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NASA NTRS1990

Process development and fabrication of space station type aluminum-clad graphite epoxy struts

The manufacture of aluminum-clad graphite epoxy struts, designed for application to the Space Station truss structure, is described. The strut requirements are identified, and the strut material selection rationale is discussed. The manufacturing procedure is described, and shop documents describing the details are included. Dry graphite fiber, Pitch-75, is pulled between two concentric aluminum tubes. Epoxy resin is then injected and cured. After reduction of the aluminum wall thickness by chemical milling the end fittings are bonded on the tubes. A discussion of the characteristics of the manufactured struts, i.e., geometry, weight, and any anomalies of the individual struts is included.

Ring, L. R.↗

DOE OSTI2018

N-BAR PROBLEMS AS APPROXIMATIONS TO THE BREE PROBLEM

The Bree solution to the problem of a ratcheting cylinder under constant pressure and cyclic thermal load is a fundamental result in nuclear engineering and widely used as the technical basis for the ASME Boiler and Pressure Vessel Code and other design methods. However, because the loading conditions in the Bree problem are difficult to achieve experimentally there have been relatively few works experimentally examining the problem and extending it to other relevant design situations, for example cladded components. In contrast, 2-bar problems are widely studied experimentally and are relatively easy to setup. These 2-bar problems are thought to be representative of Bree-type geometries, but a formal connection has not been demonstrated. This work formally establishes the connection between the Bree cylinder and an n-bar problem – a coupled bar experiment with, in general, more than two bars linked in parallel. The connection suggests that n-bar experiments using a fairly limited number of bars might be an experimentally-accessible setup that better represents ratcheting phenomenon in actual nuclear pressurized components. Such experiments could test surrogate cladded or multi-material components by using bars of different materials. Finally, this work suggests control schemes that yield optimally efficient n-bar experiments – experiments that best replicates a Bree cylinder with a limited number of bars.

Messner, Mark C.↗

DOE OSTI2019

{ital n}-bar problems as approximations to the Bree problem

The Bree solution to the problem of a ratcheting cylinder under constant pressure and cyclic thermal load is a fundamental result in nuclear engineering and widely used as the technical basis for the ASME Boiler and Pressure Vessel Code and other design methods. However, because the loading conditions in the Bree problem are difficult to achieve experimentally there have been relatively few works experimentally examining the problem and extending it to other relevant design situations, for example cladded components. In contrast, 2-bar problems are widely studied experimentally and are relatively easy to setup. These 2-bar problems are thought to be representative of Bree-type geometries, but a formal connection has not been demonstrated. This work formally establishes the connection between the Bree cylinder and an n-bar problem a coupled bar experiment with, in general, more than two bars linked in parallel. The connection suggests that n-bar experiments using a fairly limited number of bars might be an experimentally-accessible setup that better represents ratcheting phenomenon in actual nuclear pressurized components. Such experiments could test surrogate cladded or multi-material components by using bars of different materials. Finally, this work suggests control schemes that yield optimally efficient n-bar experiments experiments that best replicates a Bree cylinder with a limited number of bars.

Messner, M. C.↗

DOE OSTI2025

A review of prefabricated overclad panels for building energy retrofits

Over 50% of existing buildings in the US were constructed before 1980 and lack modern, energy-efficient technologies. Despite the large number of old buildings, less than 2% of buildings are retrofitted each year with many of those retrofits consisting of small incremental improvements that often ignore the building envelope. The opaque building envelope affects 25% of building energy use equating to 10% of total US primary energy use. To meet decarbonization goals by 2035, the US must increase the rate of retrofit of existing building envelopes. The process of prefabrication is one potential opportunity to increase the rate of envelope retrofits and reduce costs. Prefabrication is the process of manufacturing construction components in a controlled environment before shipping them on-site to perform assembly of the construction. The process of prefabrication has been demonstrated to increase production efficiency, quality control, and rate of production. Prefabricated overclad panels, consisting of a cladding material and insulation designed to be installed over the existing envelope, have been successfully used to retrofit existing buildings in countries other than the US; however, the technology still has some barriers to major market adoption within the US. This paper will present a review of current implementations and demonstrations of prefabricated overclad panels. The review will include identification of panel types, materials, and applications and aims to identify barriers to retrofit market entry and widespread adoption.

