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DOE OSTI2025

Multiscale, mechanistic modeling of cesium transport in silicon carbide for TRISO fuel performance prediction

Understanding cesium (Cs) transport in TRistructural ISOtropic (TRISO) particle fuel is crucial for predicting fission product release in high-temperature reactors. However, current challenges include significant scatter in diffusivity data and unexplained temperature-dependent diffusion regimes in the silicon carbide layer. This study addresses these challenges by developing a multiscale, mechanistic Cs transport model integrating atomistic simulations and phase field modeling. Our model quantifies temperature and grain size effects on Cs diffusivity, attributing experimentally observed regimes to a transition from bulk-dominated diffusivity at high temperatures to grain boundary-dominated diffusivity at lower temperatures. The model, validated against diffusion measurements and advanced gas reactor (AGR)-1 and AGR-2 post-irradiation fission product release data, enhances the predictive capability of the BISON fuel performance code. This study advances our understanding of Cs release from TRISO particles and its dependence on temperature and silicon carbide grain size, with implications for the safety and efficiency of high-temperature nuclear reactors.

BISON↗

DOE OSTI2022

Silicon Photomultipliers Coupled to Scintillators With the Emission Maximum at 550 nm

A majority of the silicon photomultipliers (SiPMs) are sensitive to blue and near-ultraviolet (NUV) photons that are not optimized for scintillators with the emission maximum at wavelengths longer than 500 nm. The red–green–blue (RGB) SiPM is developed for the maximum photon detection efficiency (PDE) at 550 nm, which is a good match for some high-light-yield scintillators, such as CsI:Tl (CsI) and Gd 1.5 Y 1.5 Ga 2 Al 3 O 12 :Ce (GYGAG). Comparisons are made for the performance of these scintillators coupled to the RGB SiPM and a blue-sensitive SiPM. Because it takes tens of nanoseconds for the microcells to recharge after registering a photon hit, the linearity of these scintillation detectors was studied for high-energy gammas where numerous scintillation photons are generated. In addition, the energy resolution of the 662-keV gamma emitted by 137 Cs was measured for temperatures between –20 °C and 50 °C. The nonlinearity was observed above 1 MeV for all measurements, and however, it can be corrected by energy calibration using a third-degree polynomial. For CsI, the energy resolution is better with the blue-sensitive SiPM because of the lower dark count rate (DCR). In contrast, GYGAG coupled to the RGB SiPM has a better energy resolution for temperatures below 30 °C because of the well-matched emission spectrum and PDE distribution. Nevertheless, the advantage disappears for temperatures above 30 °C due to the higher DCR. It would be useful to further develop the RGB SiPM with a lower DCR and higher operating temperatures.

42 ENGINEERING↗

DOE OSTI2020

Cesium Removal Performance Comparisons of Crystalline Silicotitanate Media Batches with Savannah River Site Waste Simulant

The Tank Closure Cesium Removal (TCCR) system uses ion exchange columns filled with Crystalline Silicotitanate (CST) media to process radioactive waste solutions for the removal of Cs137. TCCR currently focuses on dissolving Savannah River Site (SRS) Tank 10H waste (primarily sodium salt cake solids) within the tank followed by at-tank ion exchange column treatment. Plans are underway to prepare and install a second TCCR unit at SRS. Capacity and particle size differences exist between archived (IE-911) and more recently-prepared CST media batches (9120- B and 9140-B). Side-by-side comparison testing was performed to evaluate the cesium removal performance of each batch to aid in selecting the preferred CST batch and media characteristics to load into the second TCCR unit. Batch contact equilibrium and flow-through column tests have been conducted with three CST batches using an SRS Average Simulant. Simulant batch contact equilibrium cesium loading results for the three CST batches (including two lots of one batch) are provided in Table ES-1. The 35 °C data indicates that the archived IE-911 batch has a higher cesium capacity than recently-prepared CST media and that the minor TCCR CST 9120-B media lot (2099000035) has similar cesium removal performance to the major TCCR lot (2099000034). Tests conducted at 25 ºC for the archived IE-911 CST batch indicated lower cesium removal performance with this simulant batch than was observed recently with a different SRS Average Simulant batch. A dilution factor (DF) is typically utilized when modeling engineered CST media cesium loading performance to account for mass contributions from the binder material. In cases where CST performance is lower than expected, this factor includes corrections for the binder and for low performance. The ZAM (Zheng, Anthony, and Miller) Isotherm Model DF values are provided in Table ES-1 for each simulant batch contact result. DF values near 0.5 were determined for the 9120-B and 9140-B CST batch contact tests while DF values near 0.6 (20% higher) were determined for the tests with IE-911. These dilution factors are lower than recently observed with a different SRS Average Simulant batch (9140-B DF = 0.68; IE-911 DF = 1.0). Based on these results, it appears that some component in the simulant solution used for equilibrium testing may have resulted in reduced cesium loading on the CST media.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2022

