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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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DOE OSTI2024

Toward an integrated multi-gigahertz ionizing particle diagnostic

Needs arising at both current and future accelerator facilities call for the development of radiation-hardened position-sensing diagnostics that can operate with multi-GHz repetition rates. Such instruments are likely to also have applications in the diagnosis of rapid plasma behavior. Here, building on the recent work of our Advanced Accelerator Diagnostics Collaboration, we are exploring the development of integrated multi-GHz ionizing particle detection systems based on chemical-vapor deposition diamond sensors, with the initial goal of producing a quadrant detector that can determine the intensity and centroid position of a particle beam at a repetition rate between 5 and 10 GHz. Results from our initial high-speed characterization work are presented, including those from a single-channel sensor with a GHz response. Approaches to achieving multi-GHz (5–10 GHz) rate capability, including the design of a dedicated Application Specific Integrated Circuit and the use of 3D RF-solver computer aided design software, are presented and discussed in more detail. 3D RF simulations suggest clean pulses of duration less than 250 ps (FWHM less than 125 ps) can be achieved with the approaches developed by this work.

43 PARTICLE ACCELERATORS↗
DOE OSTI2024

Godiva IV Simulated Radiation Field Characterization and Variance Reduction

Godiva IV is a system comprised of highly enriched uranium alloyed with molybdenum in the form of fuel plate rings. The reactor, along with its predecessors, was designed with the unique ability to satisfy interests in the super-prompt-critical reactor operation space. Originally, the reactor was part of the Los Alamos Critical Experiments Facility (LACEF) at Technical Area-18 (TA-18). The radiation field around Godiva at this facility was well characterized and understood. As a fast neutron system, the neutron spectrum in and around Godiva was close to a Watt Fission spectrum. The Kiva where Godiva IV was located at LACEF was made of thin, sheet metal walls which did not contribute significantly to the neutron spectrum. Following the transition of LACEF to the National Critical Experiments and Research Center (NCERC) in Nevada, Godiva-IV was moved from TA-18 to the Device Assembly Facility (DAF) at the Nevada National Security Site (NNSS). Part of this move brought renewed interest in radiation field characterization. The new facility introduced significant changes to the environment surrounding Godiva, and preliminary foil irradiation results suggested that the room contribution to the neutron spectrum was significant. Unlike at TA-18, a large thermal neutron signature was added to the fast spectrum from Godiva due to significant room return. A primary goal due to the additional complexity that the room return adds to the Godiva IV radiation emission spectrum was the development of an efficient Monte Carlo N-Particle (MCNP) calculation capable of characterizing the neutron spectrum anywhere in the room around Godiva. A campaign of activation foil irradiations and analysis were completed to support the validation of the MCNP model. The modeling of these foils in MCNP can be easily done with a standard volumetric neutron flux tally. However, given the multitude of locations and reaction rates to be modeled, further steps must be taken to increase the efficiency of these calculations in MCNP. During this study, a benchmark model currently under development for Godiva IV was used. A qualitative assessment of the thermal neutron contributors was performed using spatial neutron distribution plots. Additional detail was added to the model based on the qualitative results showing the thermal spectrum’s large sensitivity to hydrogenous material. Neutron energy spectra was evaluated at discrete locations in the room around Godiva to quantify the relative contribution of various components. It was discovered that the concrete walls are the largest contributor to the thermal signature, with minor contributions from plastic components surrounding Godiva. Following these results, two different variance reduction techniques were implemented to improve the problem efficiency in these calculations. In the first approach, an F5 point detector tally was implemented in the standard Godiva IV criticality problem. The second approach involved a weight-window generator implementation with an F5 point detector tally in a fixed source problem. The weight window implementation reduced the runtime from 42739.55 minutes to 1803.34 minutes (computer time), compared to the F5 KCODE implementation.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2021

A Predictive Model for Offgas Composition in Waste Glass Melters - Spring 2021 SULI General Audience Abstract