Hayes, Nolan [ORNL] (ORCID:0000000332245718)↗

DOE OSTI2019

Fuel performance evaluation of annular metallic fuels for an advanced fast reactor concept

The annular metallic fuel design has been considered as a promising solution to relieve the fuel-cladding mechanical interaction (FCMI) caused by significant swelling of metallic fuels at high burnup. In this paper, the fuel performance of annular metallic fuel with HT9 cladding in a novel advanced fast reactor design optimized for high-burnup performance was evaluated. The BISON code was utilized to perform fuel performance simulations using a modified fuel swelling correlation for conservative evaluation. Three representative fuel pins, which feature highest peak burnup (26.8 % FIMA, fission per initial metal atom), highest fuel temperature (645 degrees C), and highest fast neutron flux (2.68 x 10 15 n/cm 2 ∙s), were investigated. A series of key fuel performance parameters, such as peak fuel temperature, reduction of fuel central void, and cumulative damage fraction (CDF) of the cladding, were predicted for both upper and lower plenum configurations. It is evident that annular fuel design with a lower smeared density is capable of accommodating the significant swelling of metallic fuel at high burnup. Finally, the adoption of a lower plenum configuration effectively lowers the gas pressure inside the rod and reduces the creep damage accumulation in the HT9 cladding.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

Development and Evaluation of Embrittlement Resistant Alloys for Advanced LWR Cladding

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2025

Development and Evaluation of Embrittlement Resistant Alloys for Advanced LWR Cladding

The primary challenge preventing the deployment of reactor designs that leverage wrought FeCrAl as an advanced light-water reactor (LWR) cladding is irradiation hardening and embrittlement. Wrought FeCrAl alloys experience a loss of post-irradiation ductility and fracture toughness under low-temperature neutron irradiation (< 350°C) resulting from the combined effects of dislocation loop formation and the irradiation-enhanced precipitation of Cr-rich alpha-prime precipitates throughout the microstructure. Over the past decade, significant improvements in our understanding regarding the effect of Cr and Al content have been enabled through neutron irradiations, but even optimized wrought alloys such as C26M remain vulnerable to brittle failure during storage, transportation, and handling following irradiation. This report summarizes multiple irradiation campaigns initiated over the past 7 years and provides recent insights into the effect of (1) minor alloying composition, (2) alloy processing strategy, (3) crystal structure, and (4) the use of advanced oxide dispersion strengthened (ODS) alloys. The first irradiation campaign irradiated body-centered cubic FeCrAl alloy C26M with and without elements such as Mo and Y. It also included face-centered cubic alumina-forming austenitic (AFA) alloys that have a steeper neutronic penalty but that have far superior high-temperature strength than wrought FeCrAl. Finally, this campaign compared conventional wrought C26M with the same material produced using modern powder metallurgical hotisostatic pressing (PM-HIP). The second irradiation campaign, initiated in 2018, irradiated wrought and FeCrAl-ODS materials up to 50 dpa to assess their irradiation resistance to high-doses. The results of this work indicate that wrought C26M suffers significant degradation following neutron irradiation at LWR-relevant temperatures regardless of Mo and Y content. The PM-HIP variant appears to be more resilient to irradiation-induced ductility loss in comparison with wrought variants but still loses almost all ductility by 8 dpa. Notably, the AFAs retained at least 5% total elongation after irradiation at the same dose and temperature condition, although additional deterioration may be expected at higher fluence levels. FeCrAl-ODS materials show the greatest promise as a transformative longer-term accident-tolerant fuel (LT-ATF) cladding material. The extruded 106ZY10C alloy retained 10% total elongation after 16 dpa irradiation and retained 8% total elongation after 50 dpa. Additional challenges with FeCrAl-ODS alloys also remain, including optimization of end cap joining methodologies, enhancement of fracture toughness, scaling production to prove economic viability, and optimizing post-pilger heat treatments to maximize ductility and irradiation resistance. This series of irradiations demonstrates the efficacy of the LWR testbed available within the United States for the rapid irradiation and down-selection of LT-ATF candidates. As the country rapidly accelerates its timeline for the deployment of advanced reactor concepts, the effective utilization of the High Flux Isotope Reactor for separate-effects style irradiations should continue to be prioritized to answer the final questions pertaining to LT-ATF candidates necessary for the deployment of advanced boiling water reactors and small modular reactors.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2021

Fuel-cladding chemical interaction of a prototype annular U-10Zr fuel with Fe-12Cr ferritic/martensitic HT-9 cladding