Design and characterization of hierarchical aluminosilicate composite materials for Cs entrapment: Adsorption efficiency tied to microstructure

The growing quantity of nuclear waste and the serious threats to the environment challenge researchers to innovate and target new waste form technologies. In the past decades, considerable efforts have been devoted to developing highly selective sorbents followed by safe disposal with the assurance of chemical stability and robust retention performance. Zeolite-containing geopolymers are regarded as a possible 2-in-1 material able to both capture and sequester elements such as Cs in bed fixed column application perspective. Here, these composites show promise for combining extraction properties of zeolite powder due to its crystalline structure (high capacity and selective adsorption), with the tunable microstructure and the shaping feasibility of the geopolymer binder. For the development of materials devoted to Cs immobilization, porous zeolite/geopolymer composites were prepared by dispersing NaY zeolite particles in a geopolymer binder. The influence of the structural properties of such composites on their ability to entrap a large amount of Cs by an ionic exchange process was notably studied. Composites' compositions, porosities, morphologies and crystallinity were analyzed by scanning electron microscopy coupled with energy dispersive x-ray spectroscopy (SEM-EDX), x-ray diffraction analysis (XRD) and nitrogen adsorption/desorption studies. Experimental Cs sorption in batch mode was used to follow the ionic exchange phenomenon in these materials. Along with 5 wt% amount of zeolite in geopolymer improves the Cs adsorption performance offering multiple new adsorption sites. Additionally, the geopolymer mesopores are beneficial facilitating the access of Cs and its role as a binder is advantageous to tailor granular hierarchical structure for safer industrial application.

36 MATERIALS SCIENCE↗

DOE OSTI2022

Engineering Evaluation of Barium Buildup in a Decayed CsCl Sealed Source and Potential Impact for Cesium Release from a Breached Source

For more than 50 years, radioactive 137 Cs has been a major source material for radioactive sealed sources, usually constructed as cesium chloride (CsCl) salt loaded into double-walled, stainless-steel capsules. A complication develops as 137 Cs decays to 137 Ba since this process creates a strongly reducing environment inside the capsule. A potential hazard exists if the capsule is breached and air ingress induces rapid exothermic oxidation, and this mechanism is suspected to be responsible for the well-known contamination incident at the Harbor View facility in Washington state. For this study, many thermodynamic evaluations were performed to assess the internal state of capsules after several decades of decay and to describe the potential oxidation if the capsule contents were to suddenly be exposed to air. Results suggest that reduction of impurities such as Cu, Fe, Pb, and Cr to metal will occur. If these impurities are lacking, it is possible that even Ba metal will form. In most cases, rapid oxidation can occur, and the exothermic reactions are sufficient to vaporize portions of the contents, which would include the remaining 137 Cs.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2020

