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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.

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All sourcesNASA NTRSDOE OSTI
DOE OSTI2026

Hydrogen Recombiner Catalyst Evaluations for Waste Storage

Radiolysis of water in nuclear waste storage generates hydrogen gas that can accumulate within sludge style waste and be rapidly released during agitation events, creating a significant flammability hazard. Engineering controls are therefore required to limit hydrogen concentrations during both quiescent storage and transient disturbances. Catalytic recombination of hydrogen in waste storage offgas is a proven mitigation strategy, maintaining hydrogen levels below flammability limits and managing sudden concentration spikes. Conventional recombiners rely on platinum and/or palladium catalysts, with development efforts focused on extending service life, increasing active surface area, and ensuring safe deployment in radioactive environments. Savannah River National Laboratory (SRNL) is evaluating a newly developed hydrogen recombiner catalyst from Canadian Nuclear Laboratories as a cost-effective and durable alternative for nuclear waste applications. Testing was conducted in SRNL’s Shielded Cells facility, which enables reduced-scale experimental modeling under radiation fields and near-use-case conditions relevant to radioactive waste storage. Catalyst performance was evaluated using a custom offgas characterization system designed for near-zero flow conditions. The experimental apparatus consisted of a gas-tight 2.7 L PTFE vessel equipped with temperature monitoring, gas flow controls, and a variable-speed mixer to simulate sludge agitation. Offgas composition was monitored using a dedicated gas chromatograph with argon carrier gas and a krypton internal standard. Measurements were obtained for an empty vessel, the vessel containing a well characterized radioactive tank waste sample, and the same configuration with the candidate catalyst installed. Results demonstrate that the new catalyst effectively reduced hydrogen concentrations in the offgas within the constraints of the experimental design. In addition to confirming catalytic activity, the testing provided valuable insights into experimental optimization and considerations for future performance evaluations. These findings support the potential scalability of the technology and highlight its applicability to broader nuclear waste management operations, offering improved safety and reduced operational costs through enhanced catalyst durability and lower replacement frequency.

Tener, Zachary P. [Savannah River National Laborat↗
DOE OSTI2020

Hydrogen Transport in a Model 9979 Shipping Package with Inner Convenience Cans

Radiolytic hydrogen production and accumulation inside containment packages is a concern at any facility responsible for their packaging, storage, transportation, and/or disposal. When hydrogen gas accumulates to concentrations above the Lower Flammability Limit (LFL) which is 4% or 40,000 ppm in air, the possibility of a deflagration or explosion increases. This concern persists over the course of the package lifetime which is unlimited when disposed of by burial or in permanent repositories. Here, we report on a numerical model used to predict the concentration of hydrogen within each layer of a Model 9979 package containing a convenience can assembly. Simulations show the hydrogen concentration to always be highest in the inner convenience can containing the radioactive source. When the radioactive source is within the Los Alamos National Laboratory (LANL) Packaging Limits, the hydrogen concentration is shown to remain well below the LFL at all times including packaging, storage, transportation, and disposal. A hydrogen transport model is presented for a Model 9979 package system containing a nested arrangement of convenience cans, which are tin oxide coated steel cans of various sizes with a slip-lid assembly. The inner convenience can contains the radioactive source material along with an unknown quantity of incidental water acquired from humid air or processing. While visible organic materials such as paper and plastics were purposely excluded from the inner can, it is not possible to claim the wastes are entirely organic free. The inner convenience can is tape sealed and placed into a plastic bag which is horsetail closed (i.e., twisted and taped). The bagged can is placed into an outer convenience can that is also tape sealed. The can assembly is then placed into the 30 gallon drum and subsequently placed inside the 55 gallon drum in the 9979 package. Here we assume hydrogen gas is produced in the inner convenience can from alpha radiolysis of water at a rate dependent on the quantity of uranium isotopes and water present. The hydrogen transport model was used to calculate hydrogen accumulations within the package’s five layers at different times and conditions. These simulations serve two purposes; (i) to build confidence in the model by comparing predicted values to measured values, and (ii) to check the steady state hydrogen concentrations that are approached at long times in the package’s lifetime. Model simulations were compared to gas samples taken from the 30 gallon drum after storage at LANL’s Chemistry and Metallurgy Research (CMR) building for around 500 days. Hydrogen concentration calculations over much longer periods (i.e., more than 270 years) included extreme storage durations, transportation at extreme cold temperatures, and disposal of packages assuming different average temperatures.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