As an alternative fuel form, the annular metallic fuel design eliminates the liquid sodium bond between the fuel and the cladding, providing back-end fuel cycle and other benefits. The fuel-cladding chemical interaction (FCCI) of annular fuel also presents new features. In this work, state-of-the-art electron microscopy and spectroscopy techniques were used to study the FCCI of a prototype annular U-10wt%Zr (U-10Zr) fuel with ferritic/martensitic HT-9 cladding irradiated to 3.3% fission per initial heavy atom. Compared with sodium-bonded solid fuels, negligible amounts of lanthanides were found in the FCCI layer in the investigated helium-bonded annular fuel. Instead, most lanthanides were retained in the newly formed UZr 2 phase in the fuel center region. The interdiffusion of iron and uranium resulted in tetragonal ( U ,Zr) 6 Fe phase (space group I4/mcm) and cubic ( U ,Zr)( Fe ,Cr) 2 phase (space group Fd 3 ¯ m). The ( U ,Zr)( Fe ,Cr) 2 phase contains a high density of voids and intergranular uranium monocarbides of NaCl-type crystal structure (space group Fm 3 ¯ m). At the interdiffusion zone and inner cladding interface, a porous lamellar structure composed of alternating Cr-rich layers and U-rich layers was observed. Next to the lamellar region, the unexpected phase transformation from body-centered cubic ferrite (α-Fe) to tetragonal binary Fe-Cr σ phase (space group P4 2 /mnm) occurred, and tetragonal Fe-Cr-U-Si phase (space group I4/mmm) was identified. Due to the diffusion of carbon into the interdiffusion zone, carbon depletion inside the HT-9 led to the disappearance of the martensite lath structure, and intergranular U-rich carbides formed as a result of the diffusion of uranium into the cladding. These detailed new findings reveal the unique features of the FCCI behavior of annular U-Zr fuels, which could be a promising alternative fuel form for high burnup fast reactor applications.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

(Oxy)hydroxides Formed on Aluminum Fuel Materials After Irradiation and Long-Term Wet Storage - 20354

The aluminum cladding of research-reactor fuel experiences general corrosion when in contact with water during in-reactor service and post-discharge wet storage, resulting in the formation of adherent aluminum (oxy)hydroxide films. These (oxy)hydroxides contain chemically-bound water that poses challenges for extended dry storage due to the risk of thermal or radiolytic decomposition releasing free water and/or hydrogen and oxygen gases. This study describes characterization of the (oxy)hydroxides present on several aluminum materials used in reactor operation and subsequently stored wet in the L-Basin storage facility at the Savannah River Site (SRS) for an extended period. Characterization data providing insight into the loading, composition, and morphology of (oxy)hydroxides to be expected on service-exposed aluminum cladding provides valuable benchmarks for designing adequate drying and dry-storage approaches. This work is part of a broader investigation to address knowledge gaps and technical data needs for dry storage of aluminum-clad spent nuclear fuel (ASNF), which included in-lab growth of (oxy)hydroxide films on aluminum alloy substrates to investigate formation behavior, investigation of drying methods to remove existing (oxy)hydroxides from ASNF cladding, and measurement of radiolytic yield of hydrogen from (oxy)hydroxide powders and films. In this study, (oxy)hydroxide films were characterized for three aluminum-alloy materials used in reactors and subsequently stored wet for up to approximately 40 years in L-Basin at SRS: one cropping from a Missouri University Research Reactor (MURR) fuel element (Al-6061 alloy), one cropping from a Universal Sleeve Housing (USH) (Al-6063 alloy), and one Mark-16B fuel assembly (either Al-6061 or Al- 6063). The USH and Mark-16B were used in SRS production reactors. Characterization of the as-received (oxy)hydroxides included scanning electron microscopy (SEM) in both plan-view and cross-section to characterize the (oxy)hydroxide layer's morphology, thickness, and structure. X-ray diffraction (XRD) was used to identify the chemical composition and distinguish between the various aluminum (oxy)hydroxides known to form under reactor and storage conditions. XRD analysis revealed both bayerite (Al(OH)3) and boehmite (AlOOH) on the surface of the MURR and USH samples, as well as bayerite, boehmite, and gibbsite (another Al(OH)3 polymorph) on the surface of the Mark-16B sample. The aluminum trihydroxides, bayerite and gibbsite, are typically associated with corrosion in low-temperature (<80 deg. C) water, while boehmite is expected to form at higher water temperature (>80 deg. C). The presence of bayerite on the USH, which is believed to have operated close to 90 deg. C, suggests that boehmite formed during in-reactor exposure was not protective against further hydroxide growth in low-temperature wet storage. Cross-section scanning electron microscopy (SEM) showed total (oxy)hydroxide layer thicknesses of ∼5- 10 μm for the MURR and ∼5-15 μm for the Mark-16B. The thickness of the USH's (oxy)hydroxide layer was indiscernible by the current mounting and imaging method, despite plan-view SEM and XRD confirming the presence of an (oxy)hydroxide layer. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2025

