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At least 289 records · Page 16
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Sister Rod Destructive Examinations (FY22) Appendix J: Leaching of High Burnup Used Nuclear Fuel in Deionized Water

The leaching experiment aims to understand the trends in the radiolysis-enhanced dissolution of HBU SNF when exposed to water (e.g., in-reactor or in-pool cladding failures). Specimens from a baseline M5-clad rod and a heat-treated M5 rod were cut from the fractured CIRFT specimens and placed in 100 mL deionized water for a period of 128 days. Both radial and axial sections were cut to provide different surface areas of fuel in contact with the leachate. During the four-month exposure period, aliquot samples of the leachate were analyzed using gamma spectroscopy and inductively coupled plasma - mass spectrometry (ICP-MS). The analysis quantified the amount of fuel leached into the solutions and provided individual isotopic release fractions (of 30+ isotopes) which were compared as a function of time and surface area of the fuel exposed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Optimizing Sub-Sampled STEM Imaging for Beam Sensitive Materials and Dynamic Processes

The goal of this research is to quantify how sub-sampling and inpainting (broadly classified as methods in Artificial Intelligence) can be used to obtain the highest resolution images under any experimental conditions, and to use this to reproducibly control dynamic in-situ processes in gases and liquids on the nanoscale. As atomic diffusion under non-extreme conditions is typically in the sub-ms time domain, the ability to use this sub-sampling approach to reduce the effect of the electron beam and to speed-up the imaging process represents the optimum combination of spatial and temporal resolution achievable in a conventional microscope. This LDRD funding demonstrated the feasibility of imaging dynamics in complex energy systems using this methodology by establishing the optimum experimental sampling conditions (optimizing the number of pixels, speed, contrast and resolution) for each experiment. For in-situ liquid cell experiments this analysis shows that simply by distributing the same dose and dose rate maximally in space and time, the kinetics for radiolysis products can be modified significantly. The final stage of this work, i.e. the experimental demonstration of its use for nucleation and growth, was severely curtailed by issues with COVID19. However, all the experiments are designed and currently underway, with the goal to complete the final demonstration and publication as soon as safety concerns for laboratory access are met.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Benchmarking the AirSquared All-Metal Scroll Pump Against the Normetex 15 - Preliminary report

Tritium operations require specialized equipment for material compatibility, safety, and waste considerations. Non-metallic pump materials (polymer, oil, or grease) on a tritium-wetted surface can absorb tritium and/or undergo isotope exchange between tritium and more common protium, creating tritiated compounds that can result in material property modification via beta-radiolysis, inventory loss, reduced performance and special waste considerations, all of which increase the operating cost for a facility. Vacuum pumps move gas via a pressure differential between volumes of an operation. Several types of all-metal vacuum pumps exist for this application including metal bellows, diaphragm and scroll pumps. A metal bellows pump incorporates only metal components at the wetted surfaces. The common metal bellows pump employed in tritium facilities, the MB601 (Senior Metal Bellows, Sharon, PA, USA) creates a gas-motive force with a vacuum pressure between 30-40 torr. These pumps can efficiently move gas from one process step to another, as long as sub-30 torr pressures are not required. However, for higher vacuum demand, an all-metal scroll pump can be used with the bellows pump serving as a “backing” pump. This configuration typically provides an achievable vacuum of less than 20 mTorr. All-metal scroll pumps are a technology uniquely applicable to tritium gas handling. The industry standard all-metal scroll pump for glovebox-scale tritium operations is the Normetex 15 m 3 /hr, typically referred to as the Normetex 15 (Normetex Pompes, Pont Audemer, France). However, the final Normetex pumps were manufactured in 2012, rendering them no longer available. Eumeca SaRL (formerly Normetex) has developed and is marketing a scroll pump based on the design of the Normetex 15. A US company, AirSquared (Broomfield, Colorado), has developed and is marketing a potential alternative to the Normetex 15. The Gas Transfer Systems group (Q-7) of the Los Alamos National Laboratory (LANL) procured serial number 0001 of the AirSquared All Metal Vacuum Scroll pump model number V16H034A-C01 in 2019. The LANL Hydrogen Processing Laboratory (HPL) has been evaluating the performance of the pump compared to a Normetex pump as a potential replacement in tritium and fusion facilities requiring this type of pump.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Solid-state transformation of uranyl peroxide materials through high-level irradiation

