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At least 307 records · Page 17
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Scoping Analysis of Pebble-Bed Reactors for the Destruction of the Transuranic Inventory of LWR Spent Nuclear Fuel

With the forecasted increase in the construction and operation of nuclear reactors, there will be a corresponding increase in the quantity of spent nuclear fuel (SNF) that requires long-term storage. In SNF, transuranic isotopes contribute the most to the long-term radiotoxicity of the fuel and pose a proliferation risk. One option that has been explored to address these issues is the removal of the transuranic isotopes from SNF and the conversion of these isotopes into transuranic fuel (TRU fuel). Here, this work sought to determine how effective a micro-modular Pebble-Bed High-Temperature Gas-Cooled Reactor (PB-HTGR); the 10-MW High Temperature Gas-cooled Test Reactor (HTR-10); and a salt-cooled small-modular pebble-bed reactor (PBR), i.e. the generic Fluoride-cooled High-temperature Reactor (gFHR), are at reducing the inventory of transuranic isotopes while still maintaining the intrinsic safety features of the PBR designs, such as negative temperature coefficients of reactivity. Optimized pebble designs utilizing TRU fuel were found for both reactors through the adjustment for the packing fraction of fuel in each pebble. The Axial Zone Equilibrium Modeling (A-ZEM) method was used in this work to help select the optimized pebble design. Once an optimized pebble design was selected and an equilibrium model was produced, the results from the deep burn (DB) HTR-10 and gFHR designs were compared to the results of two models from the literature. While both the DB gFHR and the DB HTR-10 were able to reduce the weapons-usable transuranic inventory, the performance of these reactors did not match that of the small-modular PB-HTGRs in the literature. Therefore, a need was identified for further refinement of the gFHR design using TRU fuel, as the results for this model were more promising than those of the DB HTR-10, which was strongly limited by the high leakage intrinsic to micro-modular PB-HTGRs.

Transuranic fuel↗
DOE OSTI2023

Traveling Molten Zone Refining Process Development for Innovative Fuel Cycle Solutions

Considering the phase diagrams of metallic spent fuel constituents, the melting and solidifying of spent metallic fuel causes three immiscible layers (actinide-rich, lanthanide-rich, and Group II-rich) and the condensate phase (Group I) to form. We believe this anticipated immiscibility offers an untapped opportunity for innovative fuel cycle solutions. Through the proposed project, we anticipate confirming the expected phase behavior and develop a thermal treatment process to rapidly extract actinides from spent metallic fuels. The prime apparatus for both purposes is a traveling molten zone system with induction heating. We envision that one rapid pass of the molten zone from the bottom to the top of the metallic rod incorporating species of spent metallic fuels should produce the expected immiscible layer formation and provide species partitioning data effectively and cleanly. It will also demonstrate an actinide extraction process by concentrating the impurities at the top segment of the rod and leaving the actinide species behind as the bulk rod. The successful execution of the project will demonstrate proof of concept for a transformative process path for used metal fuels in terms of economics and safeguards.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Review of Residual Stresses in Spent Nuclear Fuel Canisters

This report compiles the set of published results on the welding residual stress (WRS) distributions across the thickness section at the welds in spent nuclear fuel (SNF) canisters before and after a repair is done. The WRS information for similar steel structures with thick (~ 12.7 mm (0.5 inch)) large plate construction are also compiled to provide a comprehensive set of information on: methods to quantify; modeling to estimate; and methods to reduce WRS.

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

An introduction to Spent Nuclear Fuel decay heat for Light Water Reactors: a review from the NEA WPNCS