Damage progression and failure of SiC/SiC composite tubes under hard-contact radial expansion

The response of silicon carbide (SiC) fiber-reinforced SiC matrix (SiC/SiC) composite cladding to mechanical interaction with fissile fuel is a knowledge gap that must be overcome to design and assess SiC-based cladding systems for advanced nuclear applications. This study developed the relevant mechanical testing capability and identified the failure behavior and the critical microstructural features and processing defects. Sections of SiC composite tube were subjected to a modified expansion-due-to-compression (EDC) test in an X-ray computed tomography microscope: a polyurethane plug pressed surrogate Al 2 O 3 into the inner walls of the SiC/SiC composite tubes to achieve hard contact. A pure EDC test with just a polyurethane plug was also performed as a reference. Through the use of displacement fields, digital volume correlation revealed inhomogeneous deformation fields in the tubes, even for pure EDC, which was related to the inherent defects in the structure. Deep learning–aided segmentation and systematic data analysis revealed that the presence of inhomogeneous deformation applied by the hard contact was exaggerated by the presence of inner surface imperfections left behind from the matrix densification process. In conclusion, the findings provide insights into the applications, highlighting the necessity for improvements in inner surface roughness and the incorporation of localized contacts in pellet–cladding mechanical interaction computational models.

Composites↗

NASA NTRS1968

Design, fabrication, and fabrication evaluation of high temperature fuel pins Final report

Design, fabrication, and evaluation of molybdenum clad, high temperature fuel pins used to provide irradiation data for full length reactor fuel pin design

Gulley, R. L.↗

NASA NTRS1984

Fabrication development for ODS-superalloy, air-cooled turbine blades

MA-600 is a gamma prime and oxide dispersion strengthened superalloy made by mechanical alloying. At the initiation of this program, MA-6000 was available as an experimental alloy only and did not go into production until late in the program. The objective of this program was to develop a thermal-mechanical-processing approach which would yield the necessary elongated grain structure and desirable mechanical properties after conventional press forging. Forging evaluations were performed to select optimum thermal-mechanical-processing conditions. These forging evaluations indicated that MA-6000 was extremely sensitive to die chilling. In order to conventionally hot forge the alloy, an adherent cladding, either the original extrusion can or a thick plating, was required to prevent cracking of the workpiece. Die design must reflect the requirement of cladding. MA-6000 was found to be sensitive to the forging temperature. The correct temperature required to obtain the proper grain structure after recrystallization was found to be between 1010-1065 C (1850-1950 F). The deformation level did not affect subsequent crystallization; however, sharp transition areas in tooling designs should be avoided in forming a blade shape because of the potential for grain structure discontinuities. Starting material to be used for forging should be processed so that it is capable of being zone annealed to a coarse elongated grain structure as bar stock. This conclusion means that standard processed bar materials can be used.

Moracz, D. J.↗

DOE OSTI2022

Design of separate-effects In-Pile transient boiling experiments at the $\mathrm{TREAT}$ Facility

The cladding-to-coolant heat transfer during rapid power excursions, such as reactivity-initiated accidents, remains a crucial area of uncertainty in nuclear reactor safety. This uncertainty impacts the ability to accurately predict fuel performance behavior for these conditions. Improving our understanding of transient cladding-to-coolant heat transfer will enhance our ability to model design basis accidents and could increase design and improve safety margins for the current commercial fleet and advanced reactors. To address these issues, the Critical Heat Flux Static Environment Rodlet Transient Test Apparatus (CHF-SERTTA) experiment uses a novel approach with a borated stainless-steel heater rodlet that can replicate cladding temperatures experienced during a design-basis reactivity-initiated accident. The rodlet and experiment is instrumented to provide temperature and thermal-hydraulic conditions throughout the transient. The design of the rodlet allows for separate effect testing that eliminates complexities with a fuel and cladding specimen. We report the experiments have provided data that will be used to improve our understanding of boiling behavior, specifically critical heat flux, under transient conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2019