Approximately 56 million gallons of nuclear waste are currently stored in 177 underground storage tanks at the Hanford site in Washington state. A portion of the tank waste is targeted for immobilization by converting the nuclear waste to borosilicate glass through a process called vitrification. During vitrification, emissions are produced in addition to the intended glass product. These emissions, contained in a volume above the molten glass called the plenum, are composed of water vapor, air components, aerosols, particulates, and other gaseous species that result from evaporation, reactions, and air influx. The formation of these gasses causes the surface of the glass to foam. To reduce foaming, sucrose is added to the waste stream before the conversion to glass. Sucrose also reduces the formation of hazardous gasses in the plenum including nitrous oxides, NOx, and carboxides, COx. However, if excess sucrose is present, the reactions that produce the gasses in the plenum will be incomplete, and increased levels of NOx and COx can be observed. The Waste Treatment and Immobilization Plant at the Hanford site is equipped with the technology to treat the plenum emissions. Though many tests have been run to determine the presence of particulates in the plenum emissions, there is does not exist an established method to confidently predict the composition of the gaseous species in the plenum prior to treatment without physical testing. A model was created in the process modeling software Aspen Plus to predict the composition of NOx and COx in the plenum. The model employs a constant stirred tank reactor and user-defined reaction chemistry to determine the plenum composition. To verify the model, the offgas predicted by the model were compared to available offgas data for ten different feed compositions. It is demonstrated that the developed Aspen Plus model is capable of predicting the composition of NOx and COx in the plenum.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

The TRANSCEND University Consortium: An Overview - 20106

The UK government is committed to nuclear energy having an important role in delivering a secure, low-carbon and affordable energy future. Projections indicate that between now and 2035, ∼14 GW of new nuclear generating capacity may be built. Successful delivery of this strategy recognises the need for research and development, skills development and international collaboration as key enablers. One key area is the decommissioning and clean-up of all civil nuclear licensed sites and the implementation of geological disposal of highly active wastes, with a central component being a joined-up approach to nuclear R and D across government, industry and academia which benefits the UK economy and ensures the security of supply. Essential to public acceptance of this strategy remains a demonstrable ability to safely manage and dispose of waste, with acceptance directly linked to having scientifically robust routes for its safe clean-up and disposal. There therefore remains an ongoing need for further research in the areas of decommissioning, immobilisation and management of nuclear waste, and there is significant government and stakeholder support for such research. Given the needs of both new nuclear power generation and legacy wastes, there is a requirement to develop innovative approaches to waste management and decommissioning. Following on from earlier, highly successful projects supported by UK Research and Innovation's Energy Programme, the work of the TRANSCEND (Transformative Science and Engineering for Nuclear Decommissioning) consortium is part of the response to these needs, providing support to a strong and internationally leading group of researchers in this key technology area. The consortium is a collaboration between 11 universities, supported by 8 key industry partners, with a total funding of UK Pounds 9.5 M from the UK Engineering and Physical Research Council, the universities and key project partners. The research being carried out is multidisciplinary and covers both fundamental and applied topics, with 40 research projects clustered in four technical themes: integrated waste management; site decommissioning and remediation; spent fuels; and nuclear materials. Key aims of the consortium are to extend and further develop a relevant academic skill base in the UK associated with the needs of nuclear waste management and decommissioning, to train the next generation of UK researchers, equipping them with skills and experiences relevant to nuclear waste management and decommissioning issues, and to carry out cutting-edge research that has the potential to provide new and innovative pathways to better management of both legacy and future nuclear wastes. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Water Migration and Swelling in Engineered Barrier Materials for Radioactive Waste Disposal

Deep, underground repositories are needed to isolate radioactive waste from the biosphere. Because bentonite is an integral component of many multibarrier repository systems, information on the hydraulic behavior of bentonite is crucial for modeling the long-term viability of such systems. In this paper the hydraulic behavior of bentonite samples was analyzed as a function of aggregate size, and samples were subjected to hydrothermal treatments involving contact with NaCl, KCl, and deionized water. Neutron and X-ray imaging were used to quantify water sorption into packed bentonite samples and bentonite swelling into the water column. The distance between the original clay-water interface and the wetting front was determined as a function of time. Average water uptake exhibited a square-root-of-time dependence in freshly prepared samples, but more variable rates were observed for samples previously in contact with water. The radiography was supported by small-angle neutron scattering analysis and ultra-small-angle neutron scattering analysis of aggregate size distributions and by inelastic neutron scattering to understand the physicochemical environment of the sorbed water. Results showed that hydrothermal treatment with KCl had the greatest effect of increased water transport in the bentonite, possibly as a result of the interaction of K + with smectite layers in the clay.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Benchmark Specification for FFTF LOFWOS Test #13

