Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “Nuclear Facility”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 775 records · Page 43

A Framework to Assess Advanced Reactor Spent Fuel Management Facility Deployment

Advanced nuclear reactors offer various operational advantages over existing light water reactors but could produce types of spent nuclear fuel (SNF) with a wide variety of forms and characteristics depending on how many different concepts are deployed. Each advanced reactor SNF type potentially poses unique management challenges. New planning efforts will be necessary to anticipate how the management requirements of advanced reactor SNF will affect the deployment of an integrated waste management system. This paper applies a framework of high-level facility deployment milestones to a generic SNF management system, reviewing them together with the advanced reactor SNF characteristics and management requirements. This allows for the investigation of factors that influence facility and system deployment, and ultimately, the identification of challenges facing the deployment of different kinds of SNF management facilities. Here, the back end of the once-through fuel cycle is examined for four advanced reactor system technology types: sodium-cooled fast reactors, high-temperature gas-cooled reactors, liquid-fuel molten salt reactors, and lead-cooled fast reactors. It is observed that milestones earlier in the facility deployment process (e.g., siting and facility design) are more impacted by the uniqueness of advanced reactor SNF characteristics than others (e.g., construction and testing). Ultimately, none of the differences are seen as fundamentally disqualifying in a technical sense; however, they should be considered early, potentially as part of reactor design, to avoid issues in the future.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

APS Science 2020 (Vol. 2)

The U.S. Department of Energy's Advanced Photn Source (APS) is one of the world’s most productive x-ray light source facilities. Each year, the APS provides high-brightness x-ray beams to a diverse community of more than 5,000 researchers in materials science, chemistry, condensed matter physics, the life and environmental sciences, and applied research. Researchers using the APS produce over 2,000 publications each year detailing impactful discoveries, and solve more vital biological protein structures than users of any other x-ray light source research facility. APS x-rays are ideally suited for explorations of materials and biological structures; elemental distribution; chemical, magnetic, electronic states; and a wide range of technologically important engineering systems from batteries to fuel injector sprays, all of which are the foundations of our nation’s economic, technological, and physical well-being. The APS occupies an 80-acre site on the Argonne campus, about 25 miles from downtown Chicago, Illinois. It shares a site with the Center for Nanoscale Materials and the Advanced Protein Characterization Facility.

01 COAL, LIGNITE, AND PEAT↗

Long-term sinking of nuclear waste canisters in salt formations by low-stress creep at high temperature

Abstract Rock salt has a self-sealing capacity, low permeability, and high thermal conductivity, making it a potential host for heat-generating nuclear waste. The feasibility of nuclear waste disposal within salt formations has been investigated mostly for small-sized canisters. Geologic disposal of larger-sized canisters originally designed for spent fuel storage and transportation has lately been examined as a cost-effective alternative. This raises questions about their long-term vertical movement due to their weight and high decay heat. Low-stress creep governs this movement; however, most salt constitutive models do not incorporate it. In this paper, the Norton and the WIPP creep models are compared with the Lux/Wolters/Lerche (LWL) model and a simpler model that combines linear and Norton creep laws (named combined creep model). The LWL and combined creep models consider pressure solution creep, though all incorporate dislocation creep. The models are first applied to creep tests under various stress levels. The LWL and the combined creep models results fit the experimental data well in both high and low stress ranges, whereas the Norton and WIPP models results only fit in higher-stress ranges. The different models are further applied for analyzing long-term canister movement. A sinking rate of $$-4.4\times 10^{-7}$$ - 4.4 × 10 - 7 mm/year was predicted using the Norton and WIPP models versus $$-2.1\times 10^{-2}$$ - 2.1 × 10 - 2 mm/year and $$-3.1\times 10^{-2}$$ - 3.1 × 10 - 2 mm/year using the LWL and the combined creep models, respectively. This comparative study confirms that creep models calibrated exclusively against high-deviatoric stress data might result in an inaccurate estimation of waste packages sinking rate in salt formations.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Application of machine learning for modeling brønsted-guggenheim-scatchard specific ion interaction theory (SIT) coefficients