In-Situ Spatial Mapping of Hydrogen in Yttrium Hydrides at LANSCE (FY23 Version, Rev. 1)

This report summarizes the development of neutron imaging capabilities and experimental activities performed at the Los Alamos Neutron Science Center (LANSCE) with the main goal of measuring temperature-driven hydrogen diffusion within bulk-yttrium hydride (YH x ) materials. Yttrium hydride is the leading candidate to serve as a solid neutron moderator in microreactor cores, owing to its high density of hydrogen atoms as well as its superior thermal stability compared to all other metal hydrides. The experimental results and technique developments reported herein support the U.S. Department of Energy Office of Nuclear Energy’s (DOE-NE) Microreactor Program under Technology Maturation. In particular, it addresses the critical need to experimentally validate and verify hydrogen-diffusion models of metal hydrides used in high-temperature microreactor designs by means of high-spatial-resolution neutron imaging. These capabilities were designed to apply large temperature gradients across centimeter-sized YH x pellets to simulate conditions faced in the microreactor environment. In principle, neutron imaging, combined with in-situ sample heating, enables near real-time tracking of hydrogen diffusion in YH x on the sub-millimeter scale. In this report, an overview of neutron imaging methodology and technologies are given in the context of recent spatial measures of hydrogen concentrations in similar metal hydrides. Additionally, the commissioning and operation of a custom-built compact dual-zone furnace is given along with details on three in-situ heating measurements of YH x performed over the 2020 to 2022 LANSCE operation cycles. The aims of these experiments ranged from furnace commissioning, determining sample quality, i.e., hydrogen uniformity via neutron computed tomography, and studying the effects of applied temperature-gradients on YH x pellets. Analyses and results from these neutron imaging measurements are given along with outlooks and guidelines for optimal future hydrogen diffusion measurements. Our conclusions are as follows. Image analyses indicate that centimeter-sized yttrium hydride cylindrical pellets exhibit uniform, whole-body hydrogen desorption and absorption without clear temperature dependence as reflected in the image attenuation at the opposing ends of each sample. This suggests that despite the large magnitude in temperature gradients applied by the furnace heating elements, the sample equilibrates to an unknown intermediate temperature. The origin of this result is likely the combination of short sample length (∼1cm) and use of a TZM can for containment where the latter created a thermal short across the sample. Nevertheless, the results from the most recent measurements indicate that neither significant concentration gradients of hydrogen were formed in centimeter-sized samples through the entire temperature range (25 °C to 950 °C) nor any formed due to temperature gradients on the order of 50 °C/cm up to 700 °C/cm. Furthermore, images from the FY2021 and FY2022 measurements indicate that samples of YH x , fabricated from either the direct hydride or powder metallurgy methods, are highly uniform in their hydrogen concentration to within the measurements’ spatial resolutions. The following questions arise from these latest results: 1) What is the intermediate temperature of the pellets in the TZM cans? 2) How quickly does the temperature equilibrate within the sample? and, 3) Do the observed changes in image attenuation follow known pressure-composition-temperature relations of yttrium hydride?

