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Surface Heater Fabrication Using Micro-Lithography for Transpiration Cooling Heat Transfer Coefficient Measurements

Abstract Transpiration cooling is able to provide more uniform coolant coverage than film cooling to effectively protect the component surface from contacting the hot gas. Due to numerous coolant ejection outlets within a small area at the target surface, the experimental thermo-fluid investigation on transpiration cooling becomes a significant challenge. Two classic methods to investigate film cooling, the steady-state foil heater method and the transient thermography technique, fail for transpiration cooling because the foil heater would block numerous coolant outlets, and the semi-infinite solid conduction model no longer holds for porous plates. In this study, a micro-lithography method to fabricate a silver coil pattern on top of the additively manufactured polymer porous media as the surface heater was proposed. The circuit was deliberately designed to cover the solid surface in a combination of series connection, and parallel connection to ensure the power in each unit cell area at the target surface was identical. With uniform heat flux generation, the steady-state tests were conducted to obtain distributions of a pair of parameters, adiabatic cooling effectiveness, and heat transfer coefficient (HTC). The results showed that the adiabatic cooling effectiveness could reach 0.65 with a blowing ratio lower than 0.5. Meanwhile, the heat transfer coefficient ratio (hf/h0) of transpiration cooling was close to 1 with a small blowing ratio at 0.125. A higher HTC ratio was observed for smaller pitch-to-diameter cases due to more turbulence intensity generated at the target surface.

Engineering↗

Designing Cladded Components for High Temperature Nuclear Service: Part II—Design Rules

The challenge of using existing ASME Section III, Division 5, class A metallic materials for the construction of structural components of advanced reactors with corrosive coolants could be mitigated by allowing designers to use cladding to protect the base material from corrosion. However, the existing Section III, Division 5 rules provide no guidance on the evaluation of strain accumulation and creep-fatigue damage in cladded components. The availability of design rules for cladded components that do not require long-term clad material testing could promote the application of the cladding approach to accelerate the deployment schedule of these advanced reactor systems. To avoid long-term properties for the clad materials Part I of this work proposes two approximate design analysis methods for two types of clad materials—soft clads that creep much faster and have lower yield stress than the class A base material, and hard clads that creep much slower and have higher yield stress than the class A base material. The proposed analysis methods approximate the response of a soft clad by treating it as perfectly compliant and of a hard clad by treating it as linear elastic. Based on these approximate design analysis strategies this Part II develops a complete set of design rules for class A components cladded with either soft or hard clad materials. In conclusion, Part II discusses the reasoning behind the proposed design rules and uses example finite element analyses of representative reactor components to illustrate the use of these design methods.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Pathways for Disposal of Commercially-Generated Tritiated Waste

From a waste disposal standpoint, tritium is a major challenge. Because it behaves like hydrogen, tritium exchanges readily with hydrogen in the ground water and moves easily through the ground. Land disposal sites must control the tritium activity and mobility of incoming wastes to protect human health and the environment. Consequently, disposal of tritiated low-level wastes is highly regulated and disposal options are limited. The United States has had eight operating commercial facilities licensed for low-level radioactive waste disposal, only four of which are currently receiving waste. Each of these is licensed and regulated by its state. Only two of these sites accept waste from states outside of their specified regional compact. For waste streams that cannot be disposed directly at one of the four active commercial low-level waste disposal facilities, processing facilities offer various forms of tritiated low-level waste processing and treatment, and then transport and dispose of the residuals at a disposal facility. These processing facilities may remove and recycle tritium, reduce waste volume, solidify liquid waste, remove hazardous constituents, or perform a number of additional treatments. Waste brokers also offer many low-level and mixed waste management and transportation services. These services can be especially helpful for small-quantity tritiated-waste generators, such as universities, research institutions, medical facilities, and some industries. The information contained in this report covers general capabilities and requirements for the various disposal/processing facilities and brokerage companies, but is not considered exhaustive. Typically, each facility has extensive waste acceptance criteria and will require a generator to thoroughly characterize their wastes. Then a contractual agreement between the waste generator and the disposal/processing/broker entity must be in place before waste is accepted. Costs for tritiated waste transportation, processing and disposal vary based a number of factors. In many cases, wastes with very low radioactivity are priced primarily based on weight or volume. For higher activities, costs are based on both volume and activity, with the activity-based charges usually being much larger than volume-based charges. Other factors affecting cost include location, waste classification and form, other hazards in the waste, etc. Costs may be based on general guidelines used by an individual disposal or processing site, but final costs are established by specific contract with each generator. For this report, seven hypothetical waste streams intended to represent commercially-generated tritiated waste were defined in order to calculate comparative costs. Ballpark costs for disposition of these hypothetical waste streams were calculated. These costs ranged from thousands to millions of dollars. Due to the complexity of the cost-determining factors mentioned above, the costs calculated in this report should be understood to represent very rough cost estimates for the various hypothetical wastes. Actual costs could be higher or could be lower due to quantity discounts or other factors.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Proposed Disposition of Large-Scale Artifacts Associated with Idaho National Laboratory's Nuclear Reactor Testing Historic Context