ZAM Modeling Study to Support the Tank Closure Cesium Removal (TCCR) 1A Unit

Currently at the Savannah River Site (SRS), the Tank Closure Cesium Removal (TCCR) is an “at-tank” process designed to remove cesium from aqueous tank waste. Cesium will be removed by ion exchange using the engineered IONSIV® R9120 form of Crystalline Silicotitanate (CST) media. The current TCCR design has two columns online in a lead-lag configuration to optimize media usage and achieve the target decontamination. Once the lead column is saturated with cesium, it will be removed from service, the lag column will rotate into the lead position, and a new column with fresh ion-exchange media will be placed into the lag position. The TCCR process for cesium removal from Tank 10H is detailed in X-SOW-H- 00002. Demonstration of the system began in early calendar year 2019 with two batches of salt solution generated by dissolving saltcake in Tank 10H, followed by processing of these batches through the TCCR system. A third TCCR Tank 10H dissolved saltcake batch is scheduled for processing soon. Upon completion of the demonstration with Tank 10H dissolved saltcake, Tank 9H salt solution will be transferred to Tank 10H and subsequently processed through the TCCR unit with new CST media (referred to as R9120-B 30x60) added to new IX columns. The TCCR processing campaign of Tank 9H salt solution is referred to as TCCR-1A .

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2020

CsI-Antisolvent Adduct Formation in All-Inorganic Metal Halide Perovskites

The excellent optoelectronic properties shown by hybrid organic/inorganic metal halide perovskites are all predicated on precisely controlling the exact nucleation and crystallization dynamics that occur during film formation. In general, high-performance thin films are obtained by a method commonly called solvent engineering (or antisolvent quench) processing. The solvent engineering method removes excess solvent, but importantly leaves behind solvent that forms chemical adducts with the lead-halide precursor salts. These adduct-based precursor phases control nucleation and the growth of the polycrystalline domains. There has not yet been a comprehensive study comparing the various antisolvents used in different perovskite compositions containing cesium. In addition, there have been no reports of solvent engineering for high efficiency in all-inorganic perovskites such as CsPbI 3 . In this work, inorganic perovskite composition CsPbI 3 is specifically targeted and unique adducts formed between CsI and precursor solvents and antisolvents are found that have not been observed for other A-site cation salts. These CsI adducts control nucleation more so than the PbI 2 -dimethyl sulfoxide (DMSO) adduct and demonstrate how the A-site plays a significant role in crystallization. The use of methyl acetate (MeOAc) in this solvent engineering approach dictates crystallization through the formation of a CsI-MeOAc adduct and results in solar cells with a power conversion efficiency of 14.4%.

14 SOLAR ENERGY↗

DOE OSTI2020

Ion-Exchange Modeling of Crystalline Silicotitanate for Cesium Removal - 20283

The Tank Closure Cesium Removal (TCCR) system is a Savannah River Site (SRS) demonstration 'at-tank' process designed to remove {sup 137}Cs from the high-level aqueous tank waste so that the decontaminated solution can be disposed as low-level waste. Cesium is removed by ion exchange (IX) columns using engineered IONSIV{sup TM} R9120-B form of the Crystalline Silicotitanate (CST) media. The TCCR system is deployed at Tanks 10 and Tank 11 in the SRS H Tank Farm. Water is added to the salt-cake in Tank 10 H to dissolve it. The dissolved salt solution waste is pumped out of Tank 10H (feed tank), through filters and IX columns. The decontaminated salt solution is transferred to Tank 11 (receipt tank), and on to Tank 50H for final disposal in the Saltstone Production Facility. The current TCCR can accommodate lead-lag (two-column) or lead-lag-guard (three-column) configurations to optimize media utilization and achieve the target decontamination. To assist the TCCR operations, a parametric study was conducted to evaluate the impact of different parameters (e.g., column configurations (single column, two columns or three columns in series), waste characteristics, operating temperature, process flow rate, CST average particle size) on the IX column performance including CST bed utilization. The initial results indicate that the IX column performance is improved at slower process flow rate, at lower operating temperature, and with smaller CST average particle size. Multi-column configurations are recommended, because the single-column configuration does not utilize CST bed effectively. This paper demonstrates the versatility of the ion exchange modeling to evaluate the effects of CST characteristics and operational parameters on IX column performances. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2020