The solid-state transformation of sodium uranyl triperoxide (Na 4 (UO 2 )(O 2 ) 3 ∙9H 2 O, NaUT) to sodium uranyl tricarbonate (Na 4 (UO 2 )(CO 3 ) 3 ) by radiolysis has been observed for the first time. The exposure of NaUT to 3 MGy gamma irradiation resulted in partial breakdown of the peroxides forming a mixed peroxide and carbonate species. The effects of He ion irradiation on NaUT was also investigated up to 225 MGy using both hydrated argon and dry argon. The complete conversion to the uranyl tricarbonate phase by 56 MGy using hydrated argon, while dry argon did not fully convert showing the importance of water in the system. He-ion irradiated NaUT samples all convert to the tricarbonate phase with time in air post radiation exposure. This transition was monitored via Raman spectroscopy, infrared spectroscopy (IR), and powder X-ray diffraction (PXRD) to further confirm the identity of the final product as the sodium uranyl tricarbonate, čejkaite. This transformation outlines a mechanism for the mobility of uranyl in natural environments and in the Hanford tanks.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2024

Quantification of Residual Water in Spent Fuel Dry Storage Canisters Using Mass Spectrometry

The purpose of this report is to document updates on the apparatus to simulate commercial vacuum drying procedures at the Nuclear Energy Work Complex at Sandia National Laboratories. Validation of the extent of water removal in a dry spent nuclear fuel storage system based on drying procedures used at nuclear power plants is needed to close existing technical gaps. Operational conditions leading to incomplete drying may have potential impacts on the fuel, cladding, and other components in the system during subsequent storage and disposal, such as fuel degradation; cladding corrosion, embrittlement, or breaching; and the creation of a flammable environment via radiolysis of water. A general lack of data suitable for model validation of commercial nuclear canister drying processes necessitates well-designed investigations of drying process efficacy and water retention. Scaled tests that incorporate relevant physics and well-controlled boundary conditions are essential to provide insight and guidance to the simulation of prototypic systems undergoing drying processes. This report documents details on the quantification of residual water in the Advanced Drying Cycle Simulator (ADCS), an apparatus built to simulate commercial drying procedures and quantify the amount of residual water remaining in a pressurized water reactor (PWR) fuel assembly after drying. The ADCS was constructed with a prototypic 17×17 PWR fuel skeleton and waterproof heater rods to simulate decay heat. The ADCS is outfitted with thermocouples to measure the thermal response of the ADCS to simulated decay heats and internal helium fill pressures relevant to commercial drying procedures. The ADCS is also instrumented with pressure transducers to measure the pressures and vacuum levels observed during simulated commercial drying. The most unique instrumentation used for quantifying residual water in the ADCS is a Hiden Analytical HPR-30 mass spectrometer (MS), which measures gas compositions of the ADCS internal free volume, based on partial pressures calculated from relative proportions of gas molecules detected by the MS. This report details the methodology used to implement MS measurements in quantifying residual water in the ADCS. This methodology includes the calibration of the HPR-30 MS to a Buck Research Instruments CR-4 chilled mirror hygrometer, which itself is calibrated to a NIST-traceable standard. Data collected by both the MS and the chilled mirror hygrometer from water/helium mixtures ranging from 150 to 500,000 ppmv water in helium were used to generate calibration curves, establishing a source of verification of MS measured water contents. Details regarding water content measurement uncertainties are included in this report, defining the accuracy and verifiability of the HPR-30 MS in measuring residual water content in simulated dry storage canister environments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Computational and Experimental Investigation of Thermal-Mechanical- Chemical Mechanisms of High-burnup Spent Nuclear Fuel (SNF) Processes at Elevated Temperatures and Degradation Behavior in Geologic Repositories

