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All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Retrieval of High Level Waste for the Cigeo Project - A Practical Retrieval Test Story - 16085

Since the passing of the June 28, 2006 Act, Andra has incorporated in its step-wise design development of the French high level waste (HLW) and intermediate level long lived waste (IL-LLW) deep geological repository (DGR, aka Cigeo) various requirements linked to 'reversibility' at different stages of the repository life cycle (from its construction phase to its final closure). Reversibility is based on a number of practical dispositions including the progressivity of the development and deployment of the facility, the flexibility of the nature of operations being carried out, the adaptability of the facility to future modifications and the possible retrieval of waste. The capacity to retrieve waste from Cigeo, also called 'retrievability', bears specific requirements. On a shorter term, one must demonstrate that the waste container retrieval operations are technically safe and feasible over at least 100 years. The current paper starts with two acceptations: retrievability and reversibility. It details how those two notions are defined and inter-linked. The 'Retrievability Scale' as proposed by OECD/NEA (and adopted by Andra) is also reminded. The needs for demonstration of retrieval operations are also exposed for the two categories of waste containers disposed of in Cigeo (IL-LLW and HLW) within the context of their respective emplacement configurations (in disposal horizontal vaults and micro-tunnels) at different stages of the repository life cycle. The technical part of the paper is then focused on a practical case story: a technological test campaign (still running) of HLW container retrieval out of a steel cased disposal micro-tunnel. On a test bench built on surface to simulate (create) the environmental conditions that may prevail inside a HLW micro-tunnel at time of retrieval, it was possible to check whether the mechanical means developed for the purpose were adequate and sufficient. The test case story is exposed, c/w the problems and challenges encountered (high temperature impairing the retrieval robot's proper functioning, presence of corrosion products...). The trouble-shooting solutions implemented and the results are commented. This article concludes with a critical analysis of the methods and equipment used, the results obtained and provides some perspectives of improvement. The positive outcomes of this technical test campaign pave the way for planning the in situ retrieval tests to be programmed in the 'industrial pilot phase' of Cigeo (to be implemented over the ten first years of operations, including the active and inactive test period and the first years of disposal activities): the industrial pilot phase is presently scheduled around 2026-2036 (provided the Cigeo construction is authorized in 2021, following the licensing request filing scheduled in 2018). The issues of IL-LLW containers retrieval (for which specific solutions are also studied and technical demonstrations engaged or planned) are also briefly presented. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Efficient Discovery of Novel Multicomponent Mixtures for Hydrogen Storage: A Combined Computational/Experimental Approach

The objective of the proposed program is to discover novel mixed hydrides for hydrogen storage, which enable the DOE 2010 system-level goals. Our goal is to find a material that desorbs 8.5 wt.% H 2 or more at temperatures below 85°C. The research program will combine first-principles calculations of reaction thermodynamics and kinetics with material and catalyst synthesis, testing, and characterization. We will combine materials from distinct categories (e.g., chemical and complex hydrides) to form novel multicomponent reactions. Systems to be studied include mixtures of complex hydrides and chemical hydrides [e.g. LiNH 2+ NH 3 BH 3 ] and nitrogen-hydrogen based borohydrides [e.g. Al(BH 4 ) 3 (NH 3 ) 3 ]. The 2010 and 2015 FreedomCAR/DOE targets for hydrogen storage systems are very challenging, and cannot be met with existing materials. The vast majority of the work to date has delineated materials into various classes, e.g., complex and metal hydrides, chemical hydrides, and sorbents. However, very recent studies indicate that mixtures of storage materials, particularly mixtures between various classes, hold promise to achieve technological attributes that materials within an individual class cannot reach. Our project involves a systematic, rational approach to designing novel multicomponent mixtures of materials with fast hydrogenation/dehydrogenation kinetics and favorable thermodynamics using a combination of state-of-the-art scientific computing and experimentation. We will use the accurate predictive power of first-principles modeling to understand the thermodynamic and microscopic kinetic processes involved in hydrogen release and uptake and to design new material/catalyst systems with improved properties. Detailed characterization and atomic-scale catalysis experiments will elucidate the effect of dopants and nanoscale catalysts in achieving fast kinetics and reversibility. And, state-of-the-art storage experiments will give key storage attributes of the investigated reactions, validate computational predictions, and help guide and improve computational methods. In sum, our approach involves a powerful blend of: 1) H2 Storage measurements and characterization, 2) State-of-the-art computational modeling, 3) Detailed catalysis experiments, 4) In-depth automotive perspective.

08 HYDROGEN↗
DOE OSTI2015

Improving actinides recycling, a key step towards a more sustainable nuclear energy - 15617

Nuclear energy is thought to be one of the energy sources that could help mitigating the global climate change together with the renewables, due to its low green-house-gases emissions, its reliability and its high load power. However, nuclear energy will only develop if future nuclear energy systems meet the criteria of sustainability in terms of durability, bearability and liveability. Nowadays, many countries chose the so-called once-through cycle which basically considers spent nuclear fuel as a waste, whereas others reprocess their spent fuel to recover the energetically-valuable material Pu (and partially U) to produce Mixed Oxide Fuel (MOX) to be irradiated in a second cycle (twice-through cycle). Both options can be considered relevant, fitting diverse criteria; but if we consider a long lasting prospective for nuclear options, the once-through cycle sustainability is clearly lower (low consumption of natural resource, no recycling of potentially valuable energetic material, larger volume of waste...). Reversely, twice-through cycle can be seen as a first step towards an improved sustainability with a partial recycling of U/Pu and the reduction of the waste volume and toxicity. However, it is likely not sufficient for the long-term and implementing more efficient recycling is a key-driver to improve the environmental friendship, the economics and the social acceptability of nuclear energy. Future nuclear fuel cycles will have to better preserve the natural resource by improving the actinides recycling, to minimize the volume and toxicity of nuclear waste, to ensure their long-term confinement and to prevent any proliferation-risk. Such a gradual and progressive process calls for implementing appropriate technologies, either for reactors but also for fuel treatment/recycling facilities. This paper will describe the successive step that would allow improving the overall sustainability. Indeed, recycled Pu can be once recycled in MOX fuel feeding similar PWRs (Pressurized Water Reactor). Together with the recycling of reprocessed uranium, this first step already allows preserving ∼17% of the natural U resource and to confine ultimate wastes without Pu in a tailored durable wasteform. Subsequent Pu multi-recycling requires introducing Fast Reactors (FR) to efficiently use U-238 available in natural and depleted uranium stockpiles. In parallel, recycling processes would have to be adapted for treating spent LWR and SFR MOX spent fuels, which is currently under study thanks to an ambitious R and D program dedicated on MOX treatment and recycling. For the longer term, minor actinides recycling would allow decreasing the waste burden towards future generations: (i) Am-recycling leads to decrease the waste residual heat-power and increase the repository density and lifespan, (ii) Cm-recycling