Los Alamos National Laboratory Release of Cesium-137 Corrective Action Plan

This Corrective Action Plan (CAP) was developed by a team chartered by John Sarrao, Deputy Director for Science, Technology, and Engineering (DDSTE), to review and respond to the findings of the Joint National Nuclear Security Administration/Triad National Security (NNSA/Triad) Investigation Team (JIT) Joint Investigation Report for the “Sealed Source Recovery at the University of Washington Harborview Training and Research Facility Results in Release of Cesium-137 on May 2, 2019,” dated March 30, 2020. On May 2, 2019, International Isotopes, Inc. (INIS), a subcontractor to Triad National Security, LLC, inadvertently breached a sealed cesium-137 source at the University of Washington (UW), Harborview Medical Center Research and Training (HRT) building in downtown Seattle while attempting to recover the source for the National Nuclear Security Administration (NNSA) Off-Site Source Recovery Program (OSRP). The source breach resulted in contamination of personnel and the HRT building, as well as a release of radioactive material to the environment

07 ISOTOPE AND RADIATION SOURCES↗

DOE OSTI2020

GeoMelt{sup R} In-Container Vitrification (ICV){sup TM} for Fukushima Daiichi Water Treatment Secondary Wastes - 20212

The Japanese government is supporting development work implemented jointly by Veolia subsidiaries Kurion Japan, K.K., Veolia Nuclear Solutions (VNS), Inc., and Veolia Nuclear Solutions Federal Services, LLC for treating radioactive waste generated from Fukushima Daiichi Nuclear Power Station (NPS) water treatment using the GeoMelt{sup TM} In-Container Vitrification (ICV){sup TM} technology. The initial work consisted of glass formulation and engineering-scale testing which was completed in 2018, in the frame of an IRID (International Research Institute for Nuclear Decommissioning) program as part of a project subsidized by Japan's Ministry of Economy, Trade and Industry (METI). The Fukushima Daiichi NPS Mid- and Long-Term Road-map requires investigation of methods to stabilize solid wastes (and to immobilize radioisotopes in the wastes) generated as a result of emergency response and decommissioning activities. Cooling water treatment has resulted in a significant amount of solid and slurry secondary wastes (mostly adsorbents and ion-exchange materials) which will require processing at some point. GeoMelt{sup R} ICV{sup TM} is a joule-heated melter technology which uses a refractory-lined single-use container combining the melter and disposal container. There is no pouring required nor concerns with refractory corrosion which allows the process to accommodate a wide range of waste chemistries and high waste loadings. The testing described here consisted of three engineering-scale melts, each processing between 212 kg and 240 kg of waste simulants, glass formers, and non-radioactive cesium (Cs) and strontium (Sr) tracers. Continuous isokinetic stack sampling of off-gas emissions was performed for each test in order to calculate Cs and Sr retention in the glass wasteform. Single-pass retention of Cs in the final glass wasteform ranged from 91.46 to 99.30%, and single-pass retention of Sr ranged from 99.76 to 100%. Planned particulate recycle will increase these retention levels. Melt 1 processed a mixture of KUR-EH (a zeolite-based ion-exchange material), simulated Advanced Liquid Processing System (ALPS) Carbonate and Iron Slurries, and glass additives. Melt 2 processed a mixture of KUR-EH, KUR-TSG (a titanate-based adsorbent), and glass additives. Melt 3 processed a mixture of KUR-EH, simulated barium sulfate/iron ferrocyanide sludge (AREVA sludge), and glass additives. Waste loadings for these melts ranged from 70 weight percent (wt%) to 82 wt%. Vitrification produces a waste form much denser than the stored water treatment secondary waste wastes, resulting in significant volume reduction. Volume reductions for the three tests ranged from 74 to 79 vol%. Vitrification produces a chemically durable wasteform. Pacific Northwest National Laboratory (PNNL) tested three glass samples from each engineering-scale melt) by the Materials Characterization Center 1 (MCC-1) test, an international standard leach test of the chemical durability of nuclear waste glasses. PNNL also obtained one U.S. reference glass (EA Glass) and two Japan reference glasses P0798) and tested these under the same MCC-1 conditions (90 deg. C, 10 m-1, DIW, and 7, 14, 28-day) as the GeoMelt{sup R} ICV{sup TM} glasses. The GeoMelt{sup R} ICV{sup TM} glasses exhibited lower total normalized releases and 14- to 28-day normalized release rates than the three reference glasses. These results suggest that the GeoMelt{sup R} ICV{sup TM} glasses have durabilities on par with high-level waste glasses under standard test conditions. Post-melt process sampling and analysis indicated no Cs migration into the melter refractory materials and very little deposition of Cs or Sr onto the melter hood or off-gas piping internals. The results of the testing indicated good Cs retention in the glass, high volume reduction and waste loadings, and excellent chemical durability. These factors are important to minimize treatment costs and to protect workers and the environment. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2022