Machine learning methodologies can provide insight into Brønsted-Guggenheim-Scatchard specific ion interaction theory (SIT) parameter values where experimental data availability may be limited. This study develops and executes machine learning frameworks to model the SIT interaction coefficient, ε. Key findings include successful estimations of ε via artificial neural networks using clustering and value prediction approaches. Additionally, applicability to other chemical parameters is also assessed briefly. Models developed here provide support for a use-case of machine learning in geologic nuclear waste disposal research applications, namely in predictions of chemical behaviors of high ionic strength solutions (i.e., subsurface brines).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Multicomponent pyrochlore solid solutions with uranium incorporation – A new perspective of materials design for nuclear applications

Multicomponent pyrochlore solid solutions with and without uranium incorporation were fabricated and their thermal-mechanical properties were characterized. Multicomponent pyrochlore solid solutions without uranium exhibit comparable thermal conductivity and higher mechanical strength compared to baseline single component rare-earth titanate pyrochlore (A 2 Ti 2 O 7 ). Uranium incorporation reduces hardness as compared with single component compositions. High entropy pyrochlore with uranium displays the highest thermal conductivity within multicomponent pyrochlore solid solutions with significantly better mechanical properties than UO 2 . The measured thermal conductivity correlates well with A-site cation mixing entropy and a modified size disorder parameter, and thus the size disorder and mixing entropy could be good indicators for predicting thermal conductivity of multicomponent pyrochlore solid solutions. This work opens up the possibility of designing multicomponent oxide solid solutions by controlling their chemical disorder/mixing entropy to achieve acceptable thermal-mechanical properties, desired radiation and corrosion performance for potential nuclear waste form and inert matrix fuel applications.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Conversion degree and heat transfer in the cold cap and their effect on glass production rate in an electric melter

A predictive model of melt rate in waste glass vitrification operations is needed to inform melter operations during normal and off-normal operations. This paper describes the development of a model of the cold cap (the reacting melter feed floating on molten glass in a glass melter) that couples heat transfer with the feed-to-glass conversion kinetics. The model was applied to four melter feeds designed for high-level and low-activity nuclear waste feeds using the material properties, either measured or estimated, to obtain temperature and conversion distribution within the cold cap. The cold cap model, when coupled with a computational fluid dynamics model of a Joule-heated glass melter, allows the prediction of the glass production rate and power consumption. The results show reasonable agreement with the melting rates measured during pilot-scale melter tests.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Report on Field Test at INL Cask Farm of Single Detector Fast Neutron Spent Fuel Cask Verification System

Detecting diversion of spent fuel elements in dry storage casks is challenging due to the thick shielding used in cask construction. Measurements on top of the cask to map the underlying arrangement of the fuel elements and looking for anomalous changes over time has proven difficult to achieve using gamma rays due to the high scattering and attenuation from the thick steel structure, weakening information on the present or absence of fuel bundles. Simulations and laboratory experiments suggest that the high-energy neutron flux (>200 keV) measured directly above each fuel bundle is sufficient to produce a position map that enables detection of the present or absence of fuel bundles, and therefore diversion of a spent fuel bundle. A single-detector spent-fuel monitoring technique based on this principle was development at the Lawrence Livermore National Laboratory (LLNL). The INL Cask Farm in the INTEC technical area at Idaho National Laboratory (INL) offers the capability to test this technique on an MC-10 storage cask which has a distribution of full and empty fuel positions. An experimental test plan for the single-detector verification system was developed in consultation with INL personnel to be completed in FY2021. Due to travel advisories related to COVID-19, the experimental test plan was adapted to enable INL personnel to carry out the measurements in consultation with LLNL personnel following shipment of the LLNL system to INL. Field test measurements of the single detector verification system were successfully carried out at the INL cask farm on September 7-9, 2021. This document summarizes results of the field test.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

