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All sourcesNASA NTRSDOE OSTI
NASA NTRS2008

Molten Materials Transfer and Handling on the Lunar Surface

Electrolytic reduction processes as a means to provide pure elements for lunar resource utilization have many advantages. Such processes have. the potential of removing all the oxygen from the lunar soil for use in life support and for propellant. Electrochemical reduction also provides a direct path for the. production of pure metals and silicon which can be utilized for in situ manufacturing and power production. Some of the challenges encountered in the electrolytic reduction processes include the feeding of the electrolytic cell (the transfer of electrolyte containing lunar soil), the withdrawal of reactants and refined products such as the liquidiron~siliconalloy with a number of impurities, and the spent regolith slag, produced in the hot electrolytic cell for the reduction of lunar regolith. The paper will discuss some of the possible solutions to the challenges of handling molten materials on the lunar surface, as well as the path toward the construction and testing of a proof-of-concept facility.

Stefanescu, Doru M.↗
NASA NTRS2008

Status Update Report for the Peregrine 100km Sounding Rocket Project

The Peregrine Sounding Rocket Program is a joint basic research program of NASA Ames Research Center, NASA Wallops, Stanford University and the Space Propulsion Group, Inc. (SPG). The goal is to determine the applicability of liquifying hybrid technology to a small launch system. The approach is to design, build, test and y a stable, efficient liquefying fuel hybrid rocket vehicle to an altitude of 100 km. The program was kicked o in October of 2006 and has seen considerable progress in the subsequent 18 months. Two virtually identical vehicles will be constructed and own out of the NASA Sounding Rocket Facility at Wallops Island. This paper presents the current status of the project as of June 2008. For background on the project, the reader is referred to last year's paper.

Dyer, Jonny↗
NASA NTRS2014

Injury Surveillance Among NASA Astronauts Using the Barell Injury Diagnosis Matrix

Astronauts perform physically demanding tasks and risk incurring musculoskeletal injuries during both groundbased training and missions. Increased injury rates throughout the history of the U.S. space program have been attributed to numerous factors, including an aging astronaut corps, increased Weightless Environment Training Facility (WETF) and Neutral Buoyancy Laboratory (NBL) training to construct the International Space Station, and improved clinical operations that promote injury prevention and reporting. With NASA program changes through the years (including retirement of the Shuttle program) and an improved training environment (including a new astronaut gym), there is no surveillance program to systematically track injury rates. A limited number of research projects have been conducted over the past 20 years to evaluate musculoskeletal injuries: (1) to evaluate orthopedic injuries from 1987 to 1995, (2) to describe upper extremity injuries, (3) to evaluate EVA spacesuit training related injuries, and (4) to evaluate in-flight musculoskeletal injuries. Nevertheless, there has been no consistently performed comprehensive assessment of musculoskeletal injuries among astronauts. The Barell Injury Diagnosis Matrix was introduced at the 2001 meeting of the International Collaborative Effort (ICE) on Injury Statistics. The Matrix proposes a standardized method of classifying body region by nature of injury. Diagnoses are coded using the International Classification of Diseases, Ninth Revision, Clinical Modification (ICD-9-CM) coding system. The purpose of this study is to assess the usefulness and complexity of the Barell Injury Diagnosis Matrix to classify and track musculoskeletal injuries among NASA astronauts.

Murray, J. D.↗
NASA NTRS2015

Layered Thermal Insulation Systems for Industrial and Commercial Applications

From the high performance arena of cryogenic equipment, several different layered thermal insulation systems have been developed for industrial and commercial applications. In addition to the proven areas in cold-work applications for piping and tanks, the new Layered Composite Insulation for Extreme Environments (LCX) has potential for broader industrial use as well as for commercial applications. The LCX technology provides a unique combination of thermal, mechanical, and weathering performance capability that is both cost-effective and enabling. Industry applications may include, for example, liquid nitrogen (LN2) systems for food processing, liquefied natural gas (LNG) systems for transportation or power, and chilled water cooling facilities. Example commercial applications may include commercial residential building construction, hot water piping, HVAC systems, refrigerated trucks, cold chain shipping containers, and a various consumer products. The LCX system is highly tailorable to the end-use application and can be pre-fabricated or field assembled as needed. Product forms of LCX include rigid sheets, semi-flexible sheets, cylindrical clam-shells, removable covers, or flexible strips for wrapping. With increasing system control and reliability requirements as well as demands for higher energy efficiencies, thermal insulation in harsh environments is a growing challenge. The LCX technology grew out of solving problems in the insulation of mechanically complex cryogenic systems that must operate in outdoor, humid conditions. Insulation for cold work includes equipment for everything from liquid helium to chilled water. And in the middle are systems for LNG, LN2, liquid oxygen (LO2), liquid hydrogen (LH2) that must operate in the ambient environment. Different LCX systems have been demonstrated for sub-ambient conditions but are capable of moderately high temperature applications as well.