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Testing of In-Sodium Hydrogen Sensor

The goal of this project is to demonstrate diffusion-type in-sodium hydrogen meters (ISHMs) for real-time, in-situ monitoring of hydrogen concentration in molten sodium of a sodium-cooled fast reactor (SFR). The FY22 R&D efforts have been focused on (a) enhancement and fabrication of meshed ISHM prototype, (b) reconstruction of ISHM sodium test apparatus, (c) in-sodium test of meshed ISHM, and (d) ISHM performance and failure evaluations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Multiscale and Multimodal Characterization of 2D Titanium Carbonitride MXene

Here, a comprehensive study on the prototype solid solution phase carbonitride MXene Ti 3 CN is conducted using nuclear magnetic resonance, electron spin resonance, total and quasi-elastic neutron scattering, combined with density functional theory-based electronic structure and molecular dynamic calculations. The combination of experiment and theory lead toward rational atomic structural models of Ti 3 CN. The remnant Al ions from the etching process significantly tune the interlayer spacing, distinct from the more typical MXene, Ti 3 C 2 , prepared similarly. Neutron scattering indicates the surface terminations of Ti 3 CN display high oxygen and fluorine concentrations and rather low hydroxyl and hydrogen concentrations. Calculations show that the structure including both the residual Al ions and mixed surface terminations give the best agreement with the measurements. The water molecules in Ti 3 CN are highly immobile, in strong contrast to those in Ti 3 C 2 . The analysis of the electronic structure suggests that the nitride MXene displays higher conductivity than the carbides. Finally, the absence of hydroxyl groups in terminations, the solid-solution in the anion sites, the remnants within layers, and immobile water altogether make the carbonitrides a unique series in the MXene family, implying a further exploration of their exotic properties and applications in energy storage.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Fatigue Testing and Characterization of Pre-hydrided Zircaloy-4 Cladding Tubes

The solubility of hydrogen in Zircaloy-4 (Zry-4) at the cladding operating temperature is near 100 wt. ppm. Above this solubility limit, excess hydrogen precipitates as δ-hydride platelets in the cladding material. Because of the effect of the thermal gradient across the cladding thickness on the migration and precipitation of hydrogen, hydride rims are often observed at the cladding outer surface; under excessive corrosion conditions hydride blisters are possible. It is well known that the fatigue of metal alloys, especially high cycle fatigue, is sensitive to the status of surface including the microstructure, roughness, and residual stress. Thus, the question considered herein, is whether excessive hydrogen pickup modifies the microstructure such that it has a degradation effect on the fatigue performance of cladding during operation. This report describes the evaluation of fatigue performance of a pre-hydrided Zry-4 cladding. A commercial Zry-4 was polished and pre-hydrided to 800 ppm and 1300 ppm H contents. The fatigue testing was conducted under strain control at 5 Hz with fully-reversed bending by using a cyclic integrated reversible bending fatigue tester (CIRFT). Six specimens with 1300 ppm and one specimen with 800 ppm were tested. Significant variation in test results were observed. While two of the 1300 ppm specimens failed with less than 1 ×10 5 cycles (at 0.32% and 0.41%), the four other samples did not fail over the strain amplitude range of 0.25% to 0.38%. Interestingly, the fracture initiation site of the failed samples was on the outside diameter (OD) surface rather than on the inside diameter (ID) surface as is typical for the as-polished cladding. This suggested a degrading effect of the hydriding process. Subsequently, three of the unfailed specimens were then further tested at ~0.43% to observe where failure initiation occurred. Significant variation was also observed in these three specimens. One specimen failed at ~9000 cycles while the two others failed at 43000 and 1.06 ×10 5 cycles. The performance here correlated to the location of failure initiation; failure initiated on the OD in the sample that failed after 9000 cycles while it initiated on the ID in the sample that failed after 43000 and 1.06 ×10 5 cycles. Flat features at the OD initiation sites suggest a brittle hydride feature on the surface of those samples was the cause of the degradation in fatigue performance, though the overall hydrogen levels in all samples was similar. A summary of all the observations is provided below: • The un-failed specimens with 1300 ppm H were cycled to failure with a higher amplitude near 0.43%. In addition, the fatigue-treated specimen tended to have a longer fatigue at the same induced amplitude. • Fractography revealed a mixed failure mode for the pre-hydrided specimen. Particularly, the specimen with fracture initiation site (FIS) located on the outer diameter surface of tube tended to have a shorter fatigue life than that of FIS on the inner diameter surface. • Etched cross section was shown to have hydride platelets aligned with tube longitudinal axis as expected, and the density of hydrides is at the similar level as in literature data. Meanwhile, a LECO procedure was applied for hydrogen concentration measurement, which showed the measured hydrogen contents are close to the nominal value. • With the polished cladding tube as baseline, O’Donnell-Lager (O-L) analysis showed that a decrease of about 50% in reduction-of-area (RA) would be needed for the O-L fitting to the fatigue data of 1300 ppm cladding. The suggestion for the next steps is provided.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Investigation of desorption of hydrogen gas from polymer matrix using thermal desorption analysis and finite element modeling

