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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.

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At least 1,585 records · Page 88
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Transition of Spent Nuclear Fuel to Dry Storage: Modeling activities concerning aluminum spent nuclear fuel cladding integrity

In 2017, a report was prepared by the Spent Nuclear Fuel Working Group in 2017 and titled “Technical Considerations and Challenges for Extended (>50 Years) Dry Storage” [1-2]. In this report several key scientific issues were identified pertaining to the problem of aluminum spent nuclear fuel (ASNF) and its safe removal from wet storage with subsequent drying and transfer to dry storage at the Department of Energy (DOE) Idaho National Laboratory (INL). This report outlines two modeling analyses performed for aluminum and aluminum oxides. The first analysis examined the solubility of aluminum oxides as a function of temperature and pH. This report presents existing literature produced as far back as the 1950’s and including significant analysis and reporting performed at Alcoa in the 1980’s. The primary technical gap was the lack of solubility data beyond that reported at standard temperature (25°C). The analysis examined these solubility curves as a function of temperature as well as phase type. The pH value of minimum solubility shifted with temperature but was insensitive to the oxide phase type. The solubility increases with temperature as would be expected, where an order of magnitude increase was observed from 25°C and 90°C. This data provides greater information to understand the various conditions experiences by ASNF including in dry storage in cases where water filming may occur. A second thermodynamic analysis was performed to examine the aluminum phase behavior as a function of temperature. This was performed to understand limitations and possible problems with fuel drying at elevated temperatures. Based on what has been established in previous reports, water remains chemically entrained within boehmite until temperatures exceeding 400°C. Analysis showed that significant phase changes occurred with highly alloyed AA6061 and AA5052 while the more pure aluminum (AA1100) showed nearly ideal behavior (i.e. limited phase changes). In the case of the more alloyed AA5052 and AA6061, the phase changes could take place at relatively low temperatures (100°C to 250°C), and thus may have already experienced some of these phase changes in service. As results are thermodynamic in nature, assessment of kinetics will be examined to determine if the phase changes are a concern for the relatively short fuel drying step.

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

Gadolinium Oxalate Solubility Measurements in Nitric Acid Solutions

HB-Line will begin processing Pu solutions during FY2012 that will involve the recovery of Pu using oxalate precipitation and filtration. After the precipitation and filtration processes, the filtrate solution will be transferred from HB-Line to H-Canyon. The presence of excess oxalate and unfiltered Pu oxalate solids in these solutions create a criticality safety issue if they are sent to H-Canyon without controls in H-Canyon. One approach involves H-Canyon receiving the filtrate solution into a tank that is poisoned with soluble gadolinium (Gd). Decomposition of the oxalate will occur within a subsequent H-Canyon vessel. The receipt of excess oxalate into the H-Canyon receipt tanks has the potential to precipitate a portion of the Gd poison in the receipt tanks. Because the amount of Gd in solution determines the maximum amount of Pu solids that H-Canyon can receive, H-Canyon Engineering requested that SRNL determine the solubility of Gd in aqueous solutions of 4-10 M nitric acid (HNO 3 ), 4-12 g/L Gd, and 0.15-0.25 M oxalic acid (H 2 C 2 O 4 ) at 25 C. The target soluble Gd concentration is 6 g/L. The data indicate that the target can be achieved above 6 M HNO 3 and below 0.25 M H 2 C 2 O 4 . For 6 M HNO 3 , 10.5 g/L and 7 g/L Gd are soluble in 0.15 M and 0.25 M H 2 C 2 O 4 , respectively. In 4 M HNO 3 , the Gd solubility drops significantly to 2 g/L and 0.25 g/L in 0.15 M and 0.25 M H 2 C 2 O 4 , respectively. The solubility of Gd at 8-10 M HNO 3 exceeds the solubility at 6 M HNO 3 . The data for 4 M HNO 3 showed good agreement with data in the literature. To achieve a target of 6 g/L soluble Gd in solution in the presence of 0.15-0.25 M oxalate, the HNO 3 concentration must be maintained at or above 6 M HNO 3 .

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

Spatiotemporal Analyses of Groundwater and Shoreline Cr(VI) Concentrations in the 100 Areas at Hanford