thermal conductivity↗
DOE OSTI2023

Evaluating and Countering the Insider Threat to the Radioactive Source Supply Chain

The modern supply chain is a global enterprise and little drove this home more than the global COVID-19 pandemic which sent economic shockwaves throughout the world. Many goods became scarce, as products were delayed, in limited supply, or simply not available. The suddenly diminished supply collided with still high demand and led to greatly increased costs. This was particularly true for the radioactive source supply chain. The pandemic introduced extensive delays for construction projects, slowed the transport of radiological materials to facilities, interrupted treatment deliveries, and impaired the mobility of contractors across the industry. All of these concerns not only adversely affected the economy, but also impacted the safety and security of radiological material, potentially raising national security concerns. The vulnerability of the supply chain, a critical element in an increasingly interconnected world, was exposed. One example that challenged the adaptive capacity of the overall supply chain is the Ever Given container ship, which became stuck in the Suez Canal in 2020. This accident immediately shut down shipments that accounted for 12% of global trade, with long-term impacts estimated to be much larger. Developing the ability to anticipate and react in real-time to sudden changes has quickly become a necessity, particularly in industries that deal with the transport of hazardous material. The reaction to these dramatic incidents was to largely focus attention and resources on protecting the supply chain from external threats. However, the threat to the radioactive material supply chain from insiders intimately involved in the process may be even greater and remains a blind spot that requires increased attention. Recent events revealed the blueprint for targeting and disrupting that supply chain, so the potential for a malicious insider—or a manipulated, unwitting insider—to take advantage of this vulnerability is elevated, creating security concerns for radiological industries. This paper examines and analyzes the potential insider threat to the radioactive source supply chain and recommends steps to take to counter this possibility.

Kinney, Justin↗
DOE OSTI2023

SOFE2023 Presentation: "Tritium Safety and Sensitivity Analysis of Tritium Extraction eXperiment (TEX) PbLi Loop Using MELCOR-TMAP"

The efficient extraction of tritium from the breeder blanket (BB) in a fusion reactor is vital for maintaining the tritium economy for a sustainable and self-sufficient fusion power plant. The vacuum permeator (VP) design is one of several promising technology platforms for extracting tritium from the BB with the advantages of being capable of operating continuously at high temperatures with the aim of achieving high tritium extraction efficiencies. The Tritium Extraction eXperiment (TEX), under construction at the Safety and Tritium Applied Research (STAR) Facility at Idaho National Laboratory (INL), is designed to test the viability of VP technology by introducing atomic hydrogen (H, D, and T) from a source permeator (SP) into a forced convection lead lithium eutectic (PbLi) loop and determining the hydrogen extraction rates from the vanadium VP. Presented here is a computational analysis of the hydrogen inventory and permeation behavior in TEX using the hydrogen transport and accident analysis code MELCOR-TMAP developed at INL. The results of a sensitivity analysis based on MELCOR-TMAP simulations of hydrogen extraction efficiencies from the TEX test section using varying PbLi flow rates, hydrogen source rates from the SP, and material property uncertainties are also reported. Operation scenarios for TEX in consideration of the analyses based on results from MELCOR-TMAP are discussed.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2020

Overview of Upcoming Process Improvement Efforts at the Saltstone Facilities at the Savannah River Site - 20474