The Fast Flux Test Facility (FFTF) at the Hanford site in Washington was designed by the Westinghouse Electric Corporation for the U.S. Department of Energy. FFTF was a 400 MW thermal, oxide-fueled, liquid sodium cooled test reactor, built to assist development and testing of advanced fuels and materials for fast breeder reactors. After reaching criticality in 1980, FFTF operated until 1992, providing the U.S. Department of Energy (DOE) with the means to test fuels, materials, and other components in a fast neutron flux environment. In July 1986, a series of unprotected transients (with the plant protection system intentionally disabled) were performed in FFTF as part of the passive safety demonstration program. Among these were thirteen loss of flow without scram (LOFWOS) tests. The goals of this program included confirming the liquid metal reactor safety margins, providing data for computer code validation, and demonstrating the inherent and passive safety benefits of specific design features. The test defined in this benchmark is LOFWOS Test #13, which was initiated at 50% power and 100% flow with the pump pony motors turned off. This benchmark specification is intended to support collaborative efforts within international partnerships on the validation of simulation tools and models in the area of Sodium-cooled Fast Reactor (SFR) safety. Validated tools and models are needed to evaluate SFR inherent safety characteristics and assess the impact of passive design features in response to accident initiators. Comparisons with experimental data and the results of safety analyses from other groups create unique opportunities to improve predictive capabilities of computational codes and methods for SFR modeling and simulation. The conditions of the LOFWOS test along with the feedback from FFTF’s limited free bow core restraint system and the novel passive reactivity control Gas-Expansion Modules (GEMs) pose a very challenging and uniquely valuable benchmark exercise.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Investigation of Radiological Contamination at Parks in the Gateway National Recreation Area - 20350

In 2005, the National Park Service was notified that elevated radiation levels were detected at Great Kills Park during a baseline radiological mapping effort conducted by the Counter Terrorism Bureau of the New York City Police Department and the United States Department of Energy. Great Kills Park is in the Staten Island Unit of the Gateway National Recreation Area. A records search determined that approximately 280 acres of the park was established over an area in which the New York City Department of Sanitation placed approximately 15.1 million cubic yards of waste fill from 1944 to 1948. Since 2005, a series of response actions and investigations were conducted on behalf of the National Park Service to mitigate risk to human health and the environment and to determine the nature of radiological contamination at Great Kills Park. The findings support that radiological contamination is related to certain components of waste fill, including incinerator residue from both on and offsite incinerators, coal ash from the historical use of coal-fired boilers for heating, and incidental radiological artifacts in the waste fill due to their historical use and subsequent disposal. A majority of the radiological artifacts were determined to contain radium-226 and included radioluminescent markers, radium impregnated switch board buttons, radium paint containers, and brachytherapy devices. The National Park Service has several other Parks in the Gateway National Recreation Area that were established in areas with a similar history to Great Kills Park, and overlie waste fill historically placed during landfilling operations, including Spring Creek Park and Dead Horse Bay. Over the last year, the National Park Service has performed additional radiological investigations at Great Kills Park as part of the ongoing Remedial Investigation under the Comprehensive Environmental Response, Compensation, and Liability Act. In addition, preliminary radiological investigations were performed at Spring Creek Park and Dead Horse Bay Park in the Jamaica Bay Unit of Gateway National Recreation Area. These investigation activities identified gamma radiation levels above the ambient level, several of which were attributable to radiological artifacts similar in nature to those recovered at Great Kills Park. Ongoing investigations at Parks in the Gateway National Recreation Area support an association between radiological contamination and historical landfilling practices that were common prior to the regulation and control of radioactive materials. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Second Target Station High-Fidelity Target Activation Comparison