FY22 Laboratory Directed Research and Development Annual Report

The Laboratory Directed Research and Development (LDRD) program yields foundational scientific research and development (R&D) essential to growing SRNL’s core competencies, in alignment with SRNL’s Strategic Plan to provide long-term benefits to the Department of Energy (DOE), the National Nuclear Security Administration (NNSA), and other customers and stakeholders. Five strategic goals are outlined in SRNL’s strategic plan: 1) Provide applied science and engineering for EM’s active clean-up sites and LM’s post closure management sites; 2) Provide science-based solutions for gaps identified in nonproliferation strategic vision and support the government in actives impacting national security; 3) Lead ST&E as the central technical authority for processing tritium loaded reservoirs and support production of plutonium pits; 4) Align science and energy security programs by focusing modern modeling, simulation, and data analytics tools on materials engineering and performance applications; 5) Build a workforce for the future.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Initial Site-Specific De-Inventory Report for Yankee Rowe

This report addresses the tasks, equipment, and interfaces necessary for the complete de-inventory of the Yankee Rowe Nuclear Power Station (YR) independent spent fuel storage installation (ISFSI) site.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Initial Site-Specific De-Inventory Report for Zion

This report addresses the tasks, equipment, and interfaces necessary for the complete de-inventory of the Zion Nuclear Power Station (Zion) independent spent fuel storage installation (ISFSI) site.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Long-term Moisture Adsorption in Packaged Plutonium Oxide

In 2018, the Department of Energy National Nuclear Security Administration (DOE-NNSA) began implementing dilute and dispose to remove 34 metric tons (MT) of surplus weapons grade plutonium from the US stockpile. Under this plan, surplus plutonium material is converted into plutonium oxide (PuO 2 ) before being stored in metal containers and sent to the DOE Waste Isolation Pilot Plant (WIPP). The dilute and dispose project was implemented as a more cost-effective method for abiding by the Plutonium Management and Disposition Agreement (PMDA) between the USA and Russia, as compared to producing mixed oxide fuel (MOX). The PMDA was originally signed in 2000 and amended in 2010.2 The disassembly of pits and conversion to PuO 2 as part of dilute and dispose is carried out through the Advanced Recovery and Integrated Extraction System (ARIES) developed at Los Alamos National Laboratory (LANL). The dilute portion of dilute and dispose is carried out at Savannah River Site (SRS). With this program in place, it is necessary to ensure the safe, long-term storage of the PuO 2 generated during this process until final disposal at WIPP is accomplished.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Panel Session 14: US NRC - Current and Emerging US NRC Regulatory Topics

This panel focused on several current and emerging US Nuclear Regulatory Commission (NRC) regulatory issues impacting the commercial radioactive waste arena in the US. The panelists provided different perspectives on key issues such as the revision of 10 Code of Federal Regulations Part 61 'Licensing Requirements For Land Disposal Of Radioactive Waste', the scoping study on very low-level waste, the consolidated interim storage of spent nuclear fuel and other timely regulatory topics. Panelists with presentations: Overview of NRC Low-Level Radioactive Waste Activities and Initiatives (Patricia Holahan); US Nuclear Power Plant Waste Management - US LLW Disposal Issues (Joe Weismann); Decommissioning Rule (Vern C. Rogers); What is TRU? Definition, Legislation, and Rulemaking (David Carlson); LLRW Regulatory Topics of Interest in Texas (Bobby Janecka); International Low Level Waste Management Practices (Richard McGrath)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Optimization of Dry Storage Canister Cutting Operations - 20534