Polymer materials are used widely as sealing materials and liners in hydrogen storage and transportation infrastructure. During operating conditions, polymers are exposed to high pressure hydrogen gas periodically. When in contact, the hydrogen gas diffuses through the polymer material with ease due to its smaller sized molecules compared to the polymer molecules. In the event of rapid decompression of outside hydrogen gas, the hydrogen molecules within the polymer diffuse out. This process is known as the desorption of polymer and can cause blisters or cracking causing permanent damage to the polymer. Here, we used a combined experimental and modeling approach to investigate the desorption of the hydrogen gas in the polymer. Thermal desorption analysis (TDA) experiments were performed on the polymer sample exposed to the high-pressure hydrogen gas to measure the diffusion coefficient and equilibrium hydrogen concentration. Experiments also provided the quantity of hydrogen within the polymer during the desorption process which followed Fick’s law. However, for higher pressurization values, some anomaly was observed in the quantity of hydrogen remaining in the polymer sample. A continuum mechanics based coupled diffusion–deformation-damage model was deployed to simulate the gas desorption process in the material using the finite element method to understand its effect on the material. It was found that the anomaly in TDA results were due to the blister formation on the polymer surface which resulted in longer hydrogen retention. Simulations also showed that blister formation can be affected by multiple parameters such as the addition of filler particles, and pre-existing cavity size.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Modeling of H2 Dispersion at ARIES

Hydrogen is a versatile and clean energy carrier that can be produced from various renewable sources such as wind, solar, and hydropower and help decarbonize electricity grids, industry, and transportation. Using the Hydrogen Research Facility under Advanced Research on Integrated Energy Systems (ARIES) at the National Renewable Energy Laboratory's (NREL) Flatirons campus as a test bench, the study examines the feasibility, useability, and value of using computational fluid dynamics (CFD) techniques to model hydrogen dispersion. The ARIES facility was chosen because controlled hydrogen releases can be performed at a rate of 27 kg-H2/hr. Site-specific atmospheric and weather condition data such as wind speed and temperature were used as inputs to the model. The results show statistical distributions and ranges of hydrogen concentrations at locations throughout the domain. Wind conditions are found to significantly impact the release behavior, including the hydrogen cloud's direction and concentrations. At low wind speeds (below 1 mph), hydrogen forms a cloud and at higher wind speeds (> 2-4 mph) hydrogen plume stretches in the direction of wind momentum. From >100 simulations for ARIES site-specific conditions, statistical quantities combined with a clustering algorithm were used to propose sensor location at various elevations from ground.

dispersion↗
DOE OSTI2019

Origins of hydrogen that passivates bulk defects in silicon heterojunction solar cells

Silicon heterojunction solar cell fabrication incorporates a significant amount of hydrogen into the silicon wafer bulk, and the amount of injected hydrogen is comparable to that introduced by silicon nitride films during a high-temperature firing step. In this work, the origins of the hydrogen injected during heterojunction cell processing have been identified. We demonstrate that the hydrogen plasma treatment that is routinely included to improve surface passivation considerably increases the hydrogen concentration in the wafers. We also show that the hydrogenated amorphous silicon i/p+ stack is more effective than the i/n+ stack for bulk hydrogen incorporation, and both are more effective than intrinsic films alone.