Non‐Traditional Positively‐Biased Narrow‐Band Perovskite Single‐Crystal Photodetectors Enabled by Interfacial Engineering

Abstract The positive bias in theory narrows down the depletion region and thus results in significant charge injection, which should be detrimental to charge generation and collection performance for traditional photodetectors. Here, instead, it is found that the external quantum efficiency (EQE) is increased by more than 50 times when the photodetector is positively biased. A positive bias of +6 V drives ion migration of Br − and Cs + towards the anode and cathode, respectively, leading to self‐doping within bulk single crystals to form an advantageous p‐i‐n junction for better charge collection in the devices. Meanwhile, the injected holes are allowed to tunnel through the cesium lead bromide/fullerene interface to reach the cathode which also significantly contributes to the enhancement of EQE in the forward‐biased devices. The positively‐biased narrow‐band (full width at half maxima (FWHM) = 16 nm) photodetectors exhibit a specific detectivity of 6.5 × 10 10 Jones at 550 nm, along with the −3 dB cutoff frequency of 2776 Hz. By manipulating charge injection and ion migration using interfacial engineering, a class of non‐traditional, positively‐biased, and highly narrow‐band photodetectors is demonstrated, which offers an alternative design strategy for imaging, biosensing, automatic control, and optical communication.

Li, Junchi↗

DOE OSTI2020

Characterization and CST Batch Contact Equilibrium Testing of Modified Tank 9H Process Supernate Samples in Support of TCCR

The Tank Closure Cesium Removal (TCCR) system uses ion exchange columns filled with crystalline silicotitanate (CST) media to process radioactive waste solutions for the removal of ¹³⁷Cs. TCCR currently focuses on dissolving Savannah River Site (SRS) radioactive tank waste (primarily sodium saltcake solids) within Tank 10H followed by at-tank ion exchange column treatment. Two supernate batches from Tank 10H have been processed through the TCCR unit and processing of a third batch is expected soon. After processing of this supernate batch, plans are to replace the CST columns and process dissolved salt solution from Tank 9H through Tank 10H and then through the new CST columns installed in the TCCR unit. The new columns are expected to contain either a media similar to an archived CST batch (IE-911) or the R9120-B CST media used in the current TCCR columns (the two materials are fundamentally the same; just different specifications, product names, and preconditioning steps). Samples of Tank 9H dissolved saltcake were received at the Savannah River National Laboratory (SRNL) and characterized. The Tank 9H supernate contained a high sodium concentration (~9.6 M Na⁺) and will require dilution to near 6 M [Na⁺] prior to processing through the TCCR unit. The cesium concentration in Tank 9H is currently significantly higher than was observed with Tank 10H. Three dilutions of the Tank 9H supernate were conducted to mimic possible dilutions that could be conducted in the tank farm prior to TCCR processing using inhibited water, sodium hydroxide, and sodium nitrate solutions. Dilution #1 was prepared by diluting the Tank 9H supernate by a factor of 1.6 with inhibited water, Dilution #2 was prepared by diluting the Tank 9H supernate by a factor of 3.7 with a mixed sodium hydroxide/sodium nitrate diluent, and Dilution #3 was prepared by also diluting by a factor of 3.7 but with sodium hydroxide only. All three dilutions have similar Na⁺ concentrations (~6 M), but Dilutions #2 and #3 contain significantly less cesium. Major and key minor components of the diluted Tank 9H supernate samples are provided in Table ES-1. Batch contact equilibrium tests were conducted with the Tank 9H dilutions and the two different CST media batches being considered for use in the new TCCR columns. Results are summarized in Table ES-2 and compared to ZAM model predictions. The highest cesium distribution coefficient and percent removal were observed with Dilution #3. IE-911 CST (an archived CST media batch) was more effective at removing cesium than the more recently prepared R9120-B CST media, though both media samples removed >88% of the cesium and the differences may not be statistically different considering the overall uncertainty. Based on the results, maximum cesium loadings were calculated for each CST media type and Tank 9H dilution. In general, maximum cesium loadings from dilutions of this supernate batch are quite high (approaching 0.1 mmol total Cs⁺/g CST for Dilution #1). The highest calculated maximum ¹³⁷Cs loading for the Tank 9H dilutions using the ZAM model with input of the tank compositions and batch contact results was 207 Ci/kg CST (Dilution #1 with IE-911 CST). In all cases, higher maximum cesium loading values were predicted for IE-911 CST versus R9120-B, although the differences varied considerably between the dilutions. The maximum loading value for IE-911 CST with Tank 9H Dilution #1 was only 7% higher than the maximum loading for R9120-B. The maximum loading value for IE-911 CST with Dilution #3 was 24% higher than the maximum loading for R9120-B. Dilution #2 was intermediate between these values. Note that these maximum loading values are the theoretical calculated values, and actual operating conditions will cause differences. A CST binder dilution (correction) factor near 0.7 (relative to pure powder CST) was required for each batch contact test with IE-911 engineered CST using the three Tank 9H dilutions. Correction factors calculated for R9120-B CST ranged from 0.56 to 0.66 for the three dilutions. Following the batch contact tests, the CST samples were isolated from the Tank 9H solution, washed, dried, and digested in acid following established procedures. The analysis results are provided in Table ES-3. Total cesium loading values determined by CST digestion were similar to the calculated loading values based on solution analysis (Table ES-2) for all samples. In addition, the CST was observed to load calcium (R9120-B sample only), iron, strontium, lead, uranium, and plutonium after contact with the waste supernate, as has been observed previously.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2020