Advanced Manufacturing for Nuclear Core Design

Advanced manufacturing technologies are fundamentally altering the way components are designed and manufactured. The potential application space for these technologies within the nuclear industry is broad, with the most benefit realized from the ability to approach the nuclear core design problem from an advanced manufacturing perspective. One of the objectives of the US Department of Energy Office of Nuclear Energy (DOE-NE) Transformational Challenge Reactor (TCR) design and analysis activities is to demonstrate this approach. A mitigation of conventional constraints realized through advanced manufacturing methods opens up the design space, enabling a departure from typical axially uniform pin- and plate-type geometries common to operating nuclear reactors worldwide. The ability to use more heterogenous spatial and material flexibilities increases the demand on the predictability of higher fidelity modeling and simulation tools to be used in the design, analysis, and optimization processes. Leading candidate advanced manufacturing technologies for nuclear applications are additive methods that provide for high dimensional control and incorporation of small design features such as powder bed fusion, laser directed energy deposition, and binder jetting. Potential applications to core design identified during DOE-NE TCR design and analysis activities fall into two categories: those that are uniquely enabled by additive manufacturing technologies and those that are enabled by the characteristics (e.g., repeatability and instrumentation) inherent to these technologies. Identified applications include reducing in-core parasitic materials, improving heat conduction across fuel-cladding gaps, optimizing thermal hydraulic design with flexible coolant flow geometries, reducing uncertainties in manufactured components, optimizing core conditions with design heterogeneities, designing feedback mechanisms using material characteristics, and embedding burnable poisons. These potential applications join a host of nuclear industry advanced manufacturing activities to recapture manufacturing capabilities, reduce manufacturing costs and complexities, reduce deployment time, and enable performance and operational enhancements.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2020

Initial development for the introduction of provisions for irradiation effects in Division 5 rules

Heat pipe microreactor concept involves the use of a solid core block to encapsulate the nuclear fuel and the passive extraction of heat from the core block to the heat exchanger block by advanced heat pipes. The core block is a key component of the heat pipe microreactor design. It integrates the functions of reactor vessel, core structural components and fuel cladding of traditional Gen IV advanced reactor designs into one single structure. The effects of irradiation on the structural failure modes are outside the scope of the rules of the ASME Boiler and Pressure Vessel Code, Section III, Division 5, Subsection HB, Subpart B for metallic Class A components. Methods to account for the irradiation effect were investigated in order to develop design methods that can be applied to the core block design in the heat pipe microreactor. This report presents an initial development of an approach to account for the irradiation effects in the design rules.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2018

Purification of Zirconium Tetrachloride from UNF Cladding

The overarching objective of this project was to design and demonstrate (laboratory scale) a sublimation-based purification protocol to obtain pure zirconium chloride (ZrCl 4 ) from impure feeds derived from the direct chlorination of Zircaloy cladding materials. To achieve this objective the following goals were identified. (1) Design, construct and test an apparatus capable of subliming and purifying ZrCl 4 at room pressure and temperatures between 600 and 700K (sublimation point of ZrCl 4 606 K). (2) Investigate reactions that may be used to chemically alter the makeup of the impure ZrCl 4 feed materials to enhance their removal from ZrCl 4 in the sublimation-purification process. (3) Computationally model the reactions of metal chlorides in the gas phase under sublimation conditions (1 atm, 600 K) to determine if oligomers are formed and, if formed, how they would influence sublimation-based purification strategies. All experimental investigations performed as part of this project used stable (non-radioactive) compounds and materials as surrogates for the impure ZrCl 4 feedstocks that would be derived from the direct chlorination of Zircaloy claddings from spent nuclear fuel rods. No radiologically active materials were used at any time in this project. The impure ZrCl 4 feedstocks used in this study were derived from the direct chlorination of virgin Zircaloy claddings (Zirc-2, -4 and high Nb). The main goals of the project have been achieved during the three year period of this NEUP sponsored project. The impurities of greatest concern in the feedstocks were identified as the chloride salts of the alloying metals originally found in Zircaloy (Fe, Nb) and radiological contaminants (Cs, Sr, and Sb). A series of prototype sublimation apparatuses and heating arrangements were designed, constructed and tested. A general batch-sublimator design was identified and a 1.0 kg batch sublimator/oven system was fabricated from glass. Purification protocols developed on small scale runs were found to scale to 1.0 kg batches without problem. Purification factor ranges (PFs: wt% final / wt% initial ) for the impurity chlorides identified above have been measured: Fe (40 – 500); Nb (10 – 70); Cs (520 – 5000); Sr (250 – 1900); Sb (220 – 2300). Removal of iron is sensitive to the partial pressure of hydrogen (in nitrogen) flowing through the system. Lower PFs are obtained when the hydrogen partial pressure is lower. Lower flows are required to raise the iron PF. Removal of niobium is very sensitive to the endpoint determination of the forerun. When judged properly (color change yellow to white in sublimate) reasonable PFs (500) can be achieved. However, as discussed in the conclusions section, endpoint determination can be difficult and significantly lower PFs may be obtained when the endpoint is not correctly identified. Several methods of obtaining quantitative cesium, strontium and antimony PFs were investigated, the best being ICP-MS. The results reported here reflect very limited access to and ICP-MS instrument during the course of this project. It should be noted that, while significantly reduced amounts of antimony are observed in the product, antimony is also found in the exhaust trap after the condenser. Quantum chemical computational modeling studies of the reactions of metal chlorides in the gas phase provided important information about a chemical transport mechanism that could defeat sublimation-based purification strategies. Many metal chlorides are known to oligomerize in the gas phase to reduce their coordinative unsaturation. Dimerization of ZrCl 4 with iron or niobium chlorides would provide a mechanism by which both metal impurities could be carried over with the product during sublimation of the bulk material. Computational modeling conclusively showed that under sublimation conditions in the gas phase, dimerization reactions are not favored thermodynamically due to a significant entropic penalty for the reaction. Therefore, consistent with experimental observations, sublimation-based purification strategies remain viable. The experimental part of the report concludes with a short discussion of the issues that still remain in developing a technologically relevant Zircaloy recycling strategy. The final section of the report is a summary of the quantum chemical investigation of the reactions of metal chlorides in the gas phase under sublimation conditions.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2018