Cleanup efforts have been ongoing since the late 1990s to remediate contaminated waste sites and groundwater in the 100 Areas at the U.S. Department of Energy (DOE) Hanford Site. One of the primary contaminants of concern is hexavalent chromium (Cr(VI)), which was used as a corrosion inhibitor in cooling water for nuclear reactors that formerly operated along the shoreline of the Columbia River. Cleanup efforts have included 1) removal, treatment (as needed), and disposal of contaminated sediments; 2) in situ redox manipulation as a permeable reactive barrier; 3) pump-and-treat; 4) soil flushing; and 5) monitored natural attenuation. DOE’s annual groundwater monitoring reports document the significant reductions in Cr(VI) plume areas that have occurred over the past 10 years or more as a result of these cleanup efforts. The Record of Decision for the 100-HR-3 operable unit specified a cleanup level (CUL) for Cr(VI) in groundwater of 48 µg/L to protect human receptors, and a surface water CUL of 10 µg/L to protect aquatic organisms in the Columbia River. The Record of Decision did not specify point-of-compliance locations for the surface water CUL. Data for 2019 from the six groundwater operable units (OUs) in the 100 Areas indicate that the 48 μg/L groundwater CUL has been achieved in 100% of the wells in the 100-BC and 100-NR OUs, and in 89- 97% of the wells in the other OUs (100-KR, 100-HR-D, 100-HR-H, 100-FR). Data for 2019 indicate that 100% of the aquifer tubes monitored for Cr(VI) in the 100 Areas have concentrations below the 48 μg/L groundwater CUL. However, the 10 μg/L standard has not yet been consistently achieved for both inland groundwater monitoring wells and shoreline aquifer tubes. This report describes a series of data analyses performed to identify consistent relationships, if any, between inland well and shoreline Cr(VI) concentrations within the 100 Areas. To this end, select monitoring data for Cr(VI) measured in groundwater and aquifer tubes at the 100 Areas were analyzed for a 10-year period—2010 to 2019. Relationships between inland groundwater plumes and surface-water points of discharge in and along the Columbia River were examined through several analyses that included inland well and aquifer tube concentrations as a function of distance from the shoreline, evaluation of cumulative probability plots, trend analysis, correlation analysis, cluster analysis, and identification of plume trajectories for each of the 100 Areas. The analyses did not identify consistent relationships between inland groundwater Cr(VI) concentrations and shoreline concentrations within the 100 Areas due to several confounding factors influencing groundwater flow directions and Cr(VI) concentrations. The proximity of groundwater Cr(VI) plumes to the river, and the highly dynamic nature of the river, influence the transport behavior of the plumes and create challenges for quantifying attenuation of Cr(VI) between the inland monitoring wells and shoreline concentrations. Other factors contributing to temporal and spatial Cr(VI) concentrations, as supported by some of the data analyses, include the presence of vadose zone sources, variable sorption behavior, and complexities associated with Cr(VI) mass transfer between the upper and lower aquifers and their interactions with the river. Hence, monitoring to assess compliance with target CULs will need to be determined for each area individually since several factors influence Cr(VI) concentrations in the 100 Areas.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2013

High linear energy transfer degradation studies simulating alpha radiolysis of TRU solvent extraction processes

Treatment of used nuclear fuel through solvent extraction separation processes is hindered by radiolytic damage from radioactive isotopes present in used fuel. The nature of the damage caused by the radiation may depend on the radiation type, whether it be low linear energy transfer (LET) such as gamma radiation or high LET such as alpha radiation. Used nuclear fuel contains beta/gamma emitting isotopes but also a significant amount of transuranics which are generally alpha emitters. Studying the respective effects on matter of both of these types of radiation will allow for accurate prediction and modeling of process performance losses with respect to dose. Current studies show that alpha radiation has milder effects than that of gamma. This is important to know because it will mean that solvent extraction solutions exposed to alpha radiation may last longer than expected and need less repair and replacement. These models are important for creating robust, predictable, and economical processes that have strong potential for mainstream adoption on the commercial level. The effects of gamma radiation on solvent extraction ligands have been more extensively studied than the effects of alpha radiation. This is due to the inherent difficulty in producing a sufficient and confluent dose of alpha particles within a sample without leaving the sample contaminated with long lived radioactive isotopes. Helium ion beam and radioactive isotope sources have been studied in the literature. We have developed a method for studying the effects of high LET radiation in situ via {sup 10}B activation and the high LET particles that result from the {sup 10}B(n,a){sup 7}Li reaction which follows. Our model for dose involves solving a partial differential equation representing absorption by 10B of an isentropic field of neutrons penetrating a sample. This method has been applied to organic solutions of TBP and CMPO, two ligands common in TRU solvent extraction treatment processes. Rates of degradation of TBP and CMPO and their respective degradation products in the presence of high LET radiation are presented and discussed. These results are also compared to gamma studies performed in our lab and other gamma and alpha studies found in the literature. The possible application of this method to a variety of other solvent extraction ligands to study the effects of high LET radiation is also considered. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Reactor Vessel and Reactor Vessel Internals Segmentation at Zion Nuclear Power Station - 13230

Zion Nuclear Power Station (ZNPS) is a dual-unit Pressurized Water Reactor (PWR) nuclear power plant located on the Lake Michigan shoreline, in the city of Zion, Illinois approximately 64 km (40 miles) north of Chicago, Illinois and 67 km (42 miles) south of Milwaukee, Wisconsin. Each PWR is of the Westinghouse design and had a generation capacity of 1040 MW. Exelon Corporation operated both reactors with the first unit starting production of power in 1973 and the second unit coming on line in 1974. The operation of both reactors ceased in 1996/1997. In 2010 the Nuclear Regulatory Commission approved the transfer of Exelon Corporation's license to ZionSolutions, the Long Term Stewardship subsidiary of EnergySolutions responsible for the decommissioning of ZNPS. In October 2010, ZionSolutions awarded Siempelkamp Nuclear Services, Inc. (SNS) the contract to plan, segment, remove, and package both reactor vessels and their respective internals. This presentation discusses the tools employed by SNS to remove and segment the Reactor Vessel Internals (RVI) and Reactor Vessels (RV) and conveys the recent progress. SNS's mechanical segmentation tooling includes the C-HORCE (Circumferential Hydraulically Operated Cutting Equipment), BMT (Bolt Milling Tool), FaST (Former Attachment Severing Tool) and the VRS (Volume Reduction Station). Thermal segmentation of the reactor vessels will be accomplished using an Oxygen- Propane cutting system. The tools for internals segmentation were designed by SNS using their experience from other successful reactor and large component decommissioning and demolition (D and D) projects in the US. All of the designs allow for the mechanical segmentation of the internals remotely in the water-filled reactor cavities. The C-HORCE is designed to saw seven circumferential cuts through the Core Barrel and Thermal Shield walls with individual thicknesses up to 100 mm (4 inches). The BMT is designed to remove the bolts that fasten the Baffle Plates to the Baffle Former Plates. The FaST is designed to remove the Baffle Former Plates from the Core Barrel. The VRS further volume reduces segmented components using multiple configurations of the 38i and horizontal reciprocating saws. After the successful removal and volume reduction of the Internals, the RV will be segmented using a 'First in the US' thermal cutting process through a co-operative effort with Siempelkamp NIS Ingenieurgesellschaft mbH using their experience at the Stade NPP and Karlsruhe in Germany. SNS mobilized in the fall of 2011 to commence execution of the project in order to complete the RVI segmentation, removal and packaging activities for the first unit (Unit 2) by end of the 2012/beginning 2013 and then mobilize to the second unit, Unit 1. Parallel to the completion of the segmentation of the reactor vessel internals at Unit 1, SNS will segment the Unit 2 pressure vessel and at completion move to Unit 1. (authors)