A recent project investigated the optimal cutting method for the opening of a welded dry storage/shielded canister (DSC) or dual purpose canister (DPC) containing Used Nuclear Fuel (UNF). Due to the lack of a disposal path for UNF in the U.S., the majority of UNF is currently moved into welded DSCs and DPCs designed by Orano TN, NAC, and Holtec. As the DSCs/DPCs were neither designed nor licensed for disposal and may not be able to be emplaced in a geologic repository due to physical emplacement constraints, near-term thermal limitations, or long-term criticality issues, the UNF in these existing DSCs/DPCs may need to be repackaged into transportation, aging, and disposal canisters (TADs), generic standard TADs (STADs), transportation casks, new cask/canister systems, and/or cask/canister/package systems suitable for disposal. These DSCs/DPCs may also be opened to simply remove the UNF in preparation for recycling, re-storage, or placement into a waste package suitable for disposal. Regardless of need, the DSCs/DPCs would require to be cut open by a potentially large scale operation. This project examined the multiple means for cutting welded metal systems and considered several factors to determine an optimal approach. Issues arose around some of the cutting approaches when applied to systems with 2 or 3 lids and with pre-cutting gas testing required. The resulting preferences in cutting methods with an optimized application for the opening of DSCs/DPCs are presented. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Nuclear waste reduction: Exploring new pathways one step at a time

In my home country of Venezuela, nuclear energy is not a topic that attracts much attention. The government briefly oversaw some nuclear energy programs during the 1950s, but currently there are no active nuclear power facilities in the country. In fact, the Venezuelan government signed and ratified the treaty of the prohibition of nuclear weapons in 2021, which states that Venezuela has never owned, possessed or controlled nuclear weapons or programs of any kind. When I moved to the United States, however, nuclear energy became an extremely relevant topic. In the 1940s, the U.S. government established and oversaw the Manhattan Project to build atomic bombs for use in World War II. After the war, the government encouraged scientists to use this information on nuclear reactions to develop nuclear energy for peaceful civilian purposes instead.1 During these early days of nuclear research, there were no formal regulatory standards for nuclear waste management. Policies usually were self-regulated and often created based on existing policies of disposal for non-nuclear waste.2 As a result, there were instances of nuclear waste leaching into the environment and affecting local communities. So, much research has been conducted since then to characterize and store nuclear waste safely and securely.3 I first became interested in nuclear energy during my undergraduate studies when I worked on a project involving ligand synthesis to help extract actinides from nuclear waste. I then studied electrochemistry in molten salt systems for nuclear energy applications during my Ph.D. As I approached graduation, I started looking into national laboratories that have programs involving nuclear energy and waste management. At Idaho National Laboratory (INL), the focus is more on applied processes and how nuclear energy can be innovated to realize next-generation reactor design and technologies. This focus led me to apply for a Seaborg distinguished postdoctoral position at INL, for which I was chosen based on my proposal of a way to improve nuclear waste recycling. To understand my proposal, we must familiarize ourselves with the makeup of nuclear waste. After uranium dioxide is used as nuclear fuel in a reactor, the fuel matrix is then characterized by various fission products, including rare earth elements, alkali and alkaline earths, and actinides. Some of these fission products can potentially be recovered through pyroprocessing, 4 which involves the electrochemical dissolution of the used nuclear fuel in a molten chloride salt mixture at high temperatures. Though some of the fission products can be easily recovered—for example, uranium is reduced onto an inert cathode by applied potentials—numerous other fission products such as rare earth elements are difficult to recover due to their multivalent oxidation states and side reactions.5 To improve the recovery efficiency of rare earth elements specifically, I proposed investigating the fundamental interactions between rare earth elements in the molten chloride salt and their metallic form (Figure 1). The kinetic pathways and the chemical reactions of these elements, which will be elucidated through spectro-electrochemistry at high temperatures, will give insights on how the recovery efficiency can be improved. Although my research focuses on fundamental science, it will benefit the applied process by generating new scientific knowledge and closing the gap for efficient recycling of the waste: one step at a time.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Segmented Gamma Scanner for Radioactive Waste Assessment: A User Guide

Radioactive waste assessment is important for ensuring nuclear material security at various types of facilities, such as enrichment, fuel fabrication, and reprocessing plants. The waste generated at such nuclear facilities is stored in standard containers and is required to be characterized for material-accounting purposes. The segmented gamma scanner system is a popular, nondestructive analysis measurement system used for characterizing nuclear material, including radioactive waste. This document provides guidance on how to achieve effective performance from a segmented gamma scanner system for accurately quantifying fission products, activation products, and transuranic wastes.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