As an active scientific and engineering laboratory with a historic mission of testing and development of nuclear power, INL presents unique challenges to historic preservation. These challenges include radiologically contaminated buildings and equipment, security restrictions, and the nearly constant modification, demolition, and replacement of structures and equipment to meet changing programmatic and mission needs. Because of these challenges, what began in 1966 as a building-by-building approach to historic preservation of the INL built environment has evolved into a more holistic management strategy and systematic procedures for identifying, evaluating, and protecting important properties within a historic contextual framework. The development of contexts, identification of historic themes, and inventories of historic INL architectural properties have been expanded following the strategies and procedures outlined within the INL Cultural Resource Management Plan (CRMP). Inventories are ongoing to catalog other important INL architectural properties that may or may not be eligible to the National Register of Historic Places (NRHP), such as nuclear-era artifacts, and photographic and engineering collections.

99 GENERAL AND MISCELLANEOUS↗

The Effects of Manufacturing Techniques on Neutron-Irradiated C26M Mechanical Properties

Nuclear energy provides 13% of the total energy produced globally. To further improve the safety and reliability of nuclear reactors material challenges need to be successfully resolved. C26M/FeCrAl alloy systems can improve the safety and reliability of current Generation III reactors due to the potential benefits of a stable passivation layer from the Al, which can protect against corrosion, and promising mechanical performance at boiling- and light-water reactor temperatures. However, information is limited on how manufacturing techniques affect the mechanical performance of C26M in neutron-irradiated environments. In this research, C26M specimens were manufactured using three techniques and then irradiated at the Advanced Test Reactor at Idaho National Laboratory. This research aims to bridge this gap by investigating how the manufacturing techniques affects the mechanical performance of irradiated C26M at Generation III reactor relevant temperatures.

36 - MATERIALS SCIENCE↗

The post-2015 delivery of universal and sustainable access to infrastructure services. Working Paper

In this new working paper, the authors focus specifically on what would be necessary to achieve High Level Panel-style goals and targets for water, energy and transport, if these were to be eventually adopted by world leaders. In all three cases, much of the advocacy - and the proposed High Level Panel goals - have emphasized the need to strive for universal and sustainable access to at least basic levels of services from these sectors. Many of the proposals for post-2015 goals and targets appear ambitious, but what would it take to achieve them? This paper assesses what is needed to achieve goals for universal and sustainable access to infrastructure, specifically water, energy and transport. Using illustrative goals and targets, the paper reviews the development challenges in each sector, and what will be necessary to overcome the barriers to universal and sustainable access to water, energy and transport infrastructure services, in the areas of governance, finance, capacity development and environmental protection. The paper ends with general conclusions about infrastructure in the post-2015 development agenda.