The Saltstone Facilities at the Savannah River Site (SRS) process and dispose of low activity salt solution waste generated by other liquid waste facilities. Today in the Liquid Waste System (LWS), approximately 125 million liters of salt waste remain to be processed. The Saltstone Production Facility (SPF) receives the decontaminated salt waste, which is mixed with dry feeds consisting of cement, fly ash and slag to produce the saltstone. The mixed slurry is pumped to the Saltstone Disposal Units (SDUs) for safe disposal in the Saltstone Disposal Facility (SDF). To support continuous Salt Waste Processing Facility (SWPF) operations at annual processing volumes many times higher than present, significant modifications are required to ensure that the SPF has the capacity to support the throughput of salt waste. These changes include several infrastructure upgrades that are being worked through the Enhanced Low Activity Waste Disposal (ELAWD) project. In addition to infrastructure upgrades, Savannah River Remediation (SRR), the Savannah River Site's Liquid Waste Contractor for the U.S. Department of Energy, has identified a flowsheet improvement opportunity of implementing a cement-free saltstone formulation to eliminate cement from the current three component formulation. This two-component formulation would simplify the procurement, transport, off-loading, and storage of the premix materials, and increase the storage capacity for the individual components. This flowsheet, when implemented, will reduce operational risk and improve throughput necessary to achieve disposal rates necessary for SWPF operations. Research has been completed to confirm the feasibility of this formulation, with the new cement-free flowsheet expected to be implemented in the next few years. Improvements have also been identified downstream of the SPF to increase SDU utilization. Specifically, a concentrated effort has taken place to raise the fill height in SDU 6 to the full disposal height, 13 meters, to maximize the usage of each SDU in support of the liquid waste disposal mission. This improvement has been made through additional thermal modeling and improved process understanding to create a SDU flammability model and control strategy more representative of facility operations. This work will allow for the completion of construction and operational readiness for SDU 7 while filling SDU 6 to ensure there is no lapse in available space once SWPF starts up. A summary of these improvements as they relate to increased throughput will be provided along with associated status. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Digital Safety Analysis for Small Modular Nuclear Reactors (SMRs)

A Documented Safety Analysis (DSA) is a Department of Energy (DOE) construct that defines the extent to which a nuclear facility can be operated safely. It includes a description of hazards, safe boundaries, and hazard controls. The authors assert that a Digital Safety Analysis (DgSA) is far superior to a legacy DSA for several reasons: • The underling database is structured such that it is possible to perform a comprehensive design review and safety analysis by iterating systematically across a hierarchy of linked objects versus a redundant and spotty review by entities of various abilities under unknown resource and schedule constraints. • The analysis of a new design can discover elements that are similar to elements in previous designs. The discovery of similarities is made possible by using the same structure for the underlying database for each new DgSA. The “prior learning” from previous designs is then applied automatically to new designs. • Outputs from the DgSA are from a single source to ensure consistency among various views of the same information. After the DgSA is released, the continued use of a single source implements a configuration management program to ensure consistency between the design basis, the design, the built system, and system procedures. • The development of the DgSA is agile in that any change in a linked object triggers an analysis of impacts on other linked objects and updates of linked objects are made accordingly. After the DgSA is released, the continued maintenance of these links and objects automates the “unreviewed safety question” process.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Environmental, Health and Safety Assessment Application of a Heat Integrated, Post-combustion CO2 Capture System Using Monoethanolamine and Hitachi H3-1 Solvent at the E.W. Brown Coal-fired Power Plant

An Environmental Health and Safety (EH&S) Assessment was completed for operation of a pilot scale carbon dioxide (CO2) capture system (CCS) at the Louisville Gas & Electric (LG&E) and Kentucky Utilities (KU) Energy, LLC (LG&E-KU) E. W. Brown Generating Station in Harrodsburg, Mercer County, Kentucky. The pilot system was evaluated using two CO2 capture solvents, monoethanolamine (MEA) and Hitachi’s H3-1 solvent. The purpose of the EH&S Assessment was to determine if there were any unacceptable environmental, health or safety concerns that may prevent operation or environmental permitting of a larger scale pilot scale plant. The assessment included review of process flow diagrams, input and output flow rates for primary materials, literature describing similar processes and chemicals used or created during the process, actual air emission and wastewater test data and air emission calculations. The evaluation included identification of risks related to hazardous chemicals, air emissions, wastewater discharges, solid wastes generated and employee hazards. Most potential EH&S issues that were identified are commonly found and successfully managed at large industrial facilities. No risks were identified that could not be successfully managed or would likely prevent implementation or environmental permitting of the pilot or larger scale plant. Potential exposures and resulting health risks from exposure to low concentrations of nitrosamines generated from degradation of the Hitachi H3-1 and possibly MEA solvents may be a health concern for site workers, but additional investigation is needed to accurately assess this risk. Applying the information gained from operating the pilot plant, performing air emission testing and used solvent testing to a large-scale facility at a commercial coal-fired electric generating plant suggests additional air emission control measures may be necessary to reduce potential site worker health risks and possibly community risks. Depending upon the location and regional air quality, considerable effort may be required to obtain the necessary air permit for construction and operation of a similarly configured large-scale PCCCS facility.