This paper summarized the efforts performed to understand decay heat estimation from existing spent nuclear fuel (SNF), under the auspices of the Working Party on Nuclear Criticality Safety (WPNCS) of the OECD Nuclear Energy Agency. Needs for precise estimations are related to safety, cost, and optimization of SNF handling, storage, and repository. The physical origins of decay heat (a more correct denomination would be decay power) are then introduced, to identify its main contributors (fission products and actinides) and time-dependent evolution. Due to limited absolute prediction capabilities, experimental information is crucial; measurement facilities and methods are then presented, highlighting both their relevance and our need for maintaining the unique current full-scale facility and developing new ones. The third part of this report is dedicated to the computational aspect of the decay heat estimation: calculation methods, codes, and validation. Different approaches and implementations currently exist for these three aspects, directly impacting our capabilities to predict decay heat and to inform decision-makers. Finally, recommendations from the expert community are proposed, potentially guiding future experimental and computational developments. One of the most important outcomes of this work is the consensus among participants on the need to reduce biases and uncertainties for the estimated SNF decay heat. If it is agreed that uncertainties (being one standard deviation) are on average small (less than a few percent), they still substantially impact various applications when one needs to consider up to three standard deviations, thus covering more than 95% of cases. The second main finding is the need of new decay heat measurements and validation for cases corresponding to more modern fuel characteristics: higher initial enrichment, higher average burnup, as well as shorter and longer cooling time. Similar needs exist for fuel types without public experimental data, such as MOX, VVER, or CANDU fuels. A third outcome is related to SNF assemblies for which no direct validation can be performed, representing the vast majority of cases (due to the large number of SNF assemblies currently stored, or too short or too long cooling periods of interest). A few solutions are possible, depending on the application. For the final repository, systematic measurements of quantities related to decay heat can be performed, such as neutron or gamma emission. This would provide indications of the SNF decay heat at the time of encapsulation. For other applications (short- or long-term cooling), the community would benefit from applying consistent and accepted recommendations on calculation methods, for both decay heat and uncertainties. This would improve the understanding of the results and make comparisons easier.

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

Bounding Pressure and Flammability Evaluations of Aluminum-Clad Spent Nuclear Fuel Department of Energy Standard Canisters

This paper presents bounding pressure and flammability evaluations for DOE Standard Canister loaded with DOE-managed aluminum-clad spent nuclear fuel (ASNF). The objective of these evaluations is to gain confidence in the safety and feasibility of possible loading configurations for extended (>50 years) periods of dry storage, with particular focus on the dry storage canister pressures and potential for formation of a flammable atmosphere. The primary concern about the extended dry storage of ASNF is radiolytic gas generation. The aluminum cladding of these materials tends to corrode, and these corrosion products—typically aluminum oxides, such as boehmite, bayerite, or gibbsite—could carry water. This makes ASNF dry storage canisters difficult to dry. The gamma radiation field in dry storage environments could cause a radiolytic breakdown of residual water, forming chemical species such as molecular hydrogen (H2). The release of these species could increase the canister pressure and lead to the generation of a flammable canister atmosphere. The bounding evaluations presented within this study surmise conservative, but credible, conditions and processes. This includes the assumption of a full breakdown of a large quantity of free, physisorbed, and chemisorbed water (bound in a trihydrate, i.e., Al2O3 • 3H2O, layer). The considered dry storage configurations include a ~3 m (10 ft) long, ~46 cm (18 in) diameter (10 x 18) DOE Standard Canister loaded with 32 Advanced Test Reactor (ATR) ASNF elements, and a ~3 m (10 ft) long, ~61 cm (24 in) (10 x 24) diameter DOE Standard Canister loaded with 40 ATR ASNF elements. While the results of this study indicate the possibility of atmospheric hydrogen concentrations above the lower flammability limit, insufficient concentrations of oxygen will prevent the formation of flammable atmospheres. The maximum credible pressures remain well within the structural limits of the DOE Standard Canister.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

An Enhanced Safeguards Approach for Accelerator Driven Systems utilized to Close the Fuel Cycle

Further to our previous safeguards approach for Accelerator Driven Systems, which focused on estimates of fissile material production using relevant proton accelerator systems and corresponding safeguards needs for fuel storage, the subcritical reactor, and spent fuel storage material balances areas, this report is more expansive and considers utilization of ADS for either burning of transuranics or breeding of fissile materials. We find that the recycled fuels likely intended for ADS will be thermally and radioactively hot to such a degree that it is likely reprocessing and fuel fabrication will have to be co - located with the ADS reactor facility to avoid impractical hot fuel transportation issues. As such, we consider in detail the full ADS system to include material balance areas for spent fuel receiving, reprocessing, storage & cooling, fuel fabrication, subcritical reactor area, and waste storage & handling. Furthermore, aqueous - based separation methods like PUREX cannot tolerate the intense heat of the ADS fuels, so pyroprocessing will likely be required. With these considerations, we developed an Enhanced Safeguards Approach for ADS beyond the work done in our first report, and conclude that significant diagnostic development is needed , a nd provide safeguards recommendations. We have also included an appendix regarding some country programs, in particular the Chinese ADANES burner/breeder program a nd the Indian thorium - based breeder program.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

