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At least 955 records · Page 53
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Requalification of historical intermediate level waste (ILW) for the repository KONRAD on basis of non destructive assay (NDA) and re-evaluated nuclide vectors - 15289

The storage facility for ILW of the HDB contains a large number of historical waste products in 200-l drum geometry, which were produced in the 1970's to the 1990's. In the next years more ILW will be treated and stored at HDB, exceeding the storage capacity of the ILW store. All waste products produced at HDB are treated according to the acceptance criteria of the repository KONRAD and are planned to final storage in this repository. In order to qualify the waste materials for final storage in KONRAD and to remove parts of the inventory from the storage chambers this project was started. The main issues of the project, which will be accompanied by the regulator, are product control measures like gas chromatographic analyses of remotely sampled gas syringes and non destructive assay techniques like?-spectrometry and neutron-analytics. The product control is followed by declaration on the basis of re-evaluated nuclide vectors and a repackaging of the drums in licensed containers. The packaging concept includes a 400-l drum as qualified over-pack, because several of the historical drums cannot be qualified as IP-II or type A package. The annual space between the two drums is filled with quartz sand for stability and additional shielding. The drum set is afterwards inserted into a qualified UBA-container provided by GNS. In the case that the product control measures and the reassessment of the declarations result in a violation of the activity limits of the relevant acceptance criteria for KONRAD, the respective drum has to be physically re-qualified after correspondence with the surveillance authorities. This program will be applied to a fraction of approx. 400 canisters, which were selected on the basis of several quality aspects like dose rate below 100 mSv/h, nuclide inventory below 100 A2 according to paragraph 201 of TS-R-1 and well-known waste composition. In the future more waste products will be treated in the same way, if approved containers exist for the inventory. (authors)

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

Thermal Analysis of a Pipe Over-pack Container with a Combustible Waste Matrix for Los Alamos National Laboratory - 17307

Los Alamos National Laboratory (LANL) Technical Area 55 (TA 55) utilizes several different container types to store and transfer special nuclear material and waste for numerous programs. The Nuclear Process Infrastructure - Infrastructure Operations Group (NPI-2) analyzes and certifies containers to meet the specifications for distinct processes and storage. Engineering analysis performed on these containers include but are not limited to: leak testing, filter efficiency testing, pressure drop test, drop testing, water penetration testing, polymer (o-ring) hardness testing, and thermal steady state testing. Of all of these tests, thermal steady-state testing is most crucial to ensure critical components for specific containers do not fail under an internal heat payload. This testing provides data for setting heat payload limits for different containers. The Pipe Over-pack Container (POC) is a vented carbon steel container with a removable lid designed for storing or transferring nuclear waste. Additionally, TA 55 has been tasked to manage and store all of its Transuranic (TRU) waste for the foreseeable future. Being able to place more material into a single container conserves the physical storage space available at TA 55. The Pipe Over pack Handling and Operations Manual requires each user to ensure that the maximum temperature of each component in the POC is not exceeded. Conducting a thermal steady-state test with a combustible waste matrix on the POC can benefit stakeholders on increasing the material limits of the container. The results of these measurements will establish new wattage limits to meet LANL's transportation requirements. The current LANL Transportation Safety Document (TSD) limits the amount of heat source plutonium to 10 grams per POC, corresponding to approximately 5 Watts of heat. The result of this test was provided to the Packaging and Transportation (OS-PT) group to evaluate new wattage limits that can satisfy the needs of TA 55, while maintaining a safe packaging configuration. (authors)

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

Advanced Reactors Spent Fuel & Waste Science and Technology Program

Based on the higher interest in Advanced Reactor (AR) deployment (e.g., ARDP ) for potential new fuel cycles, the Spent Fuel & Waste Science and Technology (SFWST) Program has begun to evaluate the possible implications of long-term management and final disposition of potential Advance Reactor spent nuclear fuels (SNF) that would be generated in potential advanced reactors. Safely managing and dispositioning the potential future AR SNF, and any other associated radioactive wastes, is the primary focus of this initial preliminary assessment of those. This paper summarizes three primary tasks the Spent Fuel & Waste Science and Technology (SFWST) Program is executing (or collaborating on) related to the back end of the nuclear fuel cycle (BENFC) for potential future advanced reactors: 1. Advanced Reactors Spent Fuel and Waste Streams: Strategies for the BENFC This set of activities define a high-level strategy for how to systematically approach, identify, and close research and development (R&D) gaps/potential issues associated with long-term management and final disposition of AR SNF and other possible AR waste streams. This task involves summarizing advanced reactor concepts, their likely SNF and other waste forms, and identifying previous experience with similar materials, for example from DOE-managed SNF, with closely related characteristics to the potential future AR SNF. Technical R&D gaps between the breadth of detailed understanding for safe storage, transportation and disposal of the existing light water reactor SNF fuel cycle (e.g., see NASEM, 2022) and potential future fuel cycles based on advanced reactors would then be identified. 2. Characterization and Packaging Options of Advanced Reactor SNF These activities evaluate characteristics and packaging options for advanced reactor spent fuel forms. The fuel forms are categorized into three types: (1) tri-structural isotropic (TRISO), (2) metallic, and (3) fuel salt. Emphasis is given to TRISO and metallic SNF and additional waste streams from such AR as driven by the near-term anticipated operation of the Xe-100 and the Natrium reactors as advanced-reactor demonstrations1. Preliminary information for the spent-fuel salt discharged from molten-salt reactors (MSRs) is examined to provide a baseline for future efforts. All calculations and assumptions in this work are based on publicly available information. The following characteristics are calculated or estimated for use in the preliminary assessments: SNF volume and mass, radiation/activity levels through time, thermal conditions through time, potential radionuclide source terms, chemical interactions and evolutions, disposal inventories, and waste-form lifetime. Using those characteristics, calculations to determine the applicability of existing canister designs were performed. These evaluations included geometric (e.g., dimension, volume) and mass/weight considerations, known operational approaches and loading procedures, physical and chemical considerations/conditions for storage environments, as-loaded radiation, thermal, and criticality analyses to identify constraints for storage, transportation, and disposal. 3. Back-End Management of Advanced Reactors (BEMAR) The DOE NE-8 organization has defined an Integrated Project Team to evaluate the Back End Management of Advanced Reactors (BEMAR) (includes DOE staff from a range of organizations (e.g., NE-81, NE-82, OCED) and national laboratory technical staff within the DOE NE-81 and NE-82 programs). This BEMAR group works directly with advance reactors developers to assess for the DOE the technical feasibility of storage, transportation, and disposal of AR SNF based on the characteristics provided by the developers to DOE (much of which is proprietary). The BEMAR is also tasked to develop rough-order-of-magnitude cost estimates to compare the waste management system for individual advanced reactors to existing light-water reactor management practices. To accomplish this, the BEMAR group is implementing a Systems Engineering approach.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Development of the Mobile Systems for Conditioning of Disused Sealed Radioactive Sources in Serbia - 20105

Sealed radioactive sources (SRS) are being used worldwide in the field of medicine, agriculture, industry and research. They can be found in mobile as well as stationary devices. SRS contains radioactive material that is (a) permanently sealed in a capsule or (b) closely bounded and in a solid form. The capsule or material of an SRS should be strong enough to maintain leak tightness under the conditions of use and purpose for which the source was designed, also in case of accidents. In this case only emitted radiation is utilized. Firstly, the hazard from external radiation has to be considered, but the possibility of contamination due to fracture of the capsule should not be disregarded. The radioactive sources are composed of the radiating isotope contained in the filling medium, the single or double isotope holder that partially or totally surrounds the filling medium, the outer cover that contains the parts mentioned above and the capsule closed airtightly by welding or using some other method. The capsule must be tested for leakage periodically. If the SRS is no longer needed (e.g. replaced by a different technique) or it becomes useless for the intended application (e.g. the activity becomes too weak, the equipment containing the source works poorly or becomes obsolete, the source is damaged or leaking) it is considered disused. Disused sealed radioactive sources (DSRS) are typically conditioned and disposed if a facility is available. If the disposal option is not available, conditioned DSRS should be stored under proper conditions. In some cases, the radionuclide(s) in DSRS can be recovered/recycled or the DSRS can be repurposed for other applications. Conditioning of DSRS ensures containment of the radioactive material, provides confinement for leaking sources, provides sufficient radiation shielding, reduces storage/disposal volume by allowing consolidation of multiple sources into a single storage/disposal container, facilitates transport operations and contributes to safety and security as well. Typically, conditioning technologies are deployed either as permanently installed stationary systems in centralized or mobile on-site waste processing facilities, or in a mobile configuration. Centralized stationary facilities provide a single processing location for multiple users that requires transport of the waste to the facility. On the other hand, mobile systems may be provided for the selection and application of the optimum technology for a specific waste stream (such as DSRS) by bringing the process to the point where the waste is generated. In addition, mobile systems could offer additional flexibility by sharing equipment among multiple waste generating sites for processing campaigns that vary in duration, from very short periods to several years. The term 'mobile processing system' refers to any radioactive waste processing system or component which is designed to be transportable and which is not considered permanently installed. Two mobile system for conditioning of disused sealed radioactive sources are developed in the Public Company Nuclear Facilities of Serbia. Development of these mobile systems was supported by SRB9005 national project via Technical Cooperation of the IAEA. The first mobile system, built inside the 20 feet ISO container, will be used for conditioning of DSRS category 3 to 5. The second mobile system, built inside the 7 m long vehicle (Iveco Daily Van), will be used for dismantling of ionizing smoke detectors mostly with Am-241 sources. Designs of the mobile systems were defined in cooperation with two companies from Belgium (Belgoprocess and Leniko) and a Croatian company Ekoteh as well as with the support of the IAEA experts. The generic safety assessment and operational procedures for the mobile systems are developed. Based on safety assessment the acceptance criteria and operational limits and conditions are established. Operational procedures include: (a) equipment and material requirements, (b) assembling procedure of the mobile unit, (c) procedure for acceptance of devices for dismantling and conditioning, (d) dismantling procedure for devices to recover the DSRS, (e) characterization of DSRS, (f) encapsulation procedure of DSRS, (g) disassembling procedure of the mobile unit, and (h) keeping records, identification and traceability. In addition, radiation safety, health safety, security and emergency preparedness plans are prepared. The generic safety assessment and operational procedures could be updated with site specific requirements, DSRS inventory, and different needs for future customers. Developed mobile systems could be used in all situations when it is feasible to perform conditioning of DSRS on the spot in the county and worldwide. Development of these mobile units was just the first step to create the Reference Center for Radioactive Waste Treatment and Disused Radioactive Sources Conditioning for Small Facilities which can become a regional training center in the future, and/or as a tool for comprehensive national search and secure programmes. In the next phases development of e-learning platforms and blended learning packages as well as application for the IAEA Qualified Technical Centre (QTC) for the management of DSRS is foreseen. (authors)

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

Support for Advanced Fuel Fabrication MC&A Approaches

This report summarizes the key activities of the Material Protection, Accounting, and Control Technologies (MPACT) program’s safeguards training work package for FY25. The work focused on two main areas: domestic safeguards training and the development of new technologies for nuclear material accounting and control (NMAC). A key development during this FY was the delivery of NMAC statistics for US Nuclear Regulatory Commission requirements refresher course. The course objective was to assist industry in meeting the MC&A program requirements for fuel cycle facilities; this training focused on front-end bulk fuel cycle processes. At the same time, a report was prepared on the research output of an effort that focused on the benefits of commercial fuel debundling for NMAC and waste management. The MPACT team evaluated various technologies and proposed a fuel debundling detector that integrates gamma spectroscopy and neutron detection as the most practical solution. The scope of activities under this work package for FY26 focuses on developing a fundamental nuclear material control plan, also known as MC&A plan, for tristructural isotropic (TRISO) fuel fabrication facilities. This plan will address the various NMAC requirements for facilities that handle Category II.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Defining WAC for interim storage dedicated to tritium decay - 15164

Considering the high mobility of tritium through the package in which it is contained, the new 50-year storage concepts proposed by the CEA currently provide a solution adapted to the management of waste with high tritium concentrations, associated with adapted disposal sites. The 50-year intermediate storage corresponds to 4 tritium radioactive periods i.e. a tritium reduction by a factor 16. The waste acceptance criteria (WAC) for this interim storage concept will take into account the safety of the facility, the reference scheme for the management of tritiated waste in France, and the criteria applicable at Andra disposal centers. This will lead to define a set of waste specifications that describe the generic criteria such as acceptable waste forms, general principles and specific issues, e.g. conditioning, radioactive content, tritium content, waste tracking system, and quality control. This paper details how the work by the CEA has led to a storage facility project that not only takes into account the specificity of tritium, but also the producers' needs and Andra's disposal requirements. (authors)

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

The Time Needed to Implement the Blue Ribbon Commission Recommendation on Interim Storage - 13124

The report of the Blue Ribbon Commission on America's Nuclear Future [1] makes a number of important recommendations to be considered if Congress elects to redirect U.S. high-level radioactive waste disposal policy. Setting aside for the purposes of this discussion any issues related to political forces leading to stopping progress on the Yucca Mountain project and driving the creation of the Commission, an important recommendation of the Commission was to institute prompt efforts to develop one or more consolidated storage facilities. The Blue Ribbon Commission noted that this recommended strategy for future storage and disposal facilities and operations should be implemented regardless of what happens with Yucca Mountain. It is too easy, however, to focus on interim storage as an alternative to geologic disposal. The Blue Ribbon Commission report does not go far enough in addressing the magnitude of the contentious problems associated with reopening the issues of relative authorities of the states and federal government with which Congress wrestled in crafting the Nuclear Waste Policy Act [2]. The Blue Ribbon Commission recommendation for prompt adoption of an interim storage program does not appear to be fully informed about the actions that must be taken, the relative cost of the effort, or the realistic time line that would be involved. In essence, the recommendation leaves to others the details of the systems engineering analyses needed to understand the nature and details of all the operations required to reach an operational interim storage facility without derailing forever the true end goal of geologic disposal. The material presented identifies a number of impediments that must be overcome before the country could develop a centralized federal interim storage facility. In summary, and in the order presented, they are: 1. Change the law, HJR 87, PL 107-200, designating Yucca Mountain for the development of a repository. 2. Bring new nuclear waste legislation to the floor of the Senate, overcoming existing House support for Yucca Mountain; 3. Change the longstanding focus of Congress from disposal to storage; 4. Change the funding concepts embodied in the Nuclear Waste Policy Act to allow the Nuclear Waste fund to be used to pay for interim storage; 5. Reverse the Congressional policy not to give states or tribes veto or consent authority, and to reserve to Congress the authority to override a state or tribal disapproval; 6. Promulgate interim storage facility siting regulations to reflect the new policies after such changes to policy and law; 7. Complete already underway changes to storage and transportation regulations, possibly incorporating changes to reflect changes to waste disposal law; 8. Promulgate new repository siting regulations if the interim storage facility is to support repository development; 9. Identify volunteer sites, negotiate agreements, and get Congressional approval for negotiated benefits packages; 10. Design, License and develop the interim storage facility. The time required to accomplish these ten items depends on many factors. The estimate developed assumes that certain of the items must be completed before other items are started; given past criticisms of the current program, such an assumption appears appropriate. Estimated times for completion of individual items are based on historical precedent. (authors)

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

The diverse applications of plasma

Plasma being the fourth state of matter has always been an attraction for Physicists and Chemists. With the advent of time, plasma energy has been recognized in having widening horizons in the field of Biomedical Sciences. Plasma medicine can be subdivided into three main fields; Non-thermal atmospheric-pressure direct plasma for medical therapy; Plasma-assisted modification of bio-relevant surfaces and Plasma-based bio-decontamination and sterilization. The basis of the research is that as it has free carrier molecules, it has the ability to target specific cells and regulate functions like wound healing. Plasma does not harm healthy human cells but can kill bacteria and possibly even cancer cells to help treat various diseases. Nosocomial infection control, prevention and containment of contagious diseases, disinfection of medical devices, surface treatment (heat and UV sensitive surfaces) are research of interest. Recent success in generating plasma at very low temperature ie. Cold plasma makes the therapy painless. It has the ability to activate cellular responses and important mechanisms in the body. They target specific molecules such as prothrombin for blood coagulation, cytokines for killing bacteria, and angiogenesis for tissue regeneration. Plasma has bactericidal, fungicidal and virucidal properties. Plasma technology has flourishing future in diverse fields like Textiles, Nanofabrication, Automotives, Waste management, Microbiology, Food Hygiene, Medical Science like Skin treatments, sterilisation of wounds, Hand disinfection, Dental treatments etc. Food hygiene using plasma can be achieved in disinfection of food containers, food surface disinfection, hygiene in food handling, preparation and packaging. Therefore Plasma is most promising field for budding Scientist for fluorishing research in Biological Sciences.

60 APPLIED LIFE SCIENCES↗
NASA NTRSDate not supplied

Considerations For Waste-to-Base Future Research Paths

Developing innovative technologies to reprocess low value waste into high value base products will enable a new “Waste-to-Base” (WtB) capability leading to innovative approaches to space mission design. Systems engineers and mission architects will work together under a new paradigm in logistics planning. Equipment, components, packaging, storage, and structures will be understood not only by functional requirements, but also in terms of useful lifetimes and materials of construction. The advent of a new technical discipline - reprocess engineering - will grow from the intersection between chemical, mechanical, and systems engineering, mission design, and logistics planning. A crowdsourcing approach was employed to gather low Technology Readiness Level (TRL) concepts into workable solutions for issues in trash management, CO2 processing, foam reprocessing, and fecal processing. The results of the crowdsourcing activities are summarized along with a discussion for future technology development work.

Logistics reduction↗
DOE OSTI2021

Nevada National Security Site V3XA Spherical Explosive Confinement Vessel Awareness Package

This report summarizes the plans and activities that will facilitate shipment of the V3XA experimental spheres (a.k.a., Nevada Spheres) from the Nevada National Security Site (NNSS) to the Idaho National Laboratory’s (INL) Advanced Mixed Waste Treatment Project (AMWTP). The spheres require shipment to INL to be segmented, characterized, and repackaged for ultimate disposal at the Waste Isolation Pilot Plant (WIPP). The NNSS does not currently have the facilities or safety basis to perform these activities. The INL has the facilities, the INL M&O contractor owns an 8-120B cask, and AMWTP has a WIPP Waste Acceptance Criteria approved characterization program. The NNSS has shipped other legacy transuranic waste to INL for WIPP certified characterization in the past.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
NASA NTRSDate not supplied

Considerations For Waste-to-Base Future Research Paths

Developing innovative technologies to reprocess low value waste into high value base products will enable a new “Waste-to-Base” (WtB) capability leading to innovative approaches to space mission design. Systems engineers and mission architects will work together under a new paradigm in logistics planning. Equipment, components, packaging, storage, and structures will be understood not only by functional requirements, but also in terms of useful lifetimes and materials of construction. The advent of a new technical discipline - reprocess engineering - will grow from the intersection between chemical, mechanical, and systems engineering, mission design, and logistics planning. A crowdsourcing approach was employed to gather low Technology Readiness Level (TRL) concepts into workable solutions for issues in trash management, CO 2 processing, foam reprocessing, and fecal processing. The results of the crowdsourcing activities are summarized along with a discussion for future technology development work.

Logistics reduction↗
NASA NTRS1987

Space station resistojet system requirements and interface definition study

Preliminary resistojet design requirements were established based on initial technical requirements imposed by the results of NASA and Rocketdyne studies. The requirements are directed toward long life, simplicity, flexibility, and commonality with other space station components. The resistojet assembly is comprised of eight resistojets, fluid components downstream of the waste fluid storage system, a power controller, structure, and shielding. It consists of two identical subassemblies, one of which is redundant. Each subassembly consists of four 500-W resistojets, series redundant latch values, a power controller, a water vaporizer, two pressure regulators, filters, check valves, disconnects, fluid tubing, and electrical cables. All components are packaged at the end of the stinger aft of the JEM and Columbus modules. Different flow and power control methods were studied. A constant inlet pressure and a two-power setting controller were tentatively selected based on simplicity and reasonably high specific impulse for the range of waste gas compositions that are anticipated. The constant pressure is supplied by pressure regulators. The two set point power control includes individual power supplies to each resistojet heater and water vaporizer. An embedded data processor, a multiplexer-demultiplexer, and a network interface unit that are standard space station components are included in the power controller. The total dry weight of the resistojet assembly is approximately 172 lb. The total cost for design, development, test, evaluation, qualification, and flight hardware is estimated to be $16 million.

Heckert, Bruce J.↗
DOE OSTI2019

Dose Calculation Methodology and Data for Solid Waste Performance Assessment and Composite Analysis at the Savannah River Site

This report provides a detailed description of the methodology developed to perform dose calculations for E-Area Low Level Waste Facility Performance Assessments and Savannah River Site Composite Analyses. Chapters 2 – 8 give a complete set of equations to be used to calculate dose for individual exposure pathways. The first set of dose scenarios assume that a member of the public establishes residence near the waste site and uses contaminated groundwater or contaminated surface water for personal consumption and to irrigate a garden and pasture where produce and farm animals are raised. Products from the garden and farm animals are used for personal consumption. This scenario applies to both E-Area Performance Assessments and Savannah River Site Composite Analyses. General dose exposure pathways for the resident farmer scenario are: 1. Ingestion pathways described in Chapter 2, 2. Inhalation pathways described in Chapter 3, 3. External exposure pathways described in Chapter 4, and 4. Recreational pathways described in Chapter 5 (Composite Analysis only). Doses from all of the above pathways are directly related to the contaminant concentration in the water. Additional dose scenarios considered for E-Area Performance Assessment assume that an inadvertent intruder encroaches on the waste disposal site after loss of institutional control. General dose exposure pathways considered for the inadvertent intruder are: 1. Ingestion pathways described in Chapter 6, 2. Inhalation pathways described in Chapter 7, and 3. External exposure pathways described in Chapter 8. Doses from all of the intruder pathways are directly related to the contaminant concentration in the buried waste. Special dose considerations related to human consumption of meat and dairy, and treatment of tritium in dose calculations are discussed in Chapter 9. Chapter 10 describes an Excel database named “Radionuclide, Element and Dose Parameter Data Package” that has been created to provide a system for maintaining the data required to perform radionuclide transport and dose calculations. This data includes: Isotope physical parameters and radioactive decay data needed for both transport and dose calculations, Isotope-specific dose coefficients and soil shielding factors associated with exposure pathways, Radionuclide drinking water concentration limits either published by the Environmental Protection Agency or derived from internal dose coefficients to meet the groundwater protection requirements in Department of Energy Order 435.1, Element-specific bio-transfer factors used in dose calculations, Dose equations and values of physical parameters and human factors used in dose calculations, and Key physical constants used in dose calculations. Revision 1 incorporates updated information developed since this report was issued for use in the next revision of the E-Area Low-Level Waste Facility Performance Assessment.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2024

Concept of Operations for Advanced Reactor Spent Nuclear Fuel Management

This presentation presents a preliminary description of a concept of operations to incorporate advanced-reactor spent nuclear fuel (SNF) into an integrated waste-management system (IWMS). The evaluation includes SNF from four advanced-reactor concepts with the following fuel types: (1) small modular reactors using oxide fuels, (2) tristructural-isotropic (TRISO) fuels, (3) metallic fuels, and (4) fuel salts. To provide context for the proposed concept of operations for advanced reactors, a comparison is made with traditional light-water reactors (LWRs) to identify potential gaps in the IWMS. The technical differences between advanced reactors and LWRs are assessed to determine the feasibility of managing advanced-reactor waste streams using existing operations and technology. This presentation emphasizes fuel types from Advanced Reactor Demonstration Program reactors: Xe 100, which uses TRISO fuel and Natrium, with its metallic fuels while also analyzing management options for molten-salt reactors and advanced light-water reactors (ALWRs). Understanding the storage, transportation, and disposal requirements of SNF is dependent on both the quantity and characteristics of the SNF generated by nuclear reactors. This presentation provides a high level overview, comparing the anticipated concept of operations for different SNFs from advanced reactors. The IWMS includes at-reactor storage, transportation, potential off-site storage, potential treatment, and disposition. To assess the potential effect of advanced-reactor concept of operations on the IWMS, estimates were made for fuel characteristics that contribute to storage, transportation, disposal, and possible treatment of advanced-reactor SNF. These include canister heat load, dose rates, and criticality-control limits, which are important for determining the condition and configuration of the advanced reactor SNF. At-reactor storage of LWR SNF traditionally involves a spent-fuel pool (SFP) before transfer to an independent spent-fuel storage installation. However, some advanced-reactor concepts, particularly those using TRISO and salt fuels, do not anticipate the use of an SFP. This difference in at-reactor storage could impact the IWMS. Additionally, transportation of advanced-reactor SNF may include additional processes tied to potential off gassing, and transportation of microreactor SNF may occur within the reactor vessel. Some advanced-reactor SNF could also undergo treatment to meet requirements of an acceptable waste form for disposition, and the treatment location will be a major contributor to efficiently performing IWMS responsibilities. Moreover, the quantity of SNF generated is an important consideration for IWMS because it could affect the size of the transportation fleet and potential off-site storage requirements. Additionally, volume and heat load are the primary drivers for SNF disposition. This presentation compares potential packaging options for advanced reactor SNF with traditional LWR packaging to provide a high-level comparison for the needs of the IWMS for advanced-reactor SNF.

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

Assessment of SFR fuel pin performance codes under advanced fuel for minor actinide transmutation

Americium is a strong contributor to the long term radiotoxicity of high activity nuclear waste. Transmutation by irradiation in nuclear reactors of long-lived nuclides like {sup 241}Am is, therefore, an option for the reduction of radiotoxicity and residual power packages as well as the repository area. In the SUPERFACT Experiment four different oxide fuels containing high and low concentrations of {sup 237}Np and {sup 241}Am, representing the homogeneous and heterogeneous in-pile recycling concepts, were irradiated in the PHENIX reactor. The behavior of advanced fuel materials with minor actinide needs to be fully characterized, understood and modeled in order to optimize the design of this kind of fuel elements and to evaluate its performances. This paper assesses the current predictability of fuel performance codes TRANSURANUS and GERMINAL V2 on the basis of post irradiation examinations of the SUPERFACT experiment for pins with low minor actinide content. Their predictions have been compared to measured data in terms of geometrical changes of fuel and cladding, fission gases behavior and actinide and fission product distributions. The results are in good agreement with the experimental results, although improvements are also pointed out for further studies, especially if larger content of minor actinide will be taken into account in the codes. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Analysis of Radiation Damage in Electronics of Unmanned Aerial and Ground Robotic Platforms - 19470

Gamma ray measurements are important for monitoring of nuclear facilities. Robotic platforms are used to carry radiation detectors and cameras to enable remote sensing and visual surveillance, allowing for the operation in hazardous, hard to reach zones. The prediction of radiation damage in electronic components of robots is vital. The radiation damage analysis was carried out for two robotic platforms exposed to neutron and gamma ray flux: a ground hexapod platform PhantomX (a 'spider robot') and an aerial platform - the DJI S1000+ octocopter. The FLUKA code was used to analyze the displacement per atom values in electronic components of these robots. The packaging of the components and the shielding design was determined to increase the robot's lifetime in areas where radiation dose levels are high. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
NASA NTRSDate not supplied

Considerations For Waste-to-Base Future Research Paths

Developing innovative technologies to reprocess low value waste into high value base products will enable a new “Waste-to-Base” (WtB) capability that can spawn a new approach to mission design. Systems engineers and mission architects will work together under a new paradigm in logistics planning. Equipment, components, packaging, storage, and structures will be understood not only by functional requirements, but also in terms of useful lifetimes and materials of construction. The advent of a new technical discipline - reprocess engineering - will grow from the intersection between chemical, mechanical, and systems engineering, mission design, and logistics planning. A crowdsourcing approach was employed to gather low TRL concepts into workable solutions for issues in trash management, CO2 processing, foam re-processing, and fecal processing. The results of the crowdsourcing activities are summarized along with a roadmap for future technology development work.

Logistics reduction↗
DOE OSTI2024

M3SF-24LL010301052-Summary of SUPCRTNE testing and Rev0 database

This progress report (Level 3 Milestone Number M3SF-24LL010301052) summarizes research conducted at Lawrence Livermore National Laboratory (LLNL) within the Argillite Disposal R&D work package SF-24LL01030105. SUPCRTNE is being developed as the primary engine for thermodynamic database development in support of geologic disposal of high-level nuclear waste. As used here, “SUPCRT” refers to both a computer program and its supporting database. Additional letters or numbers refer to specific variants (SUPCRT92, (Johnson et al., 1992); SUPCRTBL, (Zimmer et al., 2016); and SUPCRTNE). A SUPCRT database contains the data required to calculate the thermodynamic properties of solid, gas, and aqueous species over a wide range of temperature and pressure. Normally, SUPCRT is used to create a higher-level database that directly supports modeling and simulation codes such as EQ3/6, GWB, PHREEQC, and PFLOTRAN.

58 GEOSCIENCES↗