Activity Waste Pretreatment System (LAWPS) for cesium and solids separation to meet LAW Facility safety basis requirements and produce an acceptable low-activity waste product. The waste feed will then be transferred to the LAW Facility where it will be immobilized in glass, poured into stainless steel containers, and then transported by truck to the Integrated Disposal Facility, a mixed waste burial ground located in the Hanford 200-East Area. The primary driver for the DFLAW approach is to enable commissioning of the LAW Facility for low-activity waste processing prior to the completion and availability of the PT and High-Level Waste (HLW) facilities. As such, the functions intended to be provided to the LAW Facility by the PT Facility and the common functionality with the HLW Facility must be refined and modified to enable DFLAW operations. The DFLAW approach will provide necessary systems and facilities to replace these enabling PT Facility functions until the PT Facility is completed and commissioned. For example, liquid effluents from the LAW Facility, primarily melter offgas condensate, will be processed in a new effluent management facility (EMF) within the WTP footprint rather than sending the effluent back to the PT Facility for processing. During DFLAW operations approximately 3 785 m{sup 3} (1 million gallons) of double-shell tank (DST) waste will be delivered to WTP for treatment each year of operation, and nominally produce 21 metric tons of immobilized low-activity waste glass per day. The resultant freed space in the DSTs will support safe tank farm operation and staging of additional waste for treatment. In December 2016, DOE executed a contract modification and approved an incremental change to the project's performance baseline to complete the WTP facilities needed to allow direct feed of low-activity waste to the LAW Facility. The revised performance baseline supports completion and cold commissioning of the WTP facilities needed for DFLAW operations no later than August 2023. The design of modifications to WTP facilities needed to enable DFLAW operations is more that 60 percent complete. DOE approved the preliminary safety basis for DFLAW modifications to WTP based on the 30 percent design package in early 2016. Early construction work for DFLAW modifications to WTP began in mid-2015. Long-lead procurement actions for some of the more complex equipment needed for LAW Facility effluent treatment were initiated in early 2016. (authors)

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DOE OSTI2020

Lessons Learned in the Development of Industry Guidance - 20279

The availability of industry guidance on different aspects of waste management forms a valuable reference for waste practitioners. The guidance typically details the principles, processes and practices that should be consulted during each stage of the relevant activities. It also includes relevant regulatory and international requirements, appropriate case studies and examples of best practices to illustrate technical concepts and make the information available to the reader in an accessible manner. Over recent years Wood have developed (or are developing) a number of industry guidance documents on a range of topics including Solid Waste Characterization, Routine Water Quality Monitoring on Nuclear Sites: Pipeline Characterization: Determining Background Radioactivity for Land Quality Assessment: Interim Storage and Aqueous Waste Management. This talk provides an overview of lessons learned along the way including: - the importance of stakeholder engagement through the process: - how to identify 'good practice'; - the importance of iterative development; and - the selection of illustrative and informative case studies. (authors)

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

Tritium Packages and 17th RH Canister Categories of Transuranic Waste Stored Below Ground within Area G

A large wildfire called the Las Conchas Fire burned large areas near Los Alamos National Laboratory (LANL) in 2011 and heightened public concern and news media attention over transuranic (TRU) waste stored at LANL’s Technical Area 54 (TA-54) Area G waste management facility. The removal of TRU waste from Area G had been placed at a lower priority in budget decisions for environmental cleanup at LANL because TRU waste removal is not included in the March 2005 Compliance Order on Consent (Reference 1) that is the primary regulatory driver for environmental cleanup at LANL. The Consent Order is a settlement agreement between LANL and the New Mexico Environment Department (NMED) that contains specific requirements and schedules for cleaning up historical contamination at the LANL site. After the Las Conchas Fire, discussions were held by the U.S. Department of Energy (DOE) with the NMED on accelerating TRU waste removal from LANL and disposing it at the Waste Isolation Pilot Plant (WIPP). This report summarizes available information on the origin, configuration, and composition of the waste containers within the Tritium Packages and 17th RH Canister categories; their physical and radiological characteristics; the results of the radioassays; and potential issues in retrieval and processing of the waste containers.

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DOE OSTI2014

New Mexico state information handbook: formerly utilized sites remedial action program

This volume is one of a series produced under contract with the DOE, by Politech Corporation to develop a legislative and regulatory data base to assist the FUSRAP management in addressing the institutional and socioeconomic issues involved in carrying out the Formerly Utilized Sites Remedial Action Program. This information Handbook series contains information about all relevant government agencies at the Federal and state levels, the pertinent programs they administer, each affected state legislature, and current Federal and state legislative and regulatory initiatives. This volume is a compilation of information about the State of New Mexico. It contains a description of the state executive branch structure; a summary of relevant state statutes and regulations; a description of the structure of the state legislature, identification of the officers and committee chairmen, and a summary of recent relevant legislative action; and the full text of relevant statutes and regulations.