29 ENERGY PLANNING, POLICY, AND ECONOMY↗

K-25 D and D Challenges - 12170

Decontamination and decommissioning (D and D) of the K-25 former gaseous diffusion plant provides lessons learned applicable to other D and D projects in the DOE Complex. The 175,000- square-meter, 1.6-kilometer-long building is contaminated with highly enriched uranium (HEU), Tc-99, trace quantities of other fission products, asbestos, PCBs, and other hazardous wastes. Safety challenges include deteriorated electrical systems, significant structural degradation, and criticality and exposure risks. The project completely revised the D and D approach after a worker fell through an operating floor and was seriously injured. For protection from deteriorated building conditions, the project reduced the number of workers in the facility, limited their hours in the building, and installed nets and barriers to protect them from falls through weakened floors and from falling material. The new plan involved removing high-risk components, removing motors and compressors, and demolishing the building from the outside with heavy equipment with most of the piping and components inside the building during demolition. The team provided temporary electrical power; reconfigured the criticality alarm system; upgraded security; performed sampling and analysis to locate and characterize HEU deposits and Tc-99; and, to establish 'criticality incredible' conditions, conducted a nondestructive assay program and injected foam into equipment and piping as a contamination fixative. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Multi-Phased, Post-Accident Support of the Fukushima Dai-Ichi Nuclear Power Plant - 12246

In the wake of the March 11 earthquake and tsunami and the subsequent flooding of several of the Fukushima Dai-Ichi reactors, Japan and the Japanese utility TEPCO faced a crisis situation with incredible challenges: substantial amounts of radioactive mixed seawater and freshwater accumulated in the basements of four reactor and other buildings on the site. This water held varying levels of contamination due to the fact that it had been in contact with damaged fuel elements in the cores and with other contaminated components. The overall water inventory was estimated at around 110,000 tons of water with contamination levels up to the order of 1 Ci/l. Time was of the essence to avoid overflow of this accumulated water into the ocean. AREVA proposed, designed and implemented a water treatment solution using a proven chemical coprecipitation process with ppFeNi reagent, which is currently in use for effluent treatment on several nuclear sites including AREVA sites. In addition to the extremely short schedule the other challenge was to adapt the chemical treatment process to the expected composition of the Fukushima water and, in particular, to evaluate the impact of salinity on process performance. It was also necessary to define operating conditions for the VEOLIA equipment that had been selected for implementation of the process in the future facility. The operation phase began on June 17, and by the end of July more than 30,000 tons of highly radioactive saltwater had been decontaminated - the Decontamination Factor (DF) for Cesium was ∼10{sup 4}. It allowed recycling the contaminated water to cool the reactors while protecting workers and the environment. This paper focuses on the Actiflo{sup TM}-Rad water treatment unit project that was part of the TEPCO general water treatment scheme. It presents a detailed look at the principles of the Actiflo{sup TM}-Rad, related on-the-fly R and D, an explanation of system implementation challenges, and a brief summary of operation results to date. AREVA's response to the Fukushima Dai-Ichi crisis was multi-phased: emergency aid and relief supply was sent within days after the accident; AREVA-Veolia engineering teams designed and implemented a water treatment solution in record time, only 3 months; and AREVA continues to support TEPCO and propose solutions for waste management, soil remediation and decontamination of the Fukushima Dai-Ichi site. Despite the huge challenges, the Actiflo{sup TM}-Rad project has been a success: the water treatment unit started on time and performed as expected. The performance is the result of many key elements: AREVA expertise in radioactive effluents decontamination, Veolia know-how in water treatment equipments in crisis environment, and of course AREVA and Veolia teams' creativity. The project success is also due to AREVA and Veolia teams' reactivity and high level of commitment with engineering teams working 24/7 in Japan, France and Germany. AREVA and Veolia deep knowledge of the Japanese industry ensured that the multi-cultural exchanges were not an issue. Finally the excellent overall project management and execution by TEPCO and other Japanese stakeholders was very efficient. The emergency water treatment was a key step of the roadmap towards restoration from the accident at Fukushima Dai-Ichi that TEPCO designed and keeps executing with success. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