The overarching goal of the combined computational and experimental R&D activities proposed in this project is to enhance understanding of the mechanisms and thermal-mechanical-chemical (TMC) parameters controlling the instant release fraction (IRF) and matrix dissolution of high-burnup (HB; burnup) spent nuclear fuels (SNFs) and the subsequent formation, stability, and phase transformations of SNF alteration products under long-term storage and geological disposal conditions. Uranium dioxide may undergo oxidative corrosion/alteration, and the IRF may be increased for HB SNF, both of which may affect environmental systems associated with SNF long-term storage and disposal. The oxidative matrix dissolution may form various complex uranyl-based phases, including a rich variety of oxides, silicates, carbonates and other secondary minerals in varied geological environments (e.g., studtite, metastudtite, amorphous uranyl peroxide, uranium trioxide, triuranium octoxide, schoepite, dehydrated schoepite, metaschoepite, becquerelite, soddyite, rutherfordine,...). These uranyl phases generally have higher mobility UO 2 +2 species than less soluble U 4+ phases. However, limited information on the thermodynamic properties and formation kinetics of these uranyl-bearing phases is available to predict explicitly paragenesis under the conditions relevant to long-term storage or disposal. The proposed project draws on complementary expertise and research backgrounds from the team members: (i) to apply a combined ab initio modeling (UNLV/UTEP and SNL) and experimental (UNLV) strategy investigating the high-temperature TMC mechanisms of alteration of SNF under α-radiolysis conditions; (ii) to investigate the mechanistic of phase transformations in UNF degradation products under various conditions expected in long-term storage systems (e.g. (UO 2 )O 2 (H 2 O) 4 → (UO 2 )O 2 (H 2 O) 2 → U 2 O 7 → UO 3 → U 3 O 8 ); (iii) to determine high-accuracy TMC parameters for complex uranyl-based phases formed in storage or geological disposal environments (e.g. UO 3 (H 2 O) 2 , Ca[(UO 2 ) 6 O 4 (OH) 8 ] 8 H 2 O, (UO 2 ) 2 (SiO 4 ) 3 2H 2 O,…). The unforeseen COVID-19 pandemic led to the laboratory/campus closure since March 2020, that resulted in a significant delay in reaching milestones in a satisfactory manner, due to (i) the statewide recommendation from stop-working to later limited work in the lab and work-from-home (WFH), (ii) no in-person interactions, and (iii) a hiring freeze at UNLV. Therefore, a no cost extension (10/01/2021- 9/30/2022) was requested to help make up the time we lost during the global pandemic in 2020-2021, leading to paradigm shifts in the focus of the project in the following three main tasks: Task 1 (Computational), Task 2 (Experimental), and Task 3 (Final report, due on 12/29/2022).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Sister Rod Destructive Examinations (FY23): Appendix J: Leaching of High Burnup Used Nuclear Fuel in Deionized Water

The leaching experiment aims to understand the trends in the radiolysis-enhanced dissolution of HBU SNF when exposed to water (e.g., in-reactor or in-pool cladding failures). Specimens from a baseline M5-clad rod and a heat-treated M5 rod were cut from the fractured CIRFT specimens and placed in 100 mL deionized water for a period of 128 days. Both radial and axial sections were cut to provide different surface areas of fuel in contact with the leachate. During the four-month exposure period, aliquot samples of the leachate were analyzed using gamma spectroscopy and inductively coupled plasma - mass spectrometry (ICP-MS). The analysis quantified the amount of fuel leached into the solutions and provided individual isotopic release fractions (of 30+ isotopes) which were compared as a function of time and surface area of the fuel exposed. Consistent with existing literature, the leaching followed a trend in which isotopes of certain elements such as Cs and Mo were among the first species in the matrix to dissolve, and with the highest release rates. This was followed by a gradual matrix dissolution consisting of uranium and other actinides and a slower-than-matrix release from some isotopes, including Ru and Rh. It was also observed that the circumferential samples having less exposed fuel surface area (to begin with) leached more than the axial samples for a majority of the isotopes during the timespan of the study. Previous experiments were conducted by decladding or exposing the pellet-clad surface completely, but the samples used in this study retained the cladding. One possible explanation for the higher leaching rate of the circumferential samples is that the pellet-clad interface, which has a greater density of grain boundaries and defects, may be the most vulnerable area to leaching of fuel in the presence of water.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