052001 -- Nuclear Fuels-- Waste Processing↗
DOE OSTI2013

Prioritization of remedial actions for radioactive waste burials in the Chernobyl exclusion zone

Large volumes of radioactive waste were urgently buried in the Chernobyl Exclusion Zone (ChEZ) in the aftermath of the 1986 accident. Twenty-six years later, decisions must be taken about whether and how to remediate these sites. An attempt to resolve two key issues is described here: 1. How to assess the hazards posed by these facilities, recognizing that, for the foreseeable future, the Chernobyl Exclusion Zone will remain under institutional control? and 2. What standards to apply in deciding the extent of remediation? This paper presents an examination of the issues and proposes a simple decision-making process. (authors)

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

GRAFEC: A New Spanish Program to Investigate Waste Management Options for Radioactive Graphite - 12399

Spain has to manage about 3700 tons of irradiated graphite from the reactor Vandellos I as radioactive waste. 2700 tons are the stack of the reactor and are still in the reactor core waiting for retrieval. The rest of the quantities, 1000 tons, are the graphite sleeves which have been already retrieved from the reactor. During operation the graphite sleeves were stored in a silo and during the dismantling stage a retrieval process was carried out separating the wires from the graphite, which were crushed and introduced into 220 cubic containers of 6 m{sup 3} each and placed in interim storage. The graphite is an intermediate level radioactive waste but it contains long lived radionuclides like {sup 14}C which disqualifies disposal at the low level waste repository of El Cabril. Therefore, a new project has been started in order to investigate two new options for the management of this waste type. The first one is based on a selective decontamination of {sup 14}C by thermal methods. This method is based on results obtained at the Research Centre Juelich (FZJ) in the Frame of the EC programs 'Raphael' and 'Carbowaste'. The process developed at FZJ is based on a preferential oxidation of {sup 14}C in comparison to the bulk {sup 12}C. Explanations for this effect are the inhomogeneous distribution and a weaker bounding of {sup 14}C which is not incorporated in the graphite lattice. However these investigations have only been performed with graphite from the high temperature reactor Arbeitsgemeinschaft Versuchsreaktor Juelich AVR which has been operated in a non-oxidising condition or research reactor graphite operated at room temperature. The reactor Vandellos I has been operated with CO{sub 2} as coolant and significant amounts of graphite have been already oxidised. The aim of the project is to validate whether a {sup 14}C decontamination can also been achieved with graphite from Vandellos I. A second possibility under investigation is the encapsulation of the graphite in a long term stable glass matrix. The principal applicability has been already proved by FNAG. Crushed graphite mixed with a suitable glass powder has been pressed at elevated temperature under vacuum. The vacuum is required to avoid gas enclosures in the obtained product. The obtained products, named IGM for 'Impermeable Graphite Matrix', have densities above 99% of theoretical density. The amount of glass has been chosen with respect to the pore volume of the former graphite parts. The method allows the production of encapsulated graphite without increasing the disposal volume. This paper will give a short overview of characterisation results of different irradiated graphite materials obtained at CIEMAT and in the Carbowaste project as well as the proposed methods and the actual status of the program including first results about leaching of non-radioactive IGM samples and hopefully first tendencies concerning the C-14 separation from graphite of Vandellos I by thermal treatment. Both processes, the thermal treatment as well as the IGM, have the potential to solve problems related to the management of irradiated graphite in Spain. However the methods have only been tested with different types of i-graphite and virgin graphite, respectively. Only investigations with real i-graphite from Spain will reveal whether the described methods are applicable to graphite from Vandellos I. However all partners are convinced that one of these new methods or a combination of them will lead to a feasible option to manage i-graphite in Spain on an industrial scale. (authors)

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

Integrated salt waste processing at the Savannah River Site - 15301

Treatment and disposition of salt waste is the critical path to completion of the Savannah River Site Liquid Waste (LW) Disposition and Cleanup Program. The new Salt Waste Processing Facility (SWPF) is a critical link in the pre-treatment chain and will process high curie salt waste removed from various SRS liquid waste storage tanks. Significant infrastructure modifications are needed in the existing LW facilities in order to integrate the SWPF into the LW system. (authors)

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

Principles and Tasks of the New Regulatory System for Radioactive Waste Management in the Russian Federation - 12020

This year the Federal Law 'On Radioactive Waste management' was adopted in the Russian Federation. The law significantly changes the existing radioactive waste management regulatory system and assigns a lot of new tasks in order to implement new principles and overcome inevitable respective difficulties. Nuclear Safety Institute was largely involved in the process of the development of the law as well as its further co-ordination among the stakeholders, during which some important initial provisions were excluded. In the paper special features of the Russian safety regulation system for radioactive waste management are analyzed. Most significant requirements adopted by the law as well as tasks and expected difficulties related to its implementation are discussed. (authors)

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