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

The Relationship Between Dose Rate and Decay Heat for Spent Nuclear Fuel Casks

Decay heat and dose rate are two important limits used for determining the allowable contents of spent fuel (SNF) in dry storage systems and transportation packages. While the decay heat limit is used to maintain fuel cladding integrity and ensure retrievability, dose rates are used to demonstrate compliance with regulatory requirements on radiation protection. Because both dose rate and decay heat result from decay of radioisotopes in SNF, this study is an attempt to examine the relationship between dose rate and decay heat for a given cask design. Dose rates were evaluated for 198 cask configurations, that include various SNF system designs (e.g., storage, transfer, transport), SNF characteristics (e.g., fuel types, burnup, cooling time), and loading maps (e.g., uniform loading, zone loading), while a constant decay heat was maintained. The decay heat was calculated using US Nuclear Regulatory Commission (NRC) Regulatory Guide (RG) 3.54, Revision 2, and verified using ORIGEN sequence within SCALE code system. The ORIGEN outputs were used as source terms in the dose analysis using 198 different configurations. A computer script was developed to calculate the cooling time necessary to achieve a given decay heat for a given enrichment, assembly average burnup, assembly mass, and in-core history using a rootfinder algorithm. Initially the computer script was developed to provide cooling time and burnup calculations directly to the analysis of dose rates and decay heats, so a comparison between Used Nuclear Fuel-Storage Transportation and Disposal Analysis Resource Data System (UNF-ST&DARDS) results and RG3.54 data was made; results are included in the appendix to this document. However, an iterative approach was used to compute cooling time, and its accuracy did not depend on the results of the RG3.54r2 algorithm, although the algorithm was still used. The results of the evaluation presented herein clearly demonstrate that a given decay heat does not correspond to a unique dose rate for a variety of cask and package designs. There is no clear pattern to develop a correlation between decay heat and the source terms. Depending on burnup, enrichment, cask type, and loading pattern, dose rates varied for the exact same decay heat—in some cases by 400% for a given cask. For cases in which decay heat was held constant through selection of the appropriate cooling time, dose rates would decrease with increasing burnup, and in other cases, dose rates would increase. The large variation in dose rates for a constant decay heat indicates that casks loaded based on decay heat—that is allowing any burnup, cooling time, and enrichment combinations that yield the qualified decay heat limit(s) —cannot ensure that an Independent Spent Fuel Storage Installation or a spent fuel transportation package will meet the regulatory limits set forth by the respective regulations, i.e., 10 CFR 72 or 10 CFR 71.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

D and D knowledge management information tool - a web based system developed to share D and D knowledge worldwide

Deactivation and decommissioning (D and D) work is a high risk and technically challenging enterprise within the U.S. Department of Energy complex. During the past three decades, the DOE's Office of Environmental Management has been in charge of carrying out one of the largest environmental restoration efforts in the world: the cleanup of the Manhattan Project legacy. In today's corporate world, worker experiences and knowledge that have developed over time represent a valuable corporate asset. The ever-dynamic workplace, coupled with an aging workforce, presents corporations with the ongoing challenge of preserving work-related experiences and knowledge for cross-generational knowledge transfer to the future workforce [5]. To prevent the D and D knowledge base and expertise from being lost over time, the DOE and the Applied Research Center at Florida International University (FIU) have developed the web-based Knowledge Management Information Tool (KM-IT) to capture and maintain this valuable information in a universally available and easily accessible and usable system. The D and D KM-IT was developed in collaboration with DOE Headquarters (HQ), the Energy Facility Contractors Group (EFCOG), and the ALARA [as low as reasonably achievable] Centers at Savannah River Sites to preserve the D and D information generated and collected by the D and D community. This is an open secured system that can be accessed from https://www.dndkm.org over the web and through mobile devices at https://m.dndkm.org. This knowledge system serves as a centralized repository and provides a common interface for D and D-related activities. It also improves efficiency by reducing the need to rediscover knowledge and promotes the reuse of existing knowledge. It is a community-driven system that facilitates the gathering, analyzing, storing, and sharing of knowledge and information within the D and D community. It assists the DOE D and D community in identifying potential solutions to their problem areas by using the vast resources and knowledge base available throughout the global D and D community. The D and D KM-IT offers a mechanism to the global D and D community for searching relevant D and D information and is focused on providing a single point of access into the collective knowledge base of the D and D community within and outside of the DOE. Collecting information from subject matter specialists, it builds a knowledge repository for future reference archiving Lessons Learned, Best Practices, ALARA reports, and other relevant documents and maintains a secured collaboration platform for the global D and D community to share knowledge. With the dynamic nature and evolution of the D and D knowledge base due to multiple factors such as changes in the workforce, new technologies and methodologies, economics, and regulations, the D and D KM-IT is being developed in a phased and modular fashion. (authors)