Thermomechanical Design Optimization of Roll-to-Roll Manufactured Hybrid Metal Polymer Heat Exchanger

Metal based heat exchangers for ultra-low temperature (<100C) waste heat recovery (WHR) face high maintenance and material costs that surpass profits from recovery[1]. Low-cost polymer pipes suffer from poor thermal conductivity (~0.2 W/mK), and thus a low overall heat transfer coefficient. Thermal conductivity of polymers can be enhanced through systematic molecular alignment by stretching[2], chemical vapor deposition of extended polymers [3], surface grating [4], etc. However, these approaches suffer from scalability issues for WHR heat exchangers in terms of cost [3], and thermomechanical considerations [5]. In this work, we propose novel hybrid metal-polymer strips that are strategically designed to enhance thermal conductivity by cladding copper around polymer strips. The polymer strips with copper-clad edges are wound helically to form the pipes for a heat exchanger. Through finite element simulations, we optimize the design for enhanced thermal performance of the polymer-metal strips. We predict up to 20 % enhancement in overall heat transfer from a base polymer with a thermal conductivity of 0.2 W/mK. We show that the thermomechanical strains induced in the pipes can be reduced to < 0.2 % by strategically stacking and welding multiple polymer-metal strips. We verified the improved thermal performance of the roll-to-roll manufactured strips through a thermal conductivity measurement setup. The optimized design of hybrid metal-polymer pipes provides a scalable and cost-effective means for harvesting waste heat from ultra-low temperature sources. Thekdi, Arvind, and Sachin U. Nimbalkar. Oak Ridge National Lab, 2015 Shen, Sheng, et al. Nature nanotechnology 5.4 (2010): 251. Xu, Yanfei, et al. Science advances 4.3 (2018): 3031. Smith, M. K., Singh, V., Kalaitzidou, K., & Cola, B. A. (2015). ACS nano, 9(2), 1080-1088. Aymonier, C., Bortzmeyer, D., Thomann, R., & Mülhaupt, R. (2003). Chemistry of Materials, 15(25), 4874-4878.

C Rajagopal, Manjunath↗

DOE OSTI2023

Editorial: Benchmark experiments, development and needs in support of advanced reactor design

Advanced nuclear reactor designs will for the most part be a departure from low enrichment light water reactor (LWR) designs currently operated around the world. Such advanced designs include but are not limited to new TRISO-fueled high temperature gas reactors, heat-pipe cooled micro-reactors, fluoride salt cooled high-temperature reactors, molten salt reactors, lead cooled fast reactors, nuclear thermal propulsion concepts, and include LWR designs with advanced fuel and clad types. Modeling and simulation methods for advanced reactors is necessary for regulators to approve license requests. However, regulators also require that modeling approaches be validated against experimental measurements. Hence, there is a crucial need for data for advanced reactor systems that will support validation of analysis methods. To this end, this Research Topic includes eleven papers organized into topical seven categories relevant for advanced reactor design.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