HELIUM LEAK TEST MODELING OF A SPENT NUCLEAR FUEL CANISTER

The U.S. Department of Energy (DOE) is considering the development of one or more federal consolidated interim storage facilities (CISFs) to be used to store commercial spent nuclear fuel (SNF) at locations in the U.S. One of the first technical challenges of a CISF is performing an inspection of SNF canisters upon their receipt to confirm they can be placed into the CISF’s licensed storage configuration. The canister receipt inspection is critical to CISF site operations. The test is conceived as being a helium (He) leak check, intended to confirm that the confinement boundary of a SNF canister is intact. SNF canisters are filled with He when they are sealed, so detection of a He leak indicates that a through-wall flaw has occurred in the canister confinement boundary. Other measurements are planned to occur upon canister receipt in addition to the He leak check such as krypton-85 measurements, which would indicate confinement breaches of one or more fuel rods in addition to a breach of the SNF canister. However, the He leak check has been identified as one of such high importance and has such significant technical challenges that a full-scale demonstration is needed to confirm the He leak test’s viability and to assist in planning relative to its operational requirements. A modeling methodology for simulating the He detection test was developed to help inform the test plan and the design of the test vessels. To develop the modeling methodology a detailed computational fluid dynamics (CFD) benchmark model was constructed to compare against leak rate test data from a transportation package for radioactive material. This report is focused on modeling efforts to simulate the benchmark leak test.

Suffield, Sarah R.↗
DOE OSTI2021

Fatigue Test Result for High Burnup Pressurized Water Reactor Rods

The High Burnup Spent Fuel Data Project, sponsored by DOE-NE, is focused on understanding the effects of long-term storage and transportation on high burnup (HBU) (>45 gigawatt days per metric ton uranium) light water reactor fuel. The experiment includes 32 HBU spent nuclear fuel (SNF) assemblies that will be stored in a typical Independent Spent Fuel Storage Installation (ISFSI) and 25 sister rods - nine SNF rods that were removed from the fuel assemblies prior to insertion to the ISFSI and 16 SNF rods removed from similar HBU assemblies. The sister rods provide a baseline of the condition of the HBU rods prior to loading, drying, and long-term dry storage. Destructive examinations of the sister rods are focused on understanding overall SNF rod strength and durability. Both composite fuel and empty cladding will be tested to derive material properties. While the data generated can be used for multiple purposes, a primary goal of the sister rod PIE data and associated measured mechanical properties is to support existing fuel storage licensing and relicensing activities by addressing identified knowledge gaps, and enhance the technical basis for post-storage transportation, handling, and consolidation activities. Fatigue tests of HBU M5-clad, ZIRLO-clad, Zircaloy-4, and low tin Zircaloy-4 segments have been completed using ORNL’s Cyclic Integral Reversible Fatigue Tester (CIRFT). This is the first time ZIRLO-clad rods have been tested. Some of the segments were heat-treated to understand the effects of thermal transients such as those encountered during dry storage vacuum drying on the fatigue lifetime. This paper presents results of the fatigue tests obtained to date and discusses planned future tests.

Montgomery, Rose↗
DOE OSTI2024

Equity Considerations in Siting Consolidated Interim Storage Facilities for Commercial Spent Nuclear Fuel

Questions of equity and fairness are often raised by representatives from State, Tribal and local governments and members of the public when resolving conflicts over siting of nuclear facilities. However, the two terms are not synonymous: equity can be quantified while fairness is much more subjective. The Department of Energy's (DOE) December 2021 Request for Information sought input on how to site federal facilities for the temporary, consolidated storage of spent nuclear fuel (SNF) using a consent-based approach. It is anticipated that just such questions about equity and fairness associated with the potential emergence of more than one host site may be raised by the twelve groups of university, nonprofit and private sector partners (Consortia) working to support community engagement, inclusively involve stakeholders, build relationships, and develop innovative forms of mutual learning and public capacity to participate in the consent-based siting process for a federal consolidated interim storage facility during the current Planning & Capacity Building stage of the consent-based siting process. In terms of DOE's Integrated Waste Management System (IWMS), sociopolitical equity concepts are a factor in all aspects of IWMS development and operations and will be particularly applicable should two or more host sites become a possibility, either by the emergence of multiple volunteer hosts or by pursuit as a IWMS program strategy. This paper summarizes recent analyses that explore how four equity metrics - number of states, division of the projected SNF total by the year 2083, current population and land area - might be perceived by host communities, states and regions if two, three or four sites for hosting federal consolidated interim storage facilities were to be contemporaneously realized. Existing institutional arrangements, including Nuclear Regulatory Commission regions, Federal Energy Regulatory Commission regions and Low-Level Radioactive Waste Disposal Compacts, were selected to create hypothetically merged two, three and four-region scenarios to calculate how the four metrics balance out. Projected SNF burden was prioritized with regional contiguity of the consolidated regions a requirement for a functional scenario. The intent was not to promote or suggest any construct as a program objective; the configurations are geopolitical abstractions only to explore potential perceptions of equity. Understanding the possible issues of equity that could arise may benefit program efforts toward achieving a cooperative federalism wherein both the federal and multiple State governments share the goal of manifesting more than one federal consolidated interim storage facility. The main finding is that it is not possible to optimize for all four metrics simultaneously. The analyses demonstrate that trying to create a sense of equity by backfitting a solution to a random population and land distribution will always contain a degree of artificiality.