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

USACE FUSRAP Records Management Challenges in the Digital Age - 20242

In late 2016, the U.S. Army Corps of Engineers (USACE) implemented a Headquarters-supported nationwide records management initiative managed by the USACE-New England District to address the digitization and organization of Formerly Utilized Sites Remedial Action Program (FUSRAP) project files. This initiative will prepare USACE FUSRAP for moving toward a fully electronic system of records management. Additionally, the records management initiative streamlines the process and prepares the Administrative Record (AR) and Permanent Record (PR) file for transfer from USACE to the U.S. Department of Energy (DOE)-Office of Legacy Management (LM) following project closeout. Over the past couple of years, USACE and DOE-LM have been working to finalize a joint guidance document to assist in the FUSRAP records transition process. However, there remain key records management topics that require further discussion and development. Specifically, these topics include the implementation, process, and procedure for the capture of social media records and compliance with Section 508 of the Rehabilitation Act of 1973 (29 USC 794d), as amended. Due to advances in technology and the ease of communication it provides, USACE is using social media more often to interface with the public than 5, 10 or 15 years ago. The identification and capture of these records are challenging because it requires each USACE district or producer of a record to track what has been released through social media so that it can be captured as part of the AR or PR. USACE is currently evaluating processes and procedures for social media records capture. This paper explores how social media is being used by USACE across FUSRAP projects and best practices for the capture of social media records. Public accessibility to the project AR is required under the Comprehensive Environmental Response, Compensation, and Liability Act, which USACE FUSRAP must follow. As USACE FUSRAP ARs are converted to a digital format, greater accessibility to the files can be provided through web sites or other electronic media. However, Section 508 of the Rehabilitation Act must be considered when releasing electronic files to the public. Under Section 508, government agencies must ensure that information is made available to the public in a format that provides equal access to persons with disabilities. This paper discusses the electronic records requirements under Section 508 and addresses broad implementation of compliance across large digital records sets. Records management challenges are presented in the context of the FUSRAP records management program; however, these records management considerations and best practices are applicable across all Federal agency records management programs. Examples presented will further aid in understanding electronic records requirements that may be included on new USACE FUSRAP contracts as well as upcoming National Archives Records Administration requirements that will move us toward an electronic government. (authors)

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

96 Park Way Vicinity Property Remediation Challenges at the FUSRAP Maywood Superfund Site, Maywood, New Jersey - 19216

The Formerly Utilized Sites Remedial Action Program (FUSRAP) Maywood Superfund Site (FMSS), 'the Project,' was transferred to the US Army Corps of Engineers (USACE) in 1997. Shortly thereafter, USACE initiated FUSRAP remediation efforts on 24 commercial and residential properties and, in addition, the Project team evaluated properties previously addressed at the FMSS. One of those properties was located at 96 Park Way in Rochelle Park, New Jersey. The property history included removal of contaminated soil in 1985 by the US Department of Energy (US DOE). Soil sampling performed after removal of contaminated soil indicated the cleanup met applicable guidelines in place at the time. A Five-Year Review of the FMSS conducted by the US Environmental Protection Agency (USEPA) in 2009 found that 96 Park Way did not meet the current residential cleanup levels for the site and recommended additional study and remediation, if warranted. The current levels are established in the approved cleanup plan for the FMSS (known as the Record of Decision [ROD]) issued in 2003. Based on USEPA's recommendation, USACE carefully reviewed all available records for the property. The review considered the differences between the soil cleanup standards used in the 1980's and the 2003 ROD standards, which are slightly more restrictive for deeper soils. The USACE conducted detailed field studies at 96 Park Way in 2014-15 to address the USEPA findings. This work included gamma walkover surveys and soil sampling to depths of approximately 3.5 meters. The results identified three discrete areas at depth that did not meet the ROD residential soil cleanup standard: two areas bordering an adjacent highway (New Jersey Route 17) and one along railroad property adjacent to the site. To address these areas, soil removal actions were conducted in 2016-17. The principal challenge with construction, engineering, and community relations was the property housed two sensitive facilities: a long-term assisted care living facility and a short-term post-surgery rehabilitation center. The operations at these facilities could not be interrupted in any way during FUSRAP activities. The three FUSRAP-impacted areas were addressed and restored in coordination with the USEPA, property management, and local officials. Site restoration presented a particular challenge well after the remedial action was completed, as the property owner and the government disagreed on the level of restoration. Those negotiations are ongoing as of this writing. (authors)

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

Leveraging 3D CAD and Virtual Reality in Design of the Calcine Disposition Project - 20512