The development of the Second Target Station (STS) target system at the Spallation Neutron Source (SNS) at Oak Ridge National Laboratory (ORNL) is well underway. The target system at STS consists of a rotating target disk that contains 21 segments of tungsten clad in tantalum clad in steel. A key aspect of the design of the target system is to account for the delayed heating and material damage caused by the delayed dose from decaying radionuclides. These radionuclides are a product of either spallation reactions or transmutation of the nuclei in the target system. These radionuclides build up in the target system components over the lifetime of the facility, and the radiation that is emitted can deposit energy in the components causing significant component heating and material damage. Monte Carlo N-Particle (MCNP) Version 6.2 transports the various particle species and calculates the spallation products and neutron fluxes throughout the target system. These spallation products and neutron fluxes along with the material definition of each component are relayed to the CINDER2008 transmutation code to calculate the radionuclide inventories and the corresponding decay gamma emission spectra. MCNP6.2 coupled with CINDER2008 is the computational method-of-choice for the analysis discussed in the following sections of this report. The analysis focuses on validating major assumptions in calculating the radionuclide inventory in the STS target system: all of the target segments are fresh, unirradiated material when the protons are incident on the segment, the average of the 21 segments of the target is sufficient to represent a single segment, and that averaging the proton pulse structure over time does not significantly affect the radionuclide inventory. The position-averaged, high-fidelity, and single-tally computational methods are used to validate the assumptions and provide a point of comparison to evaluate how the assumptions impact the radionuclide inventories. A more detailed explanation of the three computational methods is provided in Section 2. The position-averaged and single-tally methods are less computationally expensive when compared with the high-fidelity method where 54,000 individual calculations are needed to calculate 1 hr of STS operation. Section 3 details the comparison of the three methods to show that the assumptions made in the position-averaged method do not significantly impact the radionuclide inventory after 1 hr of operation. The discussions and results in this report are for 1 hr of operation. Due to the computational cost associated with calculating the transmutation and activation using the high-fidelity method, only 1 hr of operation has been calculated. The discrepancies observed after 1 hr of operation are not extrapolated out to longer operational times, and this report does not address how the discrepancies between the computational methods may manifest for longer operational periods.

43 PARTICLE ACCELERATORS↗
DOE OSTI2022

Removal of Bond Sodium from Full-Length Unirradiated Fermi-1 Blanket Elements and Assembly via Melt-Drain-Evaporate Process

Equipment was designed, fabricated, tested, and operated in an inert atmosphere radiological glovebox at Idaho National Laboratory to demonstrate the removal of bond sodium from full-length unirradiated Fermi-1 radial blanket elements and an entire radial blanket assembly using a Melt-Drain-Evaporate process. A series of three runs was performed with individual and multiple radial Fermi-1 blanket elements, and a fourth run was conducted with an entire Fermi-1 radial blanket assembly. After each run, the depleted uranium alloy slugs in every element slid out of its cladding, mechanically exhibiting the effectual absence of bond sodium. Quantitative and further qualitative analyses of the treated Fermi-1 material revealed the substantive, if not complete, absence of sodium metal in blanket element components. Indeed, entire columns of depleted uranium alloy slugs and associated cladding from select treated elements were separately contacted with water to react with residual sodium metal on the element surfaces, forming hydrogen gas. Samples of the resultant gas were analyzed to quantify the hydrogen concentration, which correlated to a residual sodium metal content for the treated element. Accordingly, no detectable sodium metal (<7 µg) was found on the surfaces of multiple depleted uranium alloy slugs after removal from its cladding. Detectable sodium metal, ranging from 14 to 30 µg, was found on the surfaces of one separated column of slugs and two separated full-length cladding segments. Each element originally contained ~25 g of bond sodium, resulting in quantitative sodium metal removal efficiencies of =99.9998%. The remaining separated uranium alloy slugs and cladding segments from the four runs were piece-wise immersed in alcohol while videorecording possible gas bubble formation as an additional qualitative analysis for residual sodium metal on treated surfaces, which identified the predominant absence of sodium reactivity on the balance of treated Fermi-1 blanket material. The results of this demonstration substantiate a path forward for treatment and disposal of 34 metric tons heavy metal of irradiated sodium-bonded Fermi-1 blanket material currently stored at Idaho National Laboratory.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Report on Initial Sodium Testing on the Thermal Hydraulic Experimental Test Article (THETA) (Fiscal Year 2024 Final Report)