gas separation↗
DOE OSTI2020

Environmental, Health and Safety Assessment Application of a Heat Integrated, Post-combustion CO 2 Capture System Using Monoethanolamine and Hitachi H3-1 Solvent at the E.W. Brown Coal-Fired Power Plant

An Environmental Health and Safety (EH&S) Assessment was completed for operation of a pilot scale carbon dioxide (CO 2 ) capture system (CCS) at the Louisville Gas & Electric (LG&E) and Kentucky Utilities (KU) Energy, LLC (LG&E-KU) E. W. Brown Generating Station in Harrodsburg, Mercer County, Kentucky. The pilot system was evaluated using two CO 2 capture solvents, monoethanolamine (MEA) and Hitachi’s H3-1 solvent. The purpose of the EH&S Assessment was to determine if there were any unacceptable environmental, health or safety concerns that may prevent operation or environmental permitting of a larger scale pilot scale plant. The assessment included review of process flow diagrams, input and output flow rates for primary materials, literature describing similar processes and chemicals used or created during the process, actual air emission and wastewater test data and air emission calculations. The evaluation included identification of risks related to hazardous chemicals, air emissions, wastewater discharges, solid wastes generated and employee hazards. Most potential EH&S issues that were identified are commonly found and successfully managed at large industrial facilities. No risks were identified that could not be successfully managed or would likely prevent implementation or environmental permitting of the pilot or larger scale plant. Potential exposures and resulting health risks from exposure to low concentrations of nitrosamines generated from degradation of the Hitachi H3-1 and possibly MEA solvents may be a health concern for site workers, but additional investigation is needed to accurately assess this risk. Applying the information gained from operating the pilot plant, performing air emission testing and used solvent testing to a large-scale facility at a commercial coal-fired electric generating plant suggests additional air emission control measures may be necessary to reduce potential site worker health risks and possibly community risks. Depending upon the location and regional air quality, considerable effort may be required to obtain the necessary air permit for construction and operation of a similarly configured large-scale PCCCS facility.

gas separation↗
NASA NTRS1993

Validation of a nonintrusive optical technique for the measurement of liquid mass distribution in a two-phase spray

This work is the continuation of an optical technique development program initiated as part of the 1992 Summer Faculty Fellowship Program. The 1992 work consisted of the formulation and implementation of a technique ivolving the spatial deconvolution of fluorescence data from a uniformly illuminated, seeded densed spray to obtain quantitative measurements of the liquid density profiles. This measurement approach largely overcomes substantial scattering problems associated with other optical approaches for two-phase flows. However, to apply this measurement approach with confidence to unknown flows, the technique must be validated. Consequently, technique validation using classical grid patternator techniques has been focus of the current work. This work has included the design and construction of a patternator rig and the implementation of a test program designed for the comparison of patternator data with the deconvolved optical data. The flow field used for the validation is the plume of an axisymetric swirl coaxial LOX injector being considered for use in the Space Transportation System Main Engine. The flow facility is an improved version of the test rig which was constructed in 1992 for the initial technique development. This report includes a brief description of the optical measurement technique and the patternator rig and a presentation of the data comparisons.