Life cycle Management of 324 Project High-Activity, Mixed/Low-Level Radioactive Waste - 20506

U.S. DOE contractor CH2M Hill Plateau Remediation Company (CHPRC) manages the 324 Building and is preparing to remotely excavate and disposition the radioactive soil beneath the building to allow further deactivation of the building. The 324 Building is a non-reactor Category 2 Nuclear Facility located in the 300 Area of the Hanford Site. Records indicate that in October 1986, approximately 516 liters of a concentrated liquid waste stream containing cesium-137 ({sup 137}Cs) and strontium-90 ({sup 90}Sr) spilled onto the floor of a hot cell, B-Cell, in the 324 Building. The spill contained an estimated 1.3 million curies (Ci) of radioactivity. A breach in the sump of B-Cell was discovered in November 2009 during characterization of the soil under the building. Exposure rates in excess of 10,000 R/hr were detected through a system of access pipes installed under the hot cell, and an estimated 1.557 E+05 Ci of {sup 137}Cs and 6.842 E+04 Ci of {sup 90}Sr are in the soil beneath B-Cell. The magnitude of the soil contamination below B-Cell represents one of the most challenging remediation activities in the DOE complex. The objective of CHPRC's 324 Project is to remotely excavate and disposition the highly contaminated soil under the hot cell so building deactivation and stabilization activities can resume. Depending on the volume and radioactivity of contaminated soil under B-Cell, the material will be dispositioned either at the Hanford Site's Environmental Restoration Disposal Facility (ERDF) or in grout monoliths that will be created in the hot cells (A, C and D) next to B-Cell during future building demolition activities. The current project scope includes installing structural supports under the walls of B-Cell and using remotely operated equipment to remove debris from the hot cell, remove debris mixed with grout on the floor, saw through and remove the cell floor and liner, and excavate contaminated soil under B-Cell to a depth of up to 3.66 meters (12 feet). Addition or removal of debris in A-Cell may be required to support B-Cell cleanout and excavation activities. The base approach is to disposition debris and soil bins in the Radiochemical Engineering Cell (REC) monoliths, and subsequently disposition the monoliths during building demolition. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2024

Ag Intercalation in Layered Cs 3 Bi 2 Br 9 Perovskite for Enhanced Light Emission with Bound Interlayer Excitons