Fluor Idaho's Calcine Disposition Project (CDP) is leveraging the power of three-dimensional (3D) visualization, light detection and ranging (Lidar), and virtual reality to improve designs and worker safety. With the use of computer-aided design (CAD) modeling software, the CDP has created 3D models of the facility and systems to aid in the design, development, and operation of its retrieval and transfer system. This by itself is nothing new and has been done in industry for years. The CDP project however is taking this a few steps farther with the use of Lidar and virtual reality software. The CRP is tasked with removing a radioactive granular material from stainless steel bins located in concrete vaults that were constructed in the late 1950's and early 1960's. While construction drawings are available, it is not certain that the drawings are as-built or how accurate they truly are. Part of the project requires precise placement of equipment on to bins around multiple pipe, electrical lines, and bin stiffeners. Relying solely on the existing drawings and models created from these drawings is highly risky and prone to failure. The first step to determine the precise location and orientation of all obstructions in the vault the project will use a Lidar system to scan the vault structure and precisely locate all vault and bin components. Once this information is obtained, the data will be integrated into the CAD model. The model will then be verified, and a true as-built model developed. Precise component location, size, and orientation will be used for final design and placement of retrieval components. The second step will be to take the verified 3D model and scan data and import it into virtual reality software creating a virtual world. While some work on the project requires all work to be completed remotely due to high radiation fields, some areas have low enough radiation levels that personnel can enter and perform work. With the use of a virtual reality headset, operators and technicians will be able to enter our virtual world and become familiar with the surroundings and perform work prior to ever entering the radiation environment. This training is expected to pay dividends by improving worker efficiency, reducing errors, and improving confidence that the work can be performed as expected. It will play a key aspect in as low as reasonably achievable (ALARA) principles. An additional aspect in using the Lidar scans and virtual environment, is that it will give engineers and managers an opportunity to revisit the facility, especially radiation areas, at any time to obtain design information and measurements that may be costly or impossible to obtain once the facility transitions into radiation operations. (authors)

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

Transfer and Transition: Interagency Coordination for Managing Public Lands at UMTRCA Title II Sites in Wyoming - 16614

By the end of fiscal year 2025, the U.S. Department of Energy (DOE) Office of Legacy Management (LM) is anticipating adding 17 sites remediated under Title II of the Uranium Mill Tailings Radiation Control Act (UMTRCA) to the current inventory of 90 sites that it manages. Among the new sites are ones where federal public lands occur within the proposed long-term care boundary, the boundary determined by the Nuclear Regulatory Commission and LM as necessary to maintain site protectiveness for the entombed uranium mill tailings and residual groundwater contamination. For these sites, public land withdrawals for land and minerals will need to be established. LM's primary mission at UMTRCA sites is to protect the public and the environment from exposure to contamination at the sites. For the sites with public lands or federally controlled minerals that will be transferring to LM, the Office will apply to the Department of the Interior (DOI) Bureau of Land Management (BLM) for new, public land and mineral withdrawals. At most current LM UMTRCA sites that involved public lands and minerals, DOI granted DOE 'full administrative jurisdiction' and permanent withdrawals. Hence, these withdrawals are, permanently, no longer subject to public land, mining, and mineral-leasing laws and regulations. LM is coordinating with DOI/BLM in Wyoming to permanently withdraw full and partial jurisdiction at future UMTRCA Title II sites in that state. This approach would allow LM to fully administer surface lands and minerals, where necessary, and DOI and LM to administer surface lands and leasable minerals where it would not jeopardize sites' radiological safety and long-term public and environmental protection. This 'shared-jurisdiction approach' will meet LM's strategic goal of protecting human health and the environment but also allow BLM to fulfill their mission to 'manage and conserve the lands under the mandate of multiple-use and sustained yield'. In addition, LM could also fulfill the fourth goal of its Strategic Plan, to optimize land use and assets. The portions of the sites where LM will ask for full jurisdiction are those that contain the disposal cell and any adjacent lands and minerals required to maintain and protect it. It is anticipated that the shared jurisdiction portion of the withdrawal would largely coincide with land where known or projected contaminated groundwater from the past ore processing exists or will migrate over time. Examples of BLM-authorized activities on the shared jurisdiction portions of the withdrawals may include livestock grazing, wildlife habitat improvement projects, recreation, and rights of ways for roads, pipelines, utilities, and communication sites. In addition, BLM could authorize subsurface uses such as oil and gas, coal, or geothermal development or groundwater pumping from deeper aquifers if the proposal would not disturb groundwater contamination. For areas of the withdrawal where the agencies will share jurisdictional authority, DOE LM and BLM will develop an interagency agreement (IA) to define acceptable uses and associated roles and responsibilities. To ensure maximum protection of human health and the environment, each site's IA will include a provision allowing DOE LM the right of first refusal against any action it believes might affect radiological safety at the site. (authors)

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

Consequence Analysis of Residual Water in a Storage Canister (Preliminary Report)