Hartfield, Roy, Jr.↗
DOE OSTI2020

A Dose Assessment Model for Radioactive Waste Exposed by Gully Erosion at West Valley - 20513

The Western New York Nuclear Service Center (WNYNSC), located approximately 48 km south of Buffalo, New York, is the site of a former nuclear fuel reprocessing and radioactive waste disposal facility. Spent nuclear fuel was processed there from 1966 to 1972, leaving behind radioactive and chemical wastes in two disposal areas and a waste tank farm. Site operations also resulted in releases of radioactivity to site soils, groundwater, and to surface waters draining the site. The New York State Energy Research and Development Authority (NYSERDA) and the U.S. Department of Energy (DOE) are collaborating in a process of decision making for decommissioning those facilities remaining at the WNYNSC following the completion of Phase 1 decommissioning. Neptune and Company, Inc. (Neptune) was contracted to develop a probabilistic performance assessment (PPA) computer model to assist the agencies in this process. The PPA Model includes a contaminant transport component focusing on the movement of contaminants within and among environmental media including groundwater and surface water transport, contaminant translocation by plants and animals, diffusion, and erosion. The model also includes evaluation of potential exposure and health effects for a Resident Farmer exposure scenario, where the Resident Farmer represents a critical group, described as that group of individuals reasonably expected to receive the greatest exposure to residual radioactivity for any applicable set of circumstances. The West Valley Site is located in the glaciated Allegheny Plateau region of western New York State. The waste reprocessing and disposal areas were constructed on a relatively fat area of plateau dissected by drainages of Buttermilk Creek, including Erdman Brook, Franks Creek, and Quarry Creek. An important aspect of the contaminant transport component of the PPA model is consideration of erosive processes such as slumping of the stream slopes and the advance of gullies from these streams. These erosion processes remove material from the plateau, growing the size of the creek valleys and making them wider and deeper. Of particular interest for the impact of erosion is radioactive waste contained in the Nuclear Regulatory Commission (NRC)-Licensed Disposal Area (NDA), the New York State-Licensed Disposal Area (SDA), and residual radiological inventory in the underground storage tanks at the Waste Tank Farm (WTF). The PPA Model is organized around geographically-defined facilities which were constructed upon the plateau, including the NDA, SDA, and WTF. Ongoing stream erosion processes will potentially transfer radioactive waste and residual inventory from these facilities to the ground surface on adjacent hillslope areas where erosion has breached the facility. Hence, it is important to evaluate the consequences of potential exposures to a Resident Farmer on the hillslopes below a breached facility. Two interrelated aspects of the dose assessment model related to hillslope exposure are discussed: 1) representation of the physical processes related to transport of radionuclides from facilities onto the hillslopes, and from the hillslopes into adjoining creeks where contaminated material migrates downstream with surface water and sediment, and; 2) adaptation of the activities associated with the Resident Farmer scenario to assess potential exposures to contamination in the hillslope areas. This discussion will cover the conceptual basis of the hillslope exposure and transport models, and also implementation in the PPA computer model. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Performance Assessment for the Environmental Restoration Disposal Facility (Annual Status Report FY 2022)

DOE O 435.1 and DOE M 435.1-1 require that a determination of continued adequacy of the performance assessment (PA) (CP-60089), composite analysis (CA), and disposal authorization statement (DAS) be made on an annual basis, and that the determination must consider the results of data collection and analysis from research, field studies, and monitoring as well as the need to update any Radioactive Waste Management Basis (RWMB) documents. Beginning in 1996, the Environmental Restoration Disposal Facility (ERDF) started accepting low-level radioactive, hazardous, and mixed wastes that were generated during cleanup activities at the Hanford Site. ERDF is composed of a series of cells or disposal areas and can accommodate future design expansions as needed. Currently, there are eight cells and two supercells in ERDF. Each supercell is the equivalent of two cells. During this reporting period (fiscal year 2022, extending from October 1, 2021, through September 30, 2022), approximately 8.52E+04 metric tons (9.39E+04 U.S. tons) of waste was disposed at ERDF. From ERDF inception through September 30, 2022, approximately 17.0 million metric tons (18.7 million U.S. tons) of waste has been disposed at ERDF, which equates to consumption of approximately 89.1% of the currently constructed disposal volume. According to the design of ERDF, the facility has the ability to be expanded as needed. As a condition of the DAS, disposal operations within ERDF must be in accordance with the waste acceptance criteria (ERDF-00011) that provide specific radionuclide disposal limits, waste form restrictions, and descriptions of acceptable waste packages in compliance with the requirements of DOE M 435-1.1. The ERDF waste acceptance criteria stipulate that waste destined for disposal at ERDF be controlled based on source, physical form, and contaminant concentration and activity levels. There have been no changes to the physical configuration of ERDF or to the waste forms (source, physical form, etc.). No new unreviewed disposal question screenings or evaluations have been generated in this reporting period. Therefore, there are no noted impacts to the PA, CA, DAS, or RWMB resulting from the evaluations and screenings. Sum-of-fractions analysis shows that the disposed inventory meets both the concentration and inventory threshold requirements. A sum-of-fractions value is computed for ERDF sensitive radionuclides contributing to the groundwater pathways and the air pathway inventory limits. Computed values were 5.50E-04 and 3.30E-03, respectively. The disposed waste inventory remained well under the PA imposed limits, as shown in Table 4 and Table 5 in the main text of this report. Required monitoring was satisfactorily completed during the fiscal year reporting period. Compliance with performance objectives were met as each of the reported values were well below the established limit. Overall, there are no substantive changes to primary PA assumptions or changes to the PA analysis conclusion; therefore, compliance with DOE O 435.1 and the DAS is maintained.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Performance Assessment for the Environmental Restoration Disposal Facility (Annual Status Report FY 2021)