Sun, Chang (ORCID:0000000282140458)↗
DOE OSTI2020

Technology Enabling Zero-EPZ Micro Modular Reactors: Fabrication of Zirconium Hydride with Controlled Hydrogen Loading

This report completes Milestone M2.2.2 - ZrH pellet fabricated with controlled hydride loading: ZrH1.5±0.1 single phase pellets fabricated with controlled and thermally stable hydride loading. Hydrogen content to be quantified and single-phase confirmed by X-ray diffraction. In this milestone, we report the successful fabrication of delta-phase zirconium hydride pellets using the ORNL bulk metal hydriding system. The successful deployment of zirconium hydride moderator in advanced reactors requires development of a consistent and affordable production pathway along with implementation of a hydrogen retention solution throughout the reactor life. Fabrication of delta-phase zirconium hydride is challenging since the absorption of a large amount of hydrogen into alpha-zirconium induces significant volume expansion and the present hydrogen concentration gradient results in cracking. A fully programmable hydriding system with continuous hydrogen partial pressure and flow control to facilitate processing of massive metal hydride has been developed at ORNL. In this report, the working principle of the hydriding system will be introduced. Characterization of the produced zirconium hydride includes X-ray powder diffraction to identify the present phases and X-ray Computed Tomography to visualize the internal microstructure (including cracks). The results indicate that single delta-phase zirconium hydride pellets with various sizes have been successfully produced.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Characterization of Fuel and Cladding In and Near the Pellet-Pellet Gap of a High-Burnup Pressurized Water Reactor Fuel Rod

At TopFuel 2021, we reported on metallographic examinations of a high burnup ZIRLO-clad spent fuel rod operated at the North Anna Power Station having a 3-mm gap between pellets at an elevation of ~1,403 mm. The rod was sectioned axially at that elevation to reveal the pellet-pellet interfaces and the pellet-pellet gap. Sections were mounted and polished to reveal the distribution of cladding inner and outer oxide layer thicknesses above, below, and within the gap. Since then, total cladding hydrogen measurements have been performed to quantify the cladding hydrogen content through the gap and above and below the pellet-pellet gap region. This paper provides the cladding hydrogen concentrations and discusses the overall findings regarding cladding waterside oxidation and cladding hydrogen diffusion to a pellet-pellet gap.

Montgomery, Rose↗
DOE OSTI2023

Effect of hydrogen on plasticity of $α$-Fe: A multi-scale assessment

A multi-scale study was carried out to quantify the effect of interstitial hydrogen concentration on plasticity in $α$-Fe. Here, in this work, the influence of hydrogen on the screw dislocation glide behavior was examined across several length-scales. The insights obtained were integrated to provide an accurate continuum description for the effect of hydrogen on the dislocation based plasticity in polycrystalline $α$-Fe. At the outset of this work, a new Fe—H interatomic potential was formulated that enhanced the atomistic estimation of the variation in dislocation glide behavior in presence of hydrogen. Next, the dislocation core reconstruction observed due to the addition of hydrogen using atomistic simulations was validated with the help of large-scale DFT calculations based on the DFT-FE framework. Several atomistic simulations were carried out to comprehensively quantify the effect of hydrogen on the non-Schmid behavior exhibited during the dislocation glide in $α$-Fe. Finally, crystal plasticity simulations were carried out to understand the effect of hydrogen on the meso-scale deformation behavior of polycrystalline $α$-Fe.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Dynamic formation of preferentially lattice oriented, self trapped hydrogen clusters