The Thermal Hydraulic Experimental Test Article (THETA) is a facility that is used to develop sodium components and instrumentation as well as to acquire experimental data for validation of reactor thermal hydraulic and safety analysis codes. The facility simulates nominal thermal hydraulic conditions as well as protected/unprotected loss of flow accidents in a sodium-cooled fast reactor (SFR). High fidelity distributed temperature profiles of the developed flow field may be acquired with Rayleigh backscatter based optical fiber temperature sensors. The facility was designed in partnership with systems code experts to tailor the experiment to ensure the most relevant and highest quality data for code validation. THETA is comprised of a traditional primary coolant and secondary coolant system. The primary system is submerged in the pool of sodium and consists of a pump, electrically heated core, intermediate heat exchanger, and connected piping and thermal barriers (redan). The secondary system, located outside of the sodium pool, consists of a pump, sodium to air heat exchanger, and connected piping and valves. In fiscal year 2023, thermal stratification tests were completed with the primary system online, while the secondary system was being constructed [1]. These tests had shown that the core barrel and intermediate heat exchanger (IHX) outlet required increased thermal insulation. The THETA primary system was removed from METL, cleaned, thermal insulators installed, and then inserted into METL Test Vessel 4. At the time of this writing the THETA primary and secondary system are operational. During this fiscal year 100+ hours of testing was completed to characterize thermal hydraulic phenomena associated with steady state and transient conditions in a pool type liquid metal cooled reactor. A majority of the testing campaign was completed to satisfy the experimental data acquisition requirements for the GAIN Voucher with Oklo, CRADA 2021-21121. THETA is still operational at the time of this publication and future testing is planned for fiscal year 2025. Work is underway to publish existing and future data to an online database to facilitate collaboration with SFR engineers looking to validate their systems code or computational fluid dynamics models.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Strontium Speciation in Relevant Tank Waste Components Examined by Electrospray Ionization Mass Spectrometry

The identification of chemical species formed in complex nuclear waste is crucial for the development and employment of advanced separations technologies to remediate the Hanford site by processing tank waste. The current Tank Side Cesium Removal (TSCR) process deployed at Hanford utilizes crystalline silicotitanate (CST) ion exchange (IX) media to aid in the separation of low-activity waste for proper treatment and disposal. The inorganic IX media is highly selective for Cs but has been shown to also remove Sr from caustic simulants and small-scale IX processing of Hanford tank waste.(Fiskum, Rovira et al. 2019, Fiskum, Campbell et al. 2021, Westesen, Campbell et al. 2022) Quantitative Sr removal has not been observed in all tank waste supernates tested; thus, to better understand Sr removal and effectively predict processing behavior through TSCR, it is necessary to first investigate Sr speciation in tank waste. This work utilized electrospray ionization mass spectrometry (ESI-MS) to identify ionic Sr complexes that form in the presence of NO 3 –, NO 2 –, OH–, and Cl–. Although our results show that NO 3 –, NO 2 –, and OH– are competitive for Sr 2+ binding, previous data from IX studies indicate that [SrOH] + is not the dominant species of concern in tank waste processing schemes.(Fiskum, Campbell and Trang-Le 2020) Our results show that the [Sr(NO3)]+ species and the [Sr(NO2)] + species form in considerable abundances, which may affect the ability to separate Sr using CST in nuclear waste separation processes.

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

Processes in Salt Repositories for Radioactive Waste Disposal

This document summarizes the key processes (thermal, hydrological, mechanical, and chemical; THMC) impacting the features of a deep geological repository for radioactive waste in salt. Some processes are natural and on-going whether the repository is there or not, and other processes are driven by the perturbation associated with the repository. The features considered here include both engineered and natural components of the repository system. The engineered barrier system (EBS) in a salt repository is quite different from those implemented for a repository in clay or crystalline rocks, because it is comprised mostly of granular salt and salt-compatible cements, rather than bentonite. When compared to other rocks (i.e., silicates), salt has unique properties that make it an excellent potential host rock. Openings and fractures in salt creep closed readily. Salt has high thermal conductivity, which can reduce peak temperatures. Additionally, far away from the excavations the porosity of salt is unconnected, which leads to essentially zero advective or diffusive transport. The small amount of hypersaline brine occurring in salt minimizes microbial activity, reduces colloid-assisted transport, and eliminates in-package criticality (i.e., chloride is a neutron poison). At the end of the report, we present a brief outline for a potential salt repository, including considerations avoided in previous repository disposal concepts. We propose considering higher-temperature processes in future disposal concepts, rather than trying to minimize the thermal perturbation of the repository. Since hot salt is drier, a dry repository would limit corrosion, gas generation, and solute transport. Openings and fractures creep shut faster in hot salt. Therefore, higher temperatures could be seen as beneficial, rather than something to minimize, through increased spacing between waste packages (increasing repository costs).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2026

Impedance and wakefield studies of the EIC RCS 591 MHz five-cell cavity

The Electron-Ion Collider (EIC) is a next-generation accelerator complex designed to enable high-luminosity collisions between highly polarized electrons and light ions (e.g., He-3). A central component of its Electron Injection System (EIS) is the Rapid Cycling Synchrotron (RCS), which accelerates a single 28 nC electron bunch from 750 MeV to 5, 10, or 18 GeV using an array of 591 MHz five-cell superconducting RF (SRF) cavities—eight at the current design stage. To ensure stable acceleration of high-charge bunches, we conducted detailed impedance and wakefield studies of the SRF cavity structure using both frequency- and time-domain methods. Wakefield solvers (ECHO3D, ECHO1D, CST), eigenmode analysis, and multi-particle tracking with ELEGANT were employed to evaluate longitudinal and transverse impedance effects and to determine instability thresholds. These studies provide critical input for the cavity design and operating parameters required to preserve beam quality and stability in the RCS.

Accelerator Physics↗
DOE OSTI2024

Dissolution Flowsheet for Non-Aluminum Spent Nuclear Fuel Campaign 1

As part of the Accelerated Basin De-inventory (ABD) program, H Canyon plans to dissolve non-aluminum spent nuclear fuel (NASNF) in the 6.3D electrolytic dissolver. NASNF Campaign 1 plans to electrolytically dissolve 68 bundles of fuel assemblies from the Carolinas-Virginia Tube Reactor (CVTR), Heavy Water Components Test Reactor (HWCTR), and Experimental Boiling Water Reactor (EBWR). The fuel assemblies are intact Zircaloy or stainless steel (SS) clad UO 2 rods, tubes, and plates. The H Canyon electrolytic dissolver previously dissolved a variety of UO 2 core fuel types in SS, Zircaloy, Nichrome, or Incoloy cladding from 1969 to 1980. The objective of this study was to identify flowsheet conditions through literature review and laboratory experimentation to safely dissolve NASNF Campaign 1 bundles in the H Canyon electrolytic dissolver. Bench-scale electrolytic dissolution tests were performed to demonstrate a flowsheet for NASNF Campaign 1 bundles. The outer bundles are composed of SS or Al alloy, Al 6061-T6, and contain intact Zircaloy or SS clad UO 2 fuel assemblies. The key objectives of these tests were to determine bounding dissolver chemistries and the sparge requirement to ensure H 2 concentration remain less than 60 vol % of the lower flammability limit (LFL) during dissolution. The impact of HNO 3 concentration and the addition of fluoride on the dissolution efficiency of Zircaloy, 304L SS, Al 6061-T6, and Inconel 625 were examined. While SS, Al, and Inconel 625 readily dissolve utilizing electrolytic dissolution, Zircaloy disintegrated anodically; the surface of Zircaloy oxidized and the oxide layer spalled off and settled at the bottom of the dissolver as an insoluble material. The black flakes were identified as ZrO 2 and 85% of the Zr processed was converted to black ZrO 2 flakes when Zr was anodically disintegrated in 9.5 M HNO 3 .

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Sensitivity analysis of in-pile critical heat flux experiments in TREAT for characterization of RIA power-transient effects

A reactivity-initiated accident (RIA) is one type of postulated design basis accident (DBA) that can cause a departure from nucleate boiling (DNB) event in pressurized-water reactors (PWRs). A DNB occurrence and its consequences depend on the thermophysical properties of the fuel components and coolant, characteristics of the transient energy insertion into the fuel rod, and the onset of the critical heat flux (CHF) phenomenon. To leverage the restart of the Transient Reactor Test (TREAT) Facility, an effort is currently underway to better understand the cladding-to-coolant heat transfer mechanisms and the CHF phenomenon under fast-transient irradiation conditions. This paper characterizes the impact of power transients on the thermal-hydraulic behavior of a TREAT Facility reactor heater rodlet CHF experiment to provide the priority of parameters that need to be investigated for an improved CHF model. Sobol sensitivity analysis methods and the Reactor Excursion and Leak Analysis Program (RELAP5-3D) code were used to identify key input parameters on the uncertainty in the prediction of peak outer- and inner-surface temperatures of the heater tube, as well as the time of the DNB event. A series of sensitivity analyses revealed the total energy deposition on the tube and the transient effects of power pulse had large impacts on the maximum temperatures. The CHF multiplier had the largest impact on the time occurrence of CHF. The overall results show the energy deposition rate in the tube is the most influencing factor to the manifestation of CHF and the resulting thermal-hydraulic behaviors of the tube. The multiplier for the CHF, which is interpreted as the predicted CHF value, has the largest Sobol indices for the time of the CHF in all cases, since it directly determines the occurrence of CHF. It is inferred that the uncertainties in the thermal-hydraulic behaviors of fuels increase with respect to the key parameters as the power pulse becomes broader, and an accurate estimation of the energy deposition rate is required to reduce the uncertainty in the evaluation of the integrity of fuel if the CHF is expected to occur near the peak power. Therefore, the outputs are expected to provide rigorous interpretation of ongoing in-pile CHF experiments in the TREAT Facility reactor regarding thermal-hydraulic behavior of the fuel system aiming for a new transient in-pile CHF model.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Accelerated Aging and Evaluation of Hose-In-Hose Transfer Lines in the Hanford Waste Transfer System - 20312

Nonmetallic materials are used in the United States Department of Energy's Hanford Site Tank Farm waste transfer system. These materials include the inner primary hoses in the hose-in-hose transfer lines (HIHTLs), Garlock{sup R} gaskets, ethylene propylene diene monomer (EPDM) O-rings, and other nonmetallic materials. These nonmetallic materials are exposed to β and γ radiation, caustic solutions as well as high temperature and pressure stressors. How the nonmetallic components react to each of these stressors individually has been well established. However, simultaneous exposure of these stressors has not been evaluated and is of great concern to Hanford Site engineers. Florida International University's Applied Research Center engineers worked closely with key Hanford's Washington River Protection Solutions personnel to develop an experimental test plan that determines how these nonmetallic components react to various simultaneous stressor exposures. In this paper, research on the experimental testing of the hose-in-hose transfer lines used in the Hanford tank farm waste transfer system under simultaneous stressor exposures is presented. The stressor exposure experiments consisted of various combinations of simultaneous stressor exposure of caustic solution, high temperature and high pressure stressors. Research efforts focused on evaluating the aging behavior of EPDM by exposing samples of HIHTLs as well as EPDM dog bone shaped specimens to a 25% NaOH solution at (38 deg. C), operating (54 deg. C) and design temperatures (77 deg. C) for 6 months and 12 months. In addition, HIHTL and the EPDM dog-bone specimens were exposed to only hot water at 77 deg. C for a duration of one year. The mechanical and material properties of the samples were characterized and compared with those of the unexposed samples (baseline). Evaluations included burst pressure tests of the EPDM hose-in-hose transfer lines and material tensile strength test of EPDM dog-bone coupons. Both the tensile strength of the EPDM material dog bones and the burst pressure of the HIHTLs significantly decreased with the increasing temperature and increasing exposure time. When compared with the baseline samples, the burst pressure decreased by 28.75% for the sample aged at 77 deg. C after 12 months, in comparison to sample aged at 38 deg. C and 54 deg. C which had about 1.34% and 5.70% loss in burst pressures, respectively, after 12-month exposure. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Declaration of the Chemical Composition of Waste-products for the German Repository Konrad Using Substance Vectors - 20500

The waste acceptance criteria of the German repository Konrad require beside a radiological declaration a compositional declaration to be able to avoid negative impact on the groundwater by a presumed leakage into the groundwater after complete corrosion of the containers in the future. The declaration of the composition has to specify compounds and substances not element analyses, because the chemical behavior and toxicity especially of organic compounds varies strongly The relevant compounds have to be described and quantified using defined substance codes provided by the operator. For simple waste types like pure concrete or steel components, the definition of the composition is rather simple. The composition of more complex waste streams is generally declared using producer specific material vectors, which can be derived for defined waste types (e.g. grouted fluids, ash from combustion, super-compacted mixed waste etc.). Additional substance vectors exist for containers, drums and grout, which together with the waste composition yield the total composition of the waste packages. The substance vectors for homogeneous waste streams like cementitious demolition waste, solidified fluids and ash are mostly based on analyses provided by the waste owner. In some cases, analyses from non-radioactive analogues or data sheets can also be used to demonstrate the composition. For each qualified treatment of radioactive waste streams the material composition is described and balanced on the basis of specific material vectors, which are composed of material codes from the material and cask list provided by the Federal company for radioactive waste disposal. For mixed waste streams like super-compacted or grouted mixed waste or materials simply packed in containers an analyses cannot be used. In these cases, the compositional information from the bookkeeping can be used for the entire waste stream to calculate a mean composition including a compositional range (min / max composition). Only waste packages with an approved material description can be delivered to Konrad. Therefore, the material declaration of all waste streams and the resulting waste products is crucial for all waste owners in Germany to be able to have enough waste packages ready, when the repository is going to open in 2027. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Implications of End Points on Remediation - 20476

The Port Hope Area Initiative (PHAI) is a federal environmental clean-up program. Its mandate is the remediation and local, long-term, safe management of approximately 1.7 million cubic metres of historic low-level radioactive waste in the adjacent communities of Port Hope and Port Granby in Southern Ontario, Canada. The PHAI is being undertaken by Canadian Nuclear Laboratories (CNL), on behalf of Atomic Energy of Canada Limited, a federal Crown corporation. It is currently one of the largest environmental remediation projects in Canada. The safety of workers, the public and the environment while delivering the PHAI compliantly, on time and within budget is CNL's number one priority. The PHAI is being implemented as two projects - the Port Hope Project and the Port Granby Project. The Port Hope Project is the larger of the two and includes the cleanup of approximately 1.2 million cubic metres of historic waste. The Port Hope Project includes a unique and important component - the remediation of approximately 1,200 private residences and businesses to include both interior and exterior verification of soil and surface-contaminated objects. Planning and consultation for the PHAI clean-up criteria and remediation verification standard operating procedures (RVSOP) were initiated in the early 2000's. The criteria were developed through a cooperative effort involving the PHAI, scientific specialists, federal and provincial government agencies, peer reviewers, the municipalities and members of the public. Subsequent verification procedures which detail the unique systematic methods to verify compliance with the clean-up criteria were developed by CNL in consultation with federal regulators and relevant stakeholders. Now, specifically with the Port Hope Small-Scale Sites project having been underway for a year and approximately 38 properties in some stage of remediation, it has become increasingly evident that the practical application of the clean-up criteria and RVSOP methodology is posing unique challenges that are having direct impact on project execution. The current RVSOP approach is based on discrete sample verification not allowing failure of any one sample. When implemented in the field, particularly at the Small-Scale Sites, it leads to a much more extensive excavation than what was accounted for in the original design, the implications of which are substantial delays in project progress and much more significant impact on the individual properties than originally anticipated. This paper will examine the evolution of PHAI clean-up criteria, subsequent RVSOP methodology and its impact on current project execution and the justification for exploring potential revisions to methods of remediation verification. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