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

Shielding an Infrasound Sensor for INL’s Fuel Conditioning Facility

Idaho National Laboratory (INL) has been pyroprocessing spent fuel from the Experimental Breeder Reactor II (EBR-II). This process occurs in the Fuel Conditioning Facility (FCF) which houses the treatment of DOE-owned sodium-bonded metal fuel. The electrometallurgical methods used in FCF treat the spent fuel from EBR-II for recycling of the uranium. The radiation field present inside FCF throughout the pyroprocessing stages affects any electronic sensors and devices that may be used for research. An example of such device is a seismoacoustic sensor which detects sound and vibration waves from the equipment in the facility. Seismoacoustic sensors are of interest to research nonproliferation techniques.

46 - INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AN↗
DOE OSTI2023

Residual Stresses in Spent Nuclear Fuel Storage Canisters with Cold Spray Coatings

This report describes the development of cold spray (CS) process and its application to provide on overlay coating of metal welds with a surface crack in a spent nuclear fuel (SNF) storage canister and thereby remediate the canister from further propagation of the crack. This work informs aging management programs for the SNF canisters.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

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

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

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Update on the Simulation of Commercial Drying of Spent Nuclear Fuel

The purpose of this report is to document improvements in the simulation of commercial vacuum drying procedures at the Nuclear Energy Work Complex at Sandia National Laboratories. Validation of the extent of water removal in a dry spent nuclear fuel storage system based on drying procedures used at nuclear power plants is needed to close existing technical gaps. Operational conditions leading to incomplete drying may have potential impacts on the fuel, cladding, and other components in the system. A general lack of data suitable for model validation of commercial nuclear canister drying processes necessitates additional, well-designed investigations of drying process efficacy and water retention. Scaled tests that incorporate relevant physics and well-controlled boundary conditions are essential to provide insight and guidance to the simulation of prototypic systems undergoing drying processes.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Large Plate Experiment of Chloride-induced Stress Corrosion Cracking in Spent Nuclear Fuel Storage Canisters

This report is the comprehensive description of the completed experiment to evaluate the corrosion behavior of a large section of welded plate material harvested from the Sandia National Laboratories’ (SNL) full-size mockup of a spent nuclear fuel (SNF) dry storage canister. The plate was subjected to conditions aggressive to chloride-induced stress corrosion cracking for a total time of 2 years. The details of the establishment of this experiment, and the corrosion results are reported. This work is intended to inform aging management programs for SNF canisters under conditions of extended dry storage.

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

Past Approaches for Spent Nuclear Fuel, Transuranic, and High-Level Waste Disposal in the United States—Part 2: Siting Process, Staged Development, and Public Preferences

This report presents pertinent aspects of the ~50-year United States experience in siting a mined geologic disposal repository for spent nuclear fuel (SNF), transuranic (TRU) waste, and high-level radioactive waste (HLW) as related to site selection and the staged process for site investigations as specified in the Nuclear Waste Policy Act of 1982 and generic and site-specific regulations of the US Department of Energy (DOE), US Environmental Protection Agency (EPA), and US Nuclear Regulatory Commission (NRC). The roles of the Environmental Impact Statement and guidance in international consensus standards by the International Atomic Energy Agency are also mentioned. The focus is on siting and developing the Waste Isolation Pilot Plant, an operating repository for TRU waste from atomic energy defense activities, and the proposed Yucca Mountain repository for commercial SNF and HLW. In the social dimension, the role of institutional stakeholders is described. Past national surveys related to waste management options for storage and disposal provide insight on public preferences of other stakeholders. The descriptions are intended to help other countries more fully understand the stages adopted for siting and developing repositories in the United States.