FY24 Progress Report: Disposition of Cracks & Features Observed in the Inner Can Closure Weld Region (ICCWR) of the 3013 Package

The long-term integrity of the 3013 containers is of interest for the safe storage of Pu materials. Although the 3013 standard embodies multiple barrier concept, the integrity of the inner container is considered to be crucial since it protects the Safety Class outer container from its contents. The container is designed to withstand high pressures that could result from the complete radiolysis of the maximum water content permissible by the 3013 standard. Shelf-life studies and destructive examination have not found high gas pressures but have found that corrosive gases are generated. Pitting and stress corrosion cracking (SCC) are considered to be critical corrosion modes for the performance of the inner container and have been observed during destructive examinations. Considerable research has been conducted on the possibility of aqueous electrolytes condensing on the inner container that can promote SCC. These studies indicate the uncertainties surrounding the formation of corrosive environments and the corrosion behavior of container materials. In this initial report, a Bayesian network (BN) model is described that can consider the uncertainties and the causal connections between various factors influencing the corrosion modes of the inner container. The BN model is preliminary and provides an initial framework to identify the necessary information. The report also provides initial experimental results on the electrochemical behavior of stainless steels in anticipated condensed environments from gas phase migration of acidic gases. The experimental results are consistent with the corrosion model. Recommendation for further work on the BN model include assembling an expert group to provide input to the BN structure and quantification of the conditional probability matrix, experimental studies to characterize the microstructure of the container, electrochemical studies to identify critical potentials for localized corrosion and SCC, and crack growth rate studies

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

DECOVALEX-2023: Task B Final Report

In all repository concepts for the geological disposal of radioactive waste, an engineered barrier system (EBS) is used to encapsulate the waste canister, or, to act as borehole or gallery seals. These systems are often based on bentonite clays due to their low permeability and high swelling capacity enabling the closure of engineering voids. However, in all repository concepts gases will be generated through the corrosion of metallic materials (under anoxic conditions), the radioactive decay of waste and the radiolysis of water. Thus, understanding the processes and mechanisms controlling the advective movement of gas (as a discrete phase) in clay-based materials is a key aspect when assessing the impact of gas flow in a repository safety case.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Multiscale Nuclear-Electronic Orbital Quantum Dynamics in Complex Environments

Many renewable energy conversion processes rely on the movement of protons as well as electrons through either electrocatalysis or photoexcitation. The simulation of such processes requires a quantum mechanical description of coupled nuclear-electronic dynamics in a solvent or heterogeneous chemical environment. The overall objective of this project is the development of theoretical and computational capabilities for simulating nuclear-electronic quantum dynamics in complex environments and the creation of high-performance, open-source software. This multiscale framework will enable simulations of the real-time dynamics of nonequilibrium excited state proton-coupled electron transfer, quantum decoherence, vibronic energy transfer, and ultrafast radiolysis, as well as their associated time-resolved multidimensional spectroscopies. An important outcome of this project will be a sustainable, reusable, and interoperable open-source software ecosystem. This software will be designed for emerging exascale and future national leadership computers. Another key outcome will be a multiscale quantum dynamics method and software enabling simulations of nonequilibrium nuclear-electronic quantum dynamics in complex environments.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
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↗
DOE OSTI2025