DOE O 435.1 and DOE M 435.1-1 require that a determination of continued adequacy of the performance assessment (PA) (CP-60089), composite analysis (CA), and disposal authorization statement (DAS) be made on an annual basis, and that the determination must consider the results of data collection and analysis from research, field studies, and monitoring as well as the need to update any Radioactive Waste Management Basis (RWMB) documents. Beginning in 1996, the Environmental Disposal Facility (ERDF) started accepting low-level radioactive, hazardous, and mixed wastes that were generated during cleanup activities at the Hanford Site. ERDF is composed of a series of cells or disposal areas and can accommodate future design expansions as needed. Currently, there are eight cells and two supercells in ERDF. Each supercell is the equivalent of two cells. During this reporting period (fiscal year 2021, extending from October 1, 2020, through September 30, 2021), approximately 9.14E+04 metric tons (1.01E+05 U.S. tons) of waste was disposed at ERDF. From ERDF inception through September 30, 2021, approximately 16.9 million metric tons (18.9 U.S. tons) of waste has been disposed of at ERDF, which equates to consumption of approximately 88.7% of the currently constructed disposal volume. According to the design of ERDF, the facility has the ability to be expanded as needed. As a condition of the DAS, disposal operation within ERDF must be in accordance with the waste acceptance criteria (ERDF-00011) that provide specific radionuclide disposal limits, waste form restrictions, and descriptions of acceptable waste packages in compliance with the requirements of DOE M 435-1.1. The ERDF waste acceptance criteria stipulate that waste destined for disposal at ERDF be controlled based on source, physical form, and contaminant concentration and activity levels. There have been no changes to the physical configuration of ERDF or to the waste forms (source, physical form, etc.). No new unreviewed disposal question screenings or evaluations have been generated in this reporting period. Therefore, there are no noted impacts to the PA, CA, DAS, or RWMB resulting from the evaluations and screenings. Sum-of-fractions analysis shows that the disposed inventory meets both the concentration and inventory threshold requirements. A sum-of-fractions value is computed for ERDF sensitive radionuclides contributing to the all pathways and the air pathway inventory limits. Computed values were 7.63E-03 and 1.49E-03, respectively. The disposed waste inventory remained well under the PA imposed limits, as shown in Table 4 and Table 5 in the main text of this report. Required monitoring was satisfactorily completed during the fiscal year reporting period. Compliance with performance objectives were met as each of the reported values were well below the established limit. Overall, there are no substantive changes to primary PA assumptions or changes to the PA analysis conclusion; therefore, compliance with DOE O 435.1 and the DAS is maintained.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Performance Assessment for the Environmental Restoration Disposal Facility (Annual Summary Report - FY 23)

Beginning in 1996, the Environmental Restoration Disposal Facility (ERDF) started accepting low-level radioactive, hazardous, and mixed wastes that were generated during cleanup activities at the Hanford Site. ERDF is composed of a series of cells or disposal areas and can accommodate future design expansions as needed. Currently, there are eight cells and two supercells in ERDF. Each supercell is the equivalent of two cells. During this reporting period (fiscal year 2023, extending from October 1, 2022, through September 30, 2023), approximately 9.37E+04 metric tons (1.03E+05 U.S. tons) of waste was disposed at ERDF. From ERDF inception through September 30, 2023, approximately 17.1 million metric tons (18.8 million U.S. tons) of waste has been disposed at ERDF, which equates to consumption of approximately 89.1% of the currently constructed disposal volume. According to the design of ERDF, the facility has the ability to be expanded as needed.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
NASA NTRS2007

Proposed Facility Modifications to Support Propulsion Systems Testing Under Simulated Space Conditions at Plum Brook Station's Spacecraft Propulsion Research Facility (B-2)

Preparing NASA's Plum Brook Station's Spacecraft Propulsion Research Facility (B-2) to support NASA's new generation of launch vehicles has raised many challenges for B-2 s support staff. The facility provides a unique capability to test chemical propulsion systems/vehicles while simulating space thermal and vacuum environments. Designed and constructed 4 decades ago to support upper stage cryogenic engine/vehicle system development, the Plum Brook Station B-2 facility will require modifications to support the larger, more powerful, and more advanced engine systems for the next generation of vehicles leaving earth's orbit. Engine design improvements over the years have included large area expansion ratio nozzles, greater combustion chamber pressures, and advanced materials. Consequently, it has become necessary to determine what facility changes are required and how the facility can be adapted to support varying customers and their specific test needs. Instrumental in this task is understanding the present facility capabilities and identifying what reasonable changes can be implemented. A variety of approaches and analytical tools are being employed to gain this understanding. This paper discusses some of the challenges in applying these tools to this project and expected facility configuration to support the varying customer needs.

Edwards, Daryl A.↗
DOE OSTI2020

Sludge Removal: Success and Partnership with T Plant - 20510

U.S. DOE contractor CH2M Hill Plateau Remediation Company (CHPRC) has safely and efficiently managed the removal of highly radioactive sludge from the 105-K West Reactor fuel storage basin (105- KW Basin) near the Columbia River. In mid-September 2019, sludge retrieval was completed after 21 storage containers were safely loaded with highly radioactive sludge and transported approximately 13 miles from the 105-KW Basin to T Plant, located in the 200 West Area on the Central Plateau of the Hanford Site. Sludge removal represents the last major source term reduction necessary before the K West Reactor and fuel storage basin can transition to closure activities in preparation for final deactivation and demolition. The technical complexity of the sludge retrieval process, coupled with the challenging physical and radiological characteristics of the waste, necessitated a methodical and deliberate approach using unique design and operational solutions to safely conduct the work. This challenge was further complicated by the need to plan and conduct the work in two separate facilities, each subject to the controls and requirements of its own nuclear facility safety basis. The project overcame these challenges through creating an integrated team that actively engaged, communicated and coordinated each phase of the project to ensure successful completion. This paper will cover how these key elements of the Sludge Removal Project led to its success: - Integrating project planning and management; - Lessons learned from design and construction; - Preparing for a successful campaign (lessons learned from testing and start-up); - Lessons learned from operating in two separate facilities, transporting sludge containers on the Hanford Site and ensuring compliance with a range of regulatory and safety basis requirements. The establishment of an integrated group spanning two facilities and multiple organizations created a team capable of overcoming the challenges necessary to successfully plan and execute the sludge retrieval mission. Sharing lessons learned from this successful project can enhance the ability of teams across the DOE complex to successfully plan and execute complex projects. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Rescue of Unique Liquid IP-2 Portable Tanks at Savannah River Site – 25078

Westinghouse Savannah River Company, the SRS Management and Operations Contractor at the time, procured six large IP-2 portable tanks capable of transporting liquids in 2003 to support the layup of the F Canyon at SRS. F Canyon was one of two full-scale hardened radiochemical separations facilities built at SRS in the 1950s as part of the initial construction of the Site. F Canyon and H Canyon provided actinide separations capabilities for SRS through the end of the Cold War. Following the end of the Cold War era, the decision was eventually made that only one of these radiochemical separations facilities at SRS would still be needed moving forward. As H Canyon was considered better suited for ongoing SRS missions, F Canyon was placed into a layup state in the first decade of the 21st Century, with F Area deactivation completing in 2006. This process resulted in the creation of excess, radioactively contaminated solutions, which required disposition. These solutions included contaminated organic solutions as well as DU-bearing aqueous solutions. The six portable tanks were procured to support the transportation of these solutions.

Fee, Nathaniel C. [Savannah River Nuclear Solution↗