Cesium bismuth bromide (CBB) has garnered considerable attention as a vacancy-ordered layered perovskite with notable optoelectronic applications. However, its use as a light source has been limited due to its weak photoluminescence (PL). Here, we demonstrate metal intercalation as a novel approach to engineer the room-temperature PL of CBB using experimental and computational methods. Ag, when introduced into CBB, occupies vacant sites in the spacer region, forming octahedral coordination with surrounding Br anions. First-principles density functional theory calculations reveal that intercalated Ag represents the most energetically stable Ag species compared to other potential forms, such as Ag substituting Bi. The intercalated Ag forms a strong polaronic trap state close to the conduction band minimum and quickly captures photoexcited electrons with holes remaining in CBB layers, leading to the formation of a bound interlayer exciton, or BIE. The radiative recombination of this BIE exhibits bright room-temperature PL at 600 nm and a decay time of 38.6 ns, 35 times greater than that of free excitons, originating from the spatial separation of photocarriers by half a unit cell separation distance. The BIE as a new form of interlayer exciton is expected to inspire new research directions for vacancy-ordered perovskites.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2022

High‐Performance and Stable Semi‐Transparent Perovskite Solar Cells through Composition Engineering

Abstract Semi‐transparent perovskite solar cells (ST‐PeSCs) have tremendous potential as solar windows owing to their higher efficiency and visible transmittance. However, studies toward this application are still nascent, particularly in unraveling the interplay between how the perovskite composition impacts the achievable device performance and stability. Here, the role of A‐ and X‐site modification in APbX 3 perovskites is studied to understand their influence on these factors. Through detailed experimental and simulation work, it is found that a perovskite composition consisting of cesium (Cs) and formamidinium (FA) at the A‐site delivers the best device performance over a range of band gaps, which are tuned by changes to the X‐site anion. Using this optimized perovskite composition, power conversion efficiencies of 15.5% and 4.1% are achieved for ST‐PeSCs with average visible transmittance values between 20.7% and 52.4%, respectively. Furthermore, the CsFA‐based ST‐PeSCs show excellent long‐term stability under continuous illumination and heating. The stability of the precursor solutions across each of the studied compositions has also been considered, showing dramatic differences in the structural properties of the perovskites and their device performance for all mixed A‐site compositions possessing the archetypal methyl ammonium species, while also confirming the superior stability of the CsFA precursor solutions.

14 SOLAR ENERGY↗

DOE OSTI2019

EGS Collab Testbed 1: Second Set Tracer Test Results

The EGS Collab project is developing ~10-20 m-scale field sites where fracture stimulation and flow models can be validated against controlled, small-scale, in-situ experiments. The first multi-well experimental site was established at the 4850 level in the Homestake Mine in Lead, South Dakota, where hydraulic fractures were created at an injection well drilled sub-horizontal from the drift. This file contains the second set of tracer data (also include the data uploaded previously, https://gdr.openei.org/submissions/1128) for the EGS Collab testbed. The tracer tests were conducted during October 2018 - November 2019. Injected tracers include DNA, C-dots (fluorescein nano particles), fluorescein, rhodamine-b, sodium chloride, lithium bromide and cesium iodine. The tracers have been detected in three flowing wells located about 7.5 to 9 meters away from the injection interval. The tracer breakthrough curves from these locations have been adjusted to account for the residence time in the injection and production tubing. The details about the tracer test can be found in Background and Methods of Tracer Tests (Mattson et al. (2019,a,b)) (also included in this package). References Mattson, E.D., Neupane, G., Plummer, M.A., Hawkins, A., Zhang, Y. and the EGS Collab Team 2019a. Preliminary Collab fracture characterization results from flow and tracer testing efforts. In Proceedings 44th Workshop on Geothermal Reservoir Engineering, edited, Stanford University, Stanford, California. Mattson, E.D., Neupane, G., Hawkins, A., Burghardt, J., Ingraham, M., Plummer, M., and the EGS Collab Team, 2019b. Fracture tracer injection response to pressure perturbations at an injection well. GRC Transactions, Vol. 43, 2019. Neupane, Ghanashyam, Earl Mattson, Adam Hawkins, Mitchell Plummer, and Yuran Zhang. EGS Collab Testbed 1: Tracer data sets. No. 1128. DOE Geothermal Data Repository; Idaho National Laboratory, 2019.

15 GEOTHERMAL ENERGY↗

DOE OSTI2020

Testing of Crystalline Silicotitanate to Support Tank-Side Cesium Removal System Operations - 20458

The direct feed of waste from the Hanford tank farms to the Low-Activity Waste Facility at the Hanford Waste Treatment and Immobilization Plant requires an intermediate treatment step that prepares the waste for vitrification. Washington River Protection Solutions (WRPS) selected a near-tank treatment system known as the Tank-Side Cesium Removal (TSCR) system to perform the required intermediate treatment functions. The approach in the TSCR system is to filter waste supernatant with a sintered metal dead-end filter and then use a series of ion exchange columns to remove cesium; this system has several similarities to the Tank Closure Cesium Removal system that has been deployed on the Savannah River Site. The ion exchange media proposed for use is crystalline silicotitanate (CST), which is a non-elutable media with a high affinity for cesium. Because the media is non-elutable, loaded columns will be blown down with compressed air, moved into interim storage, and replaced with new columns when cesium capacity is reached. The assumed extent and rate of drying that could be achieved in the TSCR column geometry lacked confirmation by experimental data. In addition, once the columns are loaded with cesium, they generate flammable gases (primarily hydrogen) via radiolysis acting on any resident liquids, i.e., moisture, remaining in the media bed. An assessment of available information on CST concluded that gas generation data that bounds expected operational conditions were needed to support the TSCR system safety basis and planned operations. Working with WRPS, Pacific Northwest National Laboratory (PNNL) designed and conducted testing with CST media to address the need for (a) representative in-column drying data, and (b) bounding flammable gas generation data. Drying testing was performed using a full-height column with a diameter of approximately 2 inches with injected air at temperatures of 18 deg. C and 30 deg. C. Gas generation testing was conducted in an engineered bunker that allowed simultaneous irradiation of up to eight samples at a time. These two experimental approaches used at PNNL for the CST testing are summarized and the major outcomes are presented. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2023

Integrated Process Testing of MSR Salt Spill Accidents

Part of the licensing process for new nuclear reactors requires vendors to assess the potential consequences of identified accident scenarios using accident progression modeling. The accident scenario that will likely be evaluated by all molten salt reactor (MSR) developers is a spill of radionuclide-bearing fuel salt onto the reactor containment floor (i.e., a salt spill accident). The development of accident progression models requires experimental data to inform which processes to incorporate, to enable the calculation of parameters to model these processes, and to validate the model predictions. The data should quantify the sensitivities of key processes (e.g., molten salt spreading, heat transfer, containment structure corrosion, radionuclide vaporization, and aerosol generation) towards the initial conditions of the spill, the ambient environment, and the features of the containment. In addition, results from integrated process tests that quantify coupled processes are required to validate systems-level models. This report documents results from integrated process tests conducted on simulated molten salt spill accidents. The generated experimental data simultaneously quantify the heat transfer behavior of the spilled salt, compositional changes to the bulk salt, and the release of surrogate fission products from the spilled salt as aerosol particles. All tests that were conducted used FLiNaK doped with surrogate fission products, and the variables that were evaluated included the initial salt temperature and the concentration of surrogate fission products present in the salt. The major accomplishments of this work include identifying surrogate fuel salt compositions that provide insight into the dispersal behavior of radionuclides of potential significance to the source term, employing previously developed methods and measurement techniques to simultaneously measure key processes, generating data on the coupled processes of molten salt heat transfer and surrogate fission product release as aerosol particles, demonstrating new test methods for real-time monitoring of the flow rate of the spill and aerosol size quantification in an argon atmosphere, and developing a mass transfer model for cesium and iodine release from molten FLiNaK to provide insight into aerosol formation by vapor condensation. The same methodology applied herein can be employed to study different salt compositions of interest to MSR developers, different environmental conditions, and other variables that are relevant to postulated accident scenarios. The insights gained from these integrated process tests conducted at a laboratory scale will be incorporated into future integral effects tests conducted at an engineering scale.

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