ALARA Review for Venting of Flanged Tritium Waste Containers (FTWCs) at TA-54

This project will vent headspace hydrogen and oxygen from specialized high-pressure storage vessels called flanged tritium waste containers (FTWCs). There are four of these containers located in Los Alamos National Laboratory’s (LANL’s) Technical Area 54 (TA-54), Building 1028. This building is in the southwest corner of LANL’s Material Disposal Area G. The vented headspace gas is expected to contain tritium in the form of water vapor, elemental hydrogen or a combination of both. The venting operation is expected to take place in 2025. There will be a series of operational readiness reviews prior to venting activities commencing. The FTWCs at TA-54 contain tritium-contaminated metal parts and molecular sieve media, which is a pebble-like material used to absorb water vapor from the air. This molecular sieve media inside the FTWCs is contained in metal canisters, along with some loose media material in bags. Over time, tritiated water vapor that had been adsorbed onto the media can become liberated into the FTWC headspace gas. Radiolysis can cause separation of the water vapor into its hydrogen and oxygen components, resulting in the potentially hazardous gas mixture within the FTWC. LANL has determined that continued tritium storage in these containers can pose an unsafe condition due to possible hydrogen and oxygen gas buildup within the FTWCs with a potential for explosion if sparks are generated within the FTWCs. To mitigate this hazard, the FTWCs will be stabilized by venting them in-place to remove hazardous gases.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Plutonium Oxidation State Distribution in the Presence of WIPP-Relevant Organics and Iron Corrosion Products

The oxidation state and solubility of plutonium (Pu) in high ionic strength synthetic WIPP (Waste Isolation Pilot Plant) brines as a function of pC H+ in the presence and absence of WIPP-relevant organic ligands (EDTA [Ethylenediaminetetraacetic acid], oxalate, citrate, acetate) and iron corrosion products (magnetite and metallic iron) at 𝑇 = 23 ± 2 ∘C was thoroughly studied by long-term batch solubility experiments (between approximately 800-1,100 days) from an undersaturation approach. The oxidation state of Pu in the WIPP environment has been a topic of interest since the initial Compliance Certification Application (CCA). This study aims to investigate the solubility of Pu under the expected WIPP conditions and to determine the oxidation state of the solid phase that will control the solubility. One of the most important results of this study is that the Pu oxidation state was analyzed both from the surface area of the corrosion products and in the precipitated solid. The analysis of the Pu oxidation state on the surface of the iron mineral from the ongoing undersaturated experiments is more relevant to the performance assessment of nuclear waste disposal than short term batch (plutonium-iron phase) experiments. The X-ray Absorption Near-Edge Spectroscopy (XANES) analysis showed that Pu oxidation state is different on the metal surface and in the precipitated solid. Pu(III) is the dominant oxidation state in the metallic iron (Fe 0 ) system in the presence and absence of organics. Pu(IV) is the dominant oxidation state in the magnetite system in the presence and absence of organics. Also, organics stabilize Pu(IV) in the magnetite system. Analysis of Pu in the precipitated solid by Extended X-ray Absorption Fine Structure (EXAFS) analysis showed that Pu formed an inner-sphere complex with iron with minor amounts of PuO 2 present. X-ray diffraction (XRD) results indicate that metallic iron and magnetite did not oxidize in three years in the alkaline and high ionic strength system. Under these conditions (8 < pC H+ < 10 at T = (22 ± 2) °C under nitrogen atmosphere, the solubility of Pu changes by up to three orders of magnitude (10 -5 and 10 -8 M). The spread in solubility is highest at pC H+ = 9. Pu(III) and Pu(IV) showed different solubility behavior in the synthetic WIPP brine. A summary of the data collected in this report will be submitted to Sandia National Laboratories as part of a parameter update report which will outline the changes to the OXSTAT parameter for the 2026 Compliance Recertification Application (CRA-2026). The experiments performed were done according to the U.S. Department of Energy (DOE) approved Test Plan entitled “Effects of Radiolysis, Organic Complexation, and Redox Conditions on the Speciation and Oxidation State Distribution of Pu(III/IV)” (LCO-ACP-25). All data reported were obtained under the Los Alamos National Laboratory-Carlsbad Office (LANL-CO) Quality Assurance Program, which is compliant with the DOE Carlsbad Field Office, Quality Assurance Program Document (CBFO/QAPD).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Van de Graaff Irradiation Studies of Oxygen Sensor and Hydrated Zirconia Sorbent