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

Corrosion-Resistant Coatings for Mitigation and Repair of Spent Nuclear Fuel Dry Storage Canisters

This report summarizes the results of a literature survey on coatings and surface treatments that are used to provide corrosion protection for exposed metal surfaces. The coatings are discussed in the context of being used on stainless steel spent nuclear fuel (SNF) dry storage canisters for potential prevention or repair of corrosion and stress corrosion cracking. The report summarizes the properties of different coating classes, including the mechanisms of protection, their physical properties, and modes of degradation (thermal, chemical, radiological). Also discussed are the current standard technologies for application of the coatings, including necessary surface pretreatments (degreasing, rust removal, grinding) and their effects on coating adhesion and performance. The coatings are also classified according their possible use for in situ repair; ex situ repair, requiring removal from the overpack; and ex situ prevention, or application prior to fuel loading to provide corrosion protection over the lifetime of the canister.

36 MATERIALS SCIENCE↗
DOE OSTI2025

DEVELOPMENT OF AN INTEGRATED SECURITY AND SAFETY MONITORING SYSTEM FOR SPENT NUCLEAR FUEL AND HIGH-LEVEL WASTE TRANSPORTATION

This paper provides an overview of the progress to date, and discussion of the path forward, related to designing, fabricating, and testing an integrated security and safety monitoring system (ISSMS) for railcars used to transport spent nuclear fuel (SNF) and high-level radioactive waste (HLW) in the United States. The system will comply with the US Department of Energy’s (DOE) Order 460.2B “Departmental Materials Transportation Management” and the Association of American Railroads’ (AAR) standard S-2043 “Performance Specification for Trains Used to Carry High-level Radioactive Material” [1] developed specifically for railcars used to transport high-level radioactive material (HLRM). DOE is in the process of developing and testing an ISSMS that will satisfy both DOE requirements and AAR standards. DOE has already developed railcar designs for transportation of HLRM. In 2024, DOE’s Atlas railcar project completed the design, fabrication and testing of three railcar types: transportation cask-carrying, buffer, and security escort, resulting in AAR conditional approval to operate on freight rail networks in North America. DOE decided to combine the required security and safety systems into one system, and this combined system is the subject of the current effort. DOE is developing and proposes to implement the ISSMS for these railcars as part of the build out of the railcar fleet. DOE began planning for development of the ISSMS in February 2020. The project is divided into seven phases starting with conceptual design and continuing through production design, as shown in Figure 1. An earlier version of the system was tested as part of the Atlas railcar consist demonstration test run in 2023. This paper describes the design features and system testing using the Atlas project railcars, and laboratory testing completed to date. The paper will also describe the activities planned to support the DOE project to ship the High Burn-Up Research Cask (HBRC) in 2027.

Schultze, Michael [ORNL] (ORCID:0000000283205671)↗
DOE OSTI2020

Large Plate Experiment of Chloride-induced Stress Corrosion Cracking in Spent Nuclear Fuel Storage Canisters

This report provides a description of the testing (in progress at the Savannah River National Laboratory) to evaluate the corrosion behavior of a large section of welded plate material harvested from the Sandia National Laboratories’ full-size mockup of a spent nuclear fuel (SNF) dry storage canister. The details of the establishment of this plate test, the interim corrosion results (from visual examination), and the general plan for examination of the plate in CY21 to evaluate crack growth behavior (if any) are provided. This work is intended to inform aging management programs for SNF canisters under conditions of extended dry storage. A “large plate” (approximately 51 x 46 cm with thickness 16 mm and weight about 30 kg) and containing a circumferential weldment from the mockup is the test specimen. The knowledge of the salt, temperature, humidity conditions for chloride-induced stress corrosion cracking (CISCC) including the threshold stress intensity factor, K iscc , were used to setup the large plate test in which a set of part-through-wall and through-wall electrical discharge machined (EDM) defects were positioned in the weld residual stress field of the welded plate material from the mockup canister. A marine salt solution was applied to this plate specimen with an air-brush spray, then dried, and then exposed to a sustained 71-73% RH condition at an ambient room temperature of 22 °C. The salt loading and environmental condition is expected to be aggressive to cause CISCC

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