The objective of this work is to examine the consequences of a residual water on change in condition of cladding and SNF during dry storage. Small amounts of residual water may remain in SNF dry storage canisters after the fuel assemblies are transferred to the canister in the spent fuel pool (SFP), the canister is removed from the SFP and drained or pumped, vacuum dried in several pressure reducing steps, and backfilled with helium. This residual water from incomplete drying could degrade fuel rod cladding, oxidize exposed fuel pellets, and cause cladding failure resulting in SNF fines’ release in canister cavity. If residual water content is high enough, it could corrode internal components inside the canister, e.g., fuel basket, and neutron absorber plates, but CNWRA work indicated that corrosion of the internal component is expected to be negligible even with 55 moles of residual water. Considering this, focus of this work to determine partitioning of residual water towards fuel and cladding oxidation. Residual water will decompose by radiolysis and create reactive oxygen species, which, at sufficient high temperatures will oxidize cladding, and UO 2 exposed by clad breaches. If a significant fraction of UO 2 in a fuel rod is oxidized to U 3 O 8 , swelling of the fuel pellets will rupture the cladding and release fuel particles, contaminating the inside of the canister and complicating retrieval of fuel assemblies, transport and ultimate disposal. The report describes an evaluation of the potential cladding and exposed SNF pellets due to residual water. An integrated quantitative approach was used to estimate the effects. The approach consisted of (i) the evolution of environmental conditions, i.e., temperature and radiation field, in the cask, (ii) the physicochemical processes that affect degradation of the materials within the cask, and (iii) the potential damage states that may result from these conditions and processes. The assessment considers the following set of cask environmental conditions: (i) the quantity of residual water after drying, (ii) internal temperature fields and their intensity decrease over time, and (iii) the strength of the internal radiation field. The quantity of residual water will affect the mass of cladding, SNF, and other components that are degraded by oxidation. Temperature controls the rate of the oxidation reactions, the phase composition of water, and, to a lesser extent, the rate at which water vapor is radiolyzed. The strength of the radiation field controls the radiolysis rate of residual water. These processes, i.e., time evolution of temperature fields and their effects on radiolysis of residual water, fuel and cladding oxidations, were modeled using the integrated approach.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Stabilization of Residual Contamination with Alternative Materials - 19404

Since 2015, Savannah River National Laboratory (SRNL) in conjunction with Florida International University (FIU) have been investigating the applicability of intumescent coatings as radiological contamination fixatives for stabilization of residual materials after decontamination efforts have concluded. Intumescent coatings are designed to prevent heat transfer to a protected substrate and are typically used in both the aerospace industry to protect components from thermal shock and in building design to protect building joints that may fail in the event of a fire. For this work, these materials are examined for their ability to reduce worker risk by fixating residual contamination and lower the impact of a release in a scenario that breaches containment of a contaminated building and subsequently induces a large-scale fire by crediting their fire-retardant attributes. Initial studies of intumescent coatings and commercial fixatives included performance in direct flame, varied environmental conditions (20-110 F or -6.67 C, 10-95% relative humidity), adhesion studies to stainless steel after curing in varied environmental conditions, and radiological exposure. Of all the materials tested, many were found to be inadequate in one or more of these areas. These materials failed in one or more of three characteristic ways: 1) through delamination at reduced temperatures (≤40 F or 4.44 C) or from even moderate handling, 2) melting off the substrate at elevated temperatures (≥90 F or 32.22 C), 3) proving to be flammable and promote smoke propagation. Failure in these ways results in the potential for a subsequent release of the fixated contamination in the event of water presence/absence or direct flame. Using these results as a platform for the best in class identified material that did not experience failure in any of these tests, SRNL and FIU targeted the plutonium-238 contaminated hot cells within Savannah River Site (SRS) Building 235-F as a characteristic test bed to benchmark an intumescent coating used as a contaminant fixative. Through discussion with the SRS 235-F Risk Reduction personnel, appropriate hazards and performance metrics were discussed and a cold demonstration was performed in a hot cell mockup located at SRS. The application methods were proven functional in the cold mock up, and these intumescent coatings were subsequently utilized in the Savannah River Site Building 235-F's Plutonium Fuel Form (PuFF) Facility to aid in stabilization of plutonium-238 in both a contaminated hot cell and Entry Hood in September 2018. Presented in this work is the overview of the application process, associated working considerations for introduction of equipment and materials into the facility, lessons learned, and lifetime monitoring data to date. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Used Fuel Dry Storage Systems and Components for Shutdown Plants to Expedite Defueling of Spent Fuel Pool - 18378