Commercial-scale production of Mo-99 demands a facility-wide assessment of radiation-induced degradation across process equipment, instrumentation, and purification media. Key factors— total ionizing dose, dose rate, radiolysis-driven chemistry, thermal loads, activation, and any neutron-related effects—should inform material selection, shielding, and component layout. Purification materials (sorbents, resins, filters, seals, housing) require particular attention to ensure that irradiation does not alter selectivity or capacity, generate fines, or introduce leachables that could compromise Mo-99 purity or downstream Tc-99m generator performance. Dose mapping, accelerated irradiation testing, and studies of extractables/leachables together provide the basis for setting maintenance, calibration, and replacement intervals and for qualifying materials that retain mechanical integrity, chemical compatibility, and radiopurity over their service life.

36 MATERIALS SCIENCE↗
DOE OSTI2026

Radiolytic Gas Generation and Pressure Buildup in a Closed System Containing Mo-99 Solution

This study measured radiolytic gas generation and pressure buildup in a sealed stainless-steel system containing alkaline Mo-99 solution with added sodium nitrate as a hydrogen suppressant. In each of two experiments, approximately 200 Ci of Mo-99 solution was transferred into a closed experimental vessel inside a hot cell, isolated, and monitored for pressure rise caused by gas generation during radioactive decay. After pressure buildup, headspace gas samples were collected and analyzed by mass spectrometry to determine hydrogen and oxygen concentrations. The purpose was to quantify the magnitude of pressure buildup in a closed system and to characterize the gas composition produced by radiolysis of the Mo-99 target solution under representative handling and storage conditions. The two experiments used similar total Mo-99 activity but differed in solution volume, headspace volume, and leak integrity. In the first experiment, 197.5 Ci of Mo-99 in 16.92 mL of solution was loaded into a vessel with a 31.1-mL headspace; a small leak was later identified, and the measured peak pressure of 46 psig was extrapolated to about 70 psig in the absence of leakage. The headspace gas from this experiment contained about 21.4% H2 and 63.4% O2, but the composition was influenced by preferential hydrogen loss through the leak. In the second experiment, 193 Ci of Mo-99 in 5.41 mL of solution with 0.46 g NaNO3 was loaded into a vessel with a larger 42.61 mL headspace, and no detectable leak was observed. This experiment reached a peak pressure of 38 psig, and the measured gas composition was approximately 28.2% H2 and 51.5% O2. Because the second experiment was leak-free, it is considered the more reliable indicator of the true pressure buildup and intrinsic radiolytic gas composition of the Mo-99 solution.

Chemerisov, Sergey D.↗
DOE OSTI2025

Why ionizing radiation enhances surface wettability

Radiation-Induced Surface Activation is an inherent phenomenon where surfaces that are exposed to gamma irradiation are observed to undergo an increase in wettability. This increase in wettability as a result of the ionizing radiation exposure has so far been demonstrated to have a pronounced impact on Leidenfrost temperature and two-phase fluid dynamics. Test results from previous experiments have shown that incorporation of this effect on heat transfer equipment design may increase the thermal-hydraulic margin leading to higher thermal efficiency. However, the mechanism behind the increased wettability is not clearly understood. In the present work, three different materials (Zircaloy-4, 316 stainless steel, and copper) were exposed at two different dose rates with use of two different gamma irradiation facilities. A detailed surface characterization on the post-irradiated samples is carried out to understand the changes in surface chemistry, wettability and surface morphology. It is observed from the experiments that the increase in wettability upon irradiation depended on the total dose and not on the dose rate. Moreover, localized oxidation and porosity induced by radiolysis was seen to be the predominant mechanism behind increased wettability which leads to improved Leidenfrost temperature.

Seshadria, Arunkumar↗