Abstract A series of MD and DFT simulations were performed to investigate hydrogen self-clustering and retention in tungsten. Using a newly develop machine learned interatomic potential, spontaneous formation of hydrogen platelets was observed after implanting low-energy hydrogen into tungsten at high fluxes and temperatures. The platelets formed along low miller index orientations and neighboring tetrahedral and octahedral sites and could grow to over 50 atoms in size. High temperatures above 600 K and high hydrogen concentrations were needed to observe significant platelet formation. A critical platelet size of six hydrogen atoms was needed for long term stability. Platelets smaller than this were found to be thermally unstable within a few nanoseconds. To verify these observations, characteristic platelets from the MD simulations were simulated using large-scale DFT. DFT corroborated the MD results in that large platelets were also found to be dynamically stable for five or more hydrogen atoms. The LDOS from the DFT simulated platelets indicated that hydrogen atoms, particularly at the periphery of the platelet, were found to be at least as stable as hydrogen atoms in bulk tungsten. In addition, electrons were found to be localized around hydrogen atoms in the platelet itself and that hydrogen atoms up to 4.2 Å away within the platelet were found to share charge suggesting that the hydrogen atoms are interacting across longer distances than previously suggested. These results reveal a self-clustering mechanisms for hydrogen within tungsten in the absence of radiation induced or microstructural defects that could be a precursor to blistering and potentially explain the experimentally observed high hydrogen retention particularly in the near surface region.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2023

Assessing the influence of microstructure on uranium hydride size distributions via small angle neutron scattering

Here, the effect of microstructure on the internal hydriding behavior of both cast (1 mm grain size) and rolled (25 μm grain size) uranium containing hydrogen concentrations between 0 and 1.8 wppm were evaluated via small angle neutron scattering (SANS). Increasing hydrogen content up to 1.8 wppm in the cast uranium only weakly affected the average uranium hydride (UH 3 ) precipitate size, calculated from the SANS data. Conversely, the UH 3 phase fraction was found to strongly depend on the hydrogen content in the same cast samples. A substantially reduced UH 3 particle size distribution was observed in the rolled uranium relative to cast uranium containing the same nominal hydrogen content. It is hypothesized that the suppression of UH 3 formation in the rolled uranium is driven by increased hydrogen trapping at grain boundaries, and theoretical calculations that account for trap density, potency, and hydrogen diffusion kinetics support this hypothesis.

36 MATERIALS SCIENCE↗
DOE OSTI2022

UV‐induced degradation of high‐efficiency silicon PV modules with different cell architectures

Abstract Degradation from ultraviolet (UV) radiation has become prevalent in the front of solar cells due to the introduction of UV‐transmitting encapsulants in photovoltaic (PV) module construction. Here, we examine UV‐induced degradation (UVID) in various commercial, unencapsulated crystalline silicon cell technologies, including bifacial silicon heterojunction (HJ), interdigitated back contact (IBC), passivated emitter and rear contact (PERC), and passivated emitter rear totally diffused (PERT) solar cells. We performed UV exposure tests using UVA‐340 fluorescent lamps at 1.24 W·m −2 (at 340 nm) and 45°C through 4.02 MJ·m −2 (2000 h). Our results showed that modern cell architectures are more vulnerable to UVID, leading to a significant power decrease (−3.6% on average; −11.8% maximum) compared with the conventional aluminum back surface field (Al‐BSF) cells (<−1% on average). The power degradation is largely caused by the decrease in short‐circuit current and open‐circuit voltage. A greater power decrease is observed in bifacial cells with rear‐side exposure compared with those with front‐side exposure, indicating that the rear side is more susceptible to UV damage. Secondary ion mass spectroscopy (SIMS) confirmed an increase in hydrogen concentration near the Si/passivation interface in HJ and IBC cells after UV exposure; the excess of hydrogen could result in hydrogen‐induced degradation and subsequently cause higher recombination losses. Additionally, surface oxidation and hot‐carrier damage were identified in PERT cells. Using a spectral‐based analysis, we obtained an acceleration factor of 5× between unpackaged cells (containing a silicon nitride antireflective coating on the front) in the UV test and an encapsulated module (with the front glass and encapsulant blocking 90% of the UV at 294 nm and 353 nm, respectively) in outdoor conditions. From the analytical calculations, we show that a UV‐blocking encapsulant can reduce UV transmission in the module by an additional factor of ~50.