ARPA-E Program: Advanced Management Protection of Energy Storage Devices (AMPED) - Fifth Quarterly Project Report - FY14 Q1

Technology has been developed that enables monitoring of individual cells in high - capacity lithium-ion battery packs, with a distributed array of wireless Bluetooth 4.0 tags and sensors, and without proliferation of extensive wiring harnesses. Given the safety challenges facing lithium-ion batteries in electric vehicle, civilian aviation and defense applications, these wireless sensors may be particularly important to these emerging markets. These wireless sensors will enhance the performance, reliability and safety of such energy storage systems. Specific accomplishments to date include, but are not limited to: (1) the development of wireless tags using Bluetooth 4.0 standard to monitor a large array of sensors in battery pack; (2) sensor suites enabling the simultaneous monitoring of cell voltage, cell current, cell temperature, and package strain, indicative of swelling and increased internal pressure, (3) small receivers compatible with USB ports on portable computers; (4) software drivers and logging software; (5) a 7S2P battery simulator, enabling the safe development of wireless BMS hardware in the laboratory; (6) demonstrated data transmission out of metal enclosures, including battery box, with small variable aperture opening; (7) test data demonstrating the accurate and reliable operation of sensors, with transmission of terminal voltage, cell temperature and package strain at distances up to 110 feet; (8) quantification of the data transmission error as a function of distance, in both indoor and outdoor operation; (9) electromagnetic interference testing during operation with live, high -capacity battery management system at Yardney Technical Products; (10) demonstrat ed operation with live high-capacity lithium-ion battery pack during charge-discharge cycling; (11) development of special polymer-gel lithium-ion batteries with embedded temperature sensors, capable of measuring the core temperature of individual of the cells during charge-discharge cycling at various temperatures, thereby enabling earlier warning of thermal runaway than possible with external sensors. Ultimately, the team plans to extend this work to include: (12) flexible wireless controllers, also using Bluetooth 4.0 standard, essential for balancing large-scale battery packs. LLNL received $925K for this project, and has $191K remaining after accomplishing these objectives.

25 ENERGY STORAGE↗

US DOE-EM On-Site Disposal Cell Working Group - Fostering Communication On Performance Assessment Challenges

On-site disposal cells are in use and being considered at several U.S. Department of Energy (USDOE) sites as the final disposition for large amounts of waste associated with cleanup of contaminated areas and facilities. These facilities are typically developed with regulatory oversight from States and/or the US Environmental Protection Agency (USEPA) in addition to USDOE. The facilities are developed to meet design standards for disposal of hazardous waste as well as the USDOE performance based standards for disposal of radioactive waste. The involvement of multiple and different regulators for facilities across separate sites has resulted in some differences in expectations for performance assessments and risk assessments (PA/RA) that are developed for the disposal facilities. The USDOE-EM Office of Site Restoration formed a working group to foster improved communication and sharing of information for personnel associated with these Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) disposal cells and work towards more consistent assumptions, as appropriate, for technical and policy considerations related to performance and risk assessments in support of a Record of Decision and Disposal Authorization Statement. The working group holds teleconferences, as needed, focusing on specific topics of interest. The topics addressed to date include an assessment of the assumptions used for performance assessments and risk assessments (PA/RAs) for on-site disposal cells, requirements and assumptions related to assessment of inadvertent intrusion, DOE Manual 435.1-1 requirements, and approaches for consideration of the long-term performance of liners and covers in the context of PAs. The working group has improved communication among the staff and oversight personnel responsible for onsite disposal cells and has provided a forum to identify and resolve common concerns.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

The GIF Proliferation Resistance and Physical Protection Working Group (PRPPWG): Achievements and Perspectives