For plants that have announced plans to shut down, their main priority is to off-load the Used Nuclear Fuel from the spent fuel pool to dry storage at an ISFSI so that they can proceed to 'Safe store' or other storage options, and start their decommissioning process as soon as possible. One of the major expenses for a decommissioned plant is the cost of maintenance of the spent fuel pool. Therefore, the ability of the dry storage system to store short cooled fuel is an important aspect of dry storage system for shutdown plants. When a shutdown plant evaluates the dry storage options, total cost of the transferring the fuel from the pool to ISFSI site needs to be evaluated. One of the major costs to the utility for dry storage is the cost of ISFSI pad (basemat) along with the cost of dry storage system hardware. Smaller ISFSI footprint saves utility for the cost of ISFSI site excavation, cost of engineered backfill and cost of ISFSI pad. In order to minimize the cost of dry storage systems, besides the size of the ISFSI pad as described above, other important considerations is to use high capacity systems. The highest capacity systems available in the dry storage market today are with 37 PWR or 89 BWR used fuel assemblies per canister. These dry storage systems are expected to be at the plant sites for extended duration due to the unavailability of Interim Storage Facilities or a permanent disposal sites like Yucca Mountain. Maintaining the integrity of the dry storage systems canisters and storage overpack so that they can safely protect the fuel and integrity of used nuclear fuel is maintained during storage period is of utmost importance. Robust Aging Management Program is an important aspect for dry storage systems not only at the shutdown plants but also at the operating plants. This paper will provide the update of a reduced footprint version of TN Americas NUHOMS{sup R} System with MATRIX HSM and NUHOMS{sup R} EOS Canisters that addresses the need for maximizing safety, ISFSI capacity while optimizing cost. It will provide an update on the current capabilities, design features and current status of the licensing and deployment of NUHOMS{sup R} EOS System. It will provide the current status of the MATRIX HSM along with the significant enhancement included for a robust Canister Evolved Safety Program for Aging Management. It will also describe the construction sequence; loading operations, and inspections during extended storage as a part of aging management program. The compact version of TN Americas NUHOMS{sup R} System with MATRIX HSM offers a 'Universal' Storage Overpack which is designed to accept all different designs dry storage canisters available in the industry. As UNF storage continues to drives the industry towards innovations that address the need for optimized dry storage systems, compact ISFSIs and prudent aging management programs, this new evolution of the TN Americas NUHOMS{sup R} system with MATRIX HSM and EOS Canisters offers a significant leap forward in addressing all of these challenges. (authors)

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DOE OSTI2013

Decommissioning an Active Historical Reactor Facility at the Savannah River Site - 13453

The Savannah River Site (SRS) is an 802 square-kilometer United States Department of Energy (US DOE) nuclear facility located along the Savannah River near Aiken, South Carolina, where Management and Operations are performed by Savannah River Nuclear Solutions (SRNS). In 2004, DOE recognized SRS as structure within the Cold War Historic District of national, state and local significance composed of the first generation of facilities constructed and operated from 1950 through 1989 to produce plutonium and tritium for our nation's defense. DOE agreed to manage the SRS 105-C Reactor Facility as a potentially historic property due to its significance in supporting the U.S. Cold War Mission and for potential for future interpretation. This reactor has five primary areas within it, including a Disassembly Basin (DB) that received irradiated materials from the reactor, cooled them and prepared the components for loading and transport to a Separation Canyon for processing. The 6,317 square meter area was divided into numerous work/storage areas. The walls between the individual basin compartments have narrow vertical openings called 'slots' that permit the transfer of material from one section to another. Data indicated there was over 830 curies of radioactivity associated with the basin sediments and approximately 9.1 M liters of contaminated water, not including a large quantity of activated reactor equipment, scrap metal, and debris on the basin floor. The need for an action was identified in 2010 to reduce risks to personnel in the facility and to eliminate the possible release of contaminants into the environment. The release of DB water could potentially migrate to the aquifer and contaminate groundwater. DOE, its regulators [U. S. Environmental Protection Agency (USEPA)-Region 4 and the South Carolina Department of Health and Environmental Control (SCDHEC)] and the SC Historical Preservation Office (SHPO) agreed/concurred to perform a non-time critical removal action for the In Situ Decommissioning (ISD) of the 105-C Disassembly Basin. ISD consisted of stabilization/isolation of remaining contaminated water, sediment, activated reactor equipment, and scrap metal by filling the DB with underwater non-structural grout to the appropriate (-4.877 meter) grade-level, thence with dry area non-structural grout to the final -10 centimeter level. The roof over the DB was preserved due to its potential historical significance and to prevent the infiltration of precipitation. Forced evaporation was the form of treatment implemented to remove the approximately 9.1 M liters of contaminated basin water. Using specially formulated grouts, irradiated materials and sediment were treated by solidification/isolation thus reducing their mobility, reducing radiation exposure and creating an engineered barrier thereby preventing access to the contaminants. Grouting provided a low permeability barrier to minimize any potential transport of contaminants to the aquifer. Efforts were made to preserve the historical significance of the Reactor in accordance with the National Historic Preservation Act. ISD provides a cost effective means to isolate and contain residual radioactivity from past nuclear operations allowing natural radioactive decay to reduce hazards to manageable levels. This method limits release of radiological contamination to the environment, minimizes radiation exposure to workers, prevents human/animal access to the hazardous substances, and allows for ongoing monitoring of the decommissioned facility. Field construction was initiated in August 2011; evaporator operations commenced January 2012 and ended July 2012 with over 9 M liters of water treated/removed. Over 8,525 cubic meters of grout were placed, completing in August 2012. The project completed with an excellent safety record, on schedule and under budget. (authors)

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DOE OSTI2019

Pneumatic System for the Neutron Generator Facility at SRNL

A neutron generator system equipped with a pneumatic sample transfer system, enables scientists to perform neutron activation analysis with increased efficiency and accuracy. The samples are placed in sealed containers called 'rabbits' and are passed through the system. Once sealed in the rabbit, the samples are loaded into the tubes via a hopper mechanism. The sample is launched into the generator with compressed air and captures neutrons, causing the sample to become radioactive. Next, the sample is sent to a detector, where the samples are counted for a set length of time on a Ge detector. Once the detection is completed, the sample is then sent to the trash. The system utilizes an interface to select which generator valve and detector location to send the sample too. This project is supports the future development of the neutron generator in the F wing of 773-A. A program is used to automate the passing of the rabbit through the system. The user is able to select the generator valve and detector to be utilized in the testing of the sample. Using the 'Add Rabbit' button the user can add as many sample rabbits to rabbit editor tab. Characteristics are then entered into the fields for each rabbit. The Pre-Det Cooldown and Post-Det Cooldown are intervals used to space out the time before the rabbit is sent into and out of the detector. After all characteristics are entered, the 'Submit to Hopper' button is selected thus preparing the program to run the rabbits in the inserted order. Small setup of the system using the available parts. Simulates the movement of the rabbit from the generator to the detector. Small valves are located before each manifold. To transport the rabbit from one side to the other, the valve of the sending side must be open and the receiving side must be closed. The manifold and valve in direct line with the compressor would extend to the hopper. Once the hopper is loaded, the valve would be open to send the rabbit to generator. Additional manifolds and other mechanical parts are necessary to complete the system design. To complete the construction of the system additional parts, such as the mechanical parts for the switching mechanism and the solenoid valves, are necessary. Further, to complete the automatization of the system, an Arduino is required to communicate with the valves and adjust the linear servos appropriately so that the rabbits are sent to the selected generator location and detector. This work is aimed to be completed in the next fiscal year.

42 ENGINEERING↗
DOE OSTI2025

Selective electrified polyethylene upcycling by pore-modulated pyrolysis

Plastic waste is a increasing problem, accumulating in landfills and the environment. Pyrolysis is a promising and industrially relevant approach for transforming plastic waste into value-added chemicals. However, the selectivity and yield of traditional plastic pyrolysis are poor, with products featuring broad molar mass distributions. Here we report a highly selective, energy-efficient and catalyst-free pyrolysis method that can upcycle plastic into value-added chemicals via pore-modulated pyrolysis. Using a Joule-heated carbon column, we demonstrate the pivotal role of the reactor’s graded porous structure in decreasing the polydispersity of the reaction intermediates, enabling high product selectivity and yield. The decreasing pore size of the reactor modulates the mass transport in an apparent gating effect—preventing high-molar-mass species from exiting the reactor before sufficient pyrolysis has occurred. Using polyethylene as a model reactant, we demonstrate a high yield of 65.9 ± 5.2% and up to 80.8% selectivity toward value-added aviation fuel precursor (C8–C18 hydrocarbons) without the use of any catalysts.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2015

The SRNL technical assistance program: transferring lessons learned from two decades of remedial successes to the DOE complex - 15438

The Savannah River Site has an established history of developing implementable risk reduction strategies, developing and deploying innovative technological solutions, and collaboratively achieving realistic and protective environmental end states for complex environmental problems. Part of this success results from the strong historical collaboration between site environmental restoration groups and the Savannah River National Laboratory which results in the promulgation of innovative technical approaches and technologies from the laboratory to ongoing remediation campaigns. Much of the underlying research and development effort was funded by DOE's Office on Environmental Management through a series of applied technology demonstration projects (Integrated Demonstration Project, Integrated Programs and the Applied Field Research Site). Concomitant with the research efforts, SRNL has managed a dynamic and efficient national program that provides teams of technical experts with a broad experience base to recommend strategies that reduce risk and/or technical uncertainty for challenging environmental problems at other DOE facilities. At many of these sites, standard remedial approaches (e.g., excavation, pump and treat, soil vapor extraction) have either failed or proved to be too costly, inadequate, or ineffective. Alternatively, other sites struggle with implementation of innovative but less well understood approaches. The technical assistance teams use a structured process that based on several basic concepts, specifically, development of site specific technical frameworks, and careful matching of remedies to site-specific chemical, geological and physical conditions. Development of technical frameworks is a key strategy to apply basic science to an applied field problem. When directed toward understanding complex real-world environmental remediation challenges, frameworks are tools that support practical identification and incorporation of the key-controlling scientific processes and principles. Frameworks can also be used to minimize technical risks, encourage efficiency and effectiveness, and provide the basis for innovative and creative solutions. Efficient and effective environmental cleanup also requires matching the character of remediation and stabilization methods to the nature of the target zone of contamination as the nature of the target zone evolves through the life of the remedial project. A contaminant plume can be divided into the following zones: the source zone, the impact zone, and the transitional zone that then can be used to identify classes of technologies that are appropriate to that zone. For example, physical and chemical methods (e.g., trapping, immobilization, destruction, or isolation) that directly address the source contaminants are often appropriate for the disturbed zone during the remedial process. A variety of methods that include both active treatments (e.g., pump and treat or active bioremediation) and enhanced attenuation technologies (e.g., geochemical manipulation or reactive barriers) are often suitable for the primary contamination zone or impact zone. Various strategies based on natural attenuation processes may be applicable to the primary contamination zone and these methods are typically applied for the transition or baseline portions of the plume. Careful matching of remedies to site-specific conditions is critical to long-term success in environmental cleanup and restoration. The matching process facilitates selection of technologies with particular strengths that align with real-world needs and constraints, encourages strategic use of multiple or combined technologies to address major plume sub-domains, and supports transitioning technologies in space and time as remediation progresses. Since 2006, the SRNL Technical Assistance program has focused on providing support across the DOE complex. During this time, over 25 teams have visited eleven DOE sites and made recommendations that yielded an estimated cost savings of $100 M. Examples will be provided that illustrate key aspects and significant successes of the technical assistance program. (authors)

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