14 SOLAR ENERGY↗
DOE OSTI2025

Recyclable Design for Retaining High Solar Absorptivity of the Media in CSP

Efficient thermal energy storage is pivotal to lowering the levelized cost of electricity (LCOE) for Concentrating Solar Power (CSP) plants. In solid-particle systems, however, prolonged high-temperature service degrades particle solar absorptivity, eroding overall efficiency. This project demonstrates a hydrogen-assisted recovery process that reliably restores absorptivity to >90 %, offering a practical route to sustain long-term CSP performance. Bench-scale investigations mapped the reduction kinetics of optically faded particles across hydrogen concentrations, temperatures, and residence times. Coupling mass-spectrometric monitoring with machine-learning optimization minimized energy demand while maximizing absorptivity gain. The resulting process window—moderate hydrogen partial pressures, 15–30 min dwell times, and temperatures well below initial calcination levels—cuts energy consumption well below that of incumbent re-blackening methods. A prototype recovery reactor processed multiple 2 kg batches with repeatable outcomes, confirming scalability and operational robustness. Integrated techno-economic analysis indicates material and operating cost reductions exceeding 15 % relative to conventional particle replacement or chemical re-coating, translating directly into lower LCOE for next-generation CSP facilities. By uniting fundamental reaction-kinetics insight with pragmatic engineering, this work advances the solid-particle pathway, delivering a cost-effective, field-deployable solution to one of CSP’s key durability challenges and strengthening the commercial outlook for high-temperature renewable power.

14 SOLAR ENERGY↗
DOE OSTI2021

Management of the Three Mile Island, Unit 2, Accident Corium and Severely Damaged Fuel Debris

The Three Mile Island, Unit Two (TMI-2) pressurized water reactor core melted down in 1979 due to an untimely combination of maintenance problems that led to a loss of feedwater, followed by a series of operational misunderstandings and errors. The melted core recovery process required development of a wide array of tools. After approximately three years of water management and other cleanup actions, the first views of the core revealed a much higher degree of damage than previously expected. Approximately 62 metric tons of the core had melted, leaving only 42 of the 177 fuel assemblies standing with a majority of rods intact. The core to be recovered was composed of loose debris and a solidified mass of formerly molten fuel. Robotic tools that had been designed for the task were of limited effectiveness due to the range of material types and phases. Over a period of several years, the central melt was broken up, primarily by use of a core-boring drill originally designed to acquire samples through the depth of the debris field. The broken pieces were loaded into specially designed debris canisters by use of long-handled pick-and-place tools and suctioned into baffled knockout canisters using an airlift vacuum system. A remotely operable underwater plasma arc torch was developed for removal of the lower core support structure to be able to retrieve the fuel pieces and secondary melt that had accumulated on the bottom reactor vessel head. Canister design played a central role in defining the initial retrieval process and later affected transportation, interim wet and its current interim dry storage. Due to concerns about the potential for radiolysis of residual water in the debris and other canister material, the canisters were fitted with hydrogen recombiner catalyst units to prevent a buildup of flammable gas and potential pressurization. To confirm the effectiveness of the recombiners during shipping, eight dewatered canisters were kept sealed for up to 205 days with periodic sampling of the headspace gas. The highest hydrogen value (9 vol%) was observed in a canister held for 147 days while the longest stored canister resulted in a 5% hydrogen concentration. The oxygen concentration never exceeded 0.5 %, and the primary backfill was >80% argon, meaning there was no flammability risk. Radiolytic hydrogen was also observed in wet pool storage, where the vented canisters discharged a portion of the water backfill as a result of gas production. The 344 canisters (268 fuel debris, 62 filter, and 12 knockout type) were dewatered, loaded in groups of seven into a double barrier shipping cask, and transported to the U.S. Department of Energy site in Idaho for 10 years of pool storage.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