The Generation IV International Forum (GIF) “Technology Roadmap” identified proliferation resistance and physical protection (PR&PP) as one of the four goal areas to advance nuclear energy into its next, “fourth” generation. It recommended the development of a methodology to define measures for PR&PP and to evaluate them for the six nuclear energy systems selected by GIF. Accordingly, the Forum formed a PR&PP Working Group (PRPPWG) to develop a methodology and to define the parameters against which PR&PP would be measured. The PRPPWG developed the methodology through a series of development and demonstration case studies, by use of a hypothetical “Example Sodium Fast Reactor” (ESFR). The PR&PP ESFR assessment was the first opportunity to exercise the full methodology on a complete system, and many insights were gained from the process. The PRPPWG and representatives of the GIF System Steering Committees (SSCs) for each of the six GIF design concepts carried out a study on the PR&PP aspects of the six GIF designs. This interaction allowed the PRPPWG to engage the designers and to raise awareness about PR&PP. A number of international workshops have also been held which have introduced the methodology to design groups and other stakeholders. In this paper, we first summarize the technical progress and the major accomplishments of the PRPPWG: the evaluation methodology and the ESFR case study used to develop and test the methodology. The interaction with the GIF System Steering Committees and the relaunching of the activity is then presented. Finally, an outline of the future challenges is provided along the lines identified in the GIF R&D outlook 2018 report.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗

Conformal Lithium Fluoride Protection Layer on Three-Dimensional Lithium by Nonhazardous Gaseous Reagent Freon

Research on lithium (Li) metal chemistry has been rapidly gaining momentum nowadays not only because of the appealing high theoretical capacity, but also its indispensable role in the next-generation Li–S and Li–air batteries. However, two root problems of Li metal, namely high reactivity and infinite relative volume change during cycling, bring about numerous other challenges that impede its practical applications. In the past, extensive studies have targeted these two root causes by either improving interfacial stability or constructing a stable host. However, efficient surface passivation on three-dimensional (3D) Li is still absent. Here in this paper, we develop a conformal LiF coating technique on Li surface with commercial Freon R134a as the reagent. In contrast to solid/liquid reagents, gaseous Freon exhibits not only nontoxicity and well-controlled reactivity, but also much better permeability that enables a uniform LiF coating even on 3D Li. By applying a LiF coating onto 3D layered Li-reduced graphene oxide (Li-rGO) electrodes, highly reduced side reactions and enhanced cycling stability without overpotential augment for over 200 cycles were proven in symmetric cells. Furthermore, Li–S cells with LiF protected Li-rGO exhibit significantly improved cyclability and Coulombic efficiency, while excellent rate capability (~800 mAh g -1 at 2 C) can still be retained.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Interfacial transmetallation synthesis of a platinadithiolene nanosheet as a potential 2D topological insulator

The construction of two-dimensional metal complex materials is fascinating because of the structural and functional diversity of these materials. Previously, we have reported the synthesis of electroconductive nickelladithiolene (NiDT) and palladadithiolene (PdDT) nanosheets using benzenehexathiol (BHT). Down the group from Ni, Pd to Pt, there is a distinct positive shift in the reduction potential; as a result, it becomes synthetically more challenging to stabilize Pt 2+ than to form metallic Pt(0) in the presence of BHT as a reducing agent. Herein, a novel synthetic strategy for the preparation of platinadithiolene nanosheet (PtDT) using a dibutyltin-protected BHT ligand is reported, leading to transmetallation in the presence of dioxygen. Both free-standing stacked sheets and atomic layer sheets were obtained and characterized by microscopic techniques such as AFM, SEM, and TEM. To study the morphology of the sheets and determine their charge neutrality, X-ray photoelectron (XP) and infrared (IR) spectroscopic techniques were used. Powder X-ray diffraction analysis of the multilayer PtDT indicates a half-way slipped hexagonal configuration in the P 3¯1 m space group. The band structure of this PtDT exhibits a band gap at the Fermi level, which is different from that of NiDT in the staggered configuration, and a Dirac gap, indicating the possibility of 2D topological insulation at room temperature. PtDT is insulating but chemically activated by oxidation with I 2 to increase the conductivity by more than 106 folds up to 0.39 S cm –1 . The MDT sheets exhibit electrocatalytic activity for the hydrogen evolution reaction, and the activity order is NiDT < PdDT < PtDT.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

TH-AB-206-02: Nuclear Medicine Theronostics: Wave of the Future; Pre-Clinical and Clinical Opportunities

In the past few decades, the field of nuclear medicine has made long strides with the continued advancement of related sciences and engineering and the availability of diagnostic and therapeutic radionuclides. Leveraging these advancements while combining the advantages of therapeutic and diagnostic radionuclides into one radiopharmaceutical has also created a new subfield “theranostics” in nuclear medicine that has the potential to further propel the field into the future. This session is composed of two talks; one focused on the physics principles of theranostics from properties of beta and alpha emitting radionuclides to dosimetric models and quantification; while the second describes preclinical and clinical applications of theranostics and discusses the challenges and opportunities of bringing them to the clinic. At the end of the session the listener should be able to identify: The different properties of beta and alpha emitting radionuclides Which radionuclides are selected for which nuclear medicine therapies and why How PET can be used to accurately quantify the uptake of tumor targeting molecules How individualized dosimetry can be performed from the management of thyroid cancer to novel radiolabeled antibody therapies Promising pre-clinical radiopharmaceutical pairs in prostate cancer and melanoma. Promising clinical Theranostics in neuroendocrine cancers. Challenges of bringing Theranostics to the clinic. E. Delpassand, RITA Foundation -Houston; SBIR Grant; CEO and share holder of RadioMedix.

61 RADIATION PROTECTION AND DOSIMETRY↗

Karst characterization in a semi-arid region using gravity, seismic, and resistivity geophysical techniques.

We proposed to customize emerging in situ geophysical monitoring technology to generate time-series data during sporadic rain events in a semi-arid region. Electrodes were to be connected to wireless \nodes" which can be left in the eld for many months. Embedded software would then increase sampling frequency during periods of rainfall. We hypothesized that this contrast between no-volume ow in karst passageways dur- ing dry periods and partial- or saturated-volume ow during a rain event is detectable by these Wireless Sensor Network (WSN) geophysical nodes, we call this a Wireless Resistivity Network (WRN). The development of new methodologies to characterize semi-arid karst hydrology is intended to augment Sandia National Laboratorys mission to lead e orts in energy technologies, waste disposal and climate security by helping to identify safe and secure regions and those that are at risk. Development and initial eld testing identi ed technological barriers to using WRNs for identifying semi-arid karst, exposing R&D which can be targeted in the future. Gravity, seismic, and resis- tivity surveys elucidated how each technique might e ectively be used to characterize semi-arid karst. This research brings to light the importance and challenges with char- acterizing semi-arid karst through a multi-method geophysical study. As there have been very few studies with this emphasis, this study has expanded the body of practical experience needed to protect the nations water and energy security interests.

Barnhart, Kevin Scott↗

Delayed Gamma-ray Spectroscopy for Non-Destructive Assay of Nuclear Materials

This project addresses the need for improved non-destructive assay techniques for quantifying the actinide composition of spent nuclear fuel and for the independent verification of declared quantities of special nuclear materials at key stages of the fuel cycle. High-energy delayed gamma-ray spectroscopy following neutron irradiation is a potential technique for directly assaying spent fuel assemblies and achieving the safeguards goal of quantifying nuclear material inventories for spent fuel handling, interim storage, reprocessing facilities, repository sites, and final disposal. Other potential applications include determination of MOX fuel composition, characterization of nuclear waste packages, and challenges in homeland security and arms control verification.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗

Chrysler Upset Protrusion Joining Techniques for Joining Dissimilar Metals

The project goal was to develop and demonstrate a robust, cost effective, and versatile joining technique, known as Upset Protrusion Joining (UPJ), for joining challenging dissimilar metal com-binations, especially those where one of the metals is a die cast magnesium (Mg) component. Since two of the key obstacles preventing more widespread use of light metals (especially in high volume automotive applications) are 1) a lack of robust joining techniques and 2) susceptibility to galvanic corrosion, and since the majority of the joint combinations evaluated in this project include die cast Mg (the lightest structural metal) as one of the two materials being joined, and since die casting is the most common and cost effective process for producing Mg components, then successful project completion provides a key enabler to high volume application of lightweight materials, thus potentially leading to reduced costs, and encouraging implementation of lightweight multi-material vehicles for significant reductions in energy consumption and reduced greenhouse gas emissions. Eco-nomic benefits to end-use consumers are achieved primarily via the reduction in fuel consumption. Unlike currently available commercial processes, the UPJ process relies on a very robust mechanical joint rather than intermetallic bonding, so the more cathodic material can be coated prior to joining, thus creating a robust isolation against galvanic attack on the more anodic material. Additionally, since the UPJ protrusion is going through a hole that can be pre-drilled or pre-punched prior to coating, the UPJ process is less likely to damage the coating when the joint is being made. Further-more, since there is no additional cathodic material (such as a steel fastener) used to create the joint, there is no joining induced galvanic activity beyond that of the two parent materials. In accordance with its originally proposed plan, this project has successfully developed process variants of UPJ to enable joining of Mg die castings to aluminum (Al) and steel sheet components of various thicknesses, strengths and coating configurations. While most development focused on the simpler round boss version of the process, an additional phase of the work focused on devel-opment of an oval boss version to support applications with narrow flanges, while yet another vari-ant of the process, known as Upset Cast Riveting (UCR), was developed and evaluated for joining mixed metals that may not necessarily include Mg or Al die cast components. Although each varia-tion posed unique challenges described later in the report, all variations were successfully produced and evaluated, and each could be further developed for specific types of commercial applications. In this project, UPJ performed favorably against the benchmark self-pierce riveting (SPR) process in Mg AM60B to Al 6013 combinations although significant corrosion challenges were observed in both processes, especially for the bare Mg to bare Al configurations. Additional challenges were observed in joining Mg to steel with the UPJ process (SPR was not evaluated for this combination as it was not considered viable). To pass FCA’s specified corrosion tests with Mg/steel combina-tions, new steel treatments were evaluated, as well as adhesives and sealed edges. These showed significant improvement. In general, UPJ performed very well in Mg to Al 6016 combinations, even in corrosion evaluation of the bare Mg to bare Al configuration (again, SPR was not evaluated for this material combination as the 1.1 mm thick Al6016 sheet thickness was considered too thin for the SPR process). The improvement in corrosion performance of the Mg to Al 6016 combina-tion over the Mg to Al 6013 combination was thought to be a result of the lower copper content in the Al 6016 alloy. Oval boss joints showed substantial improvement in all joint strength criteria compared to 8.0-mm diameter round boss joints but were not evaluated for corrosion performance. The improved joint strength is likely a result of larger shear area. Cosmetic corrosion performance of all test assemblies (UPJ, UCR and SPR) was a challenge due to exposed edges and crevices al-lowing undercutting of the coatings. In real world component applications, the exposed edges, so prevalent on the joining test coupons, would be less prevalent and easier to protect.

36 MATERIALS SCIENCE↗

Material Control and Accounting for Liquid-Fueled Molten Salt Reactors: Material Control and Holdup Considerations

The US Nuclear Regulatory Commission (NRC) will likely require license applicants for liquid-fueled molten salt reactors (MSRs) with circulating fuel to submit a nuclear material control and accounting (MC&A) plan or detailed MC&A program description for the facility. In liquid-fueled MSRs with special nuclear material (SNM) in bulk (i.e., not in discrete items) form and rapidly changing quantities due to fuel transmutation and depletion, using traditional nuclear material accounting methods with material balance evaluations is challenging. In reactors with changing inventories, these expected quantities of SNM must be calculated based on operational parameters. Reducing uncertainties on these expected quantities is challenging in the case of MSRs without decades of operational experience to verify and validate predictive computational codes. Moreover, many areas in MSRs are inaccessible because of high-temperature and high-radiation environments, making measurements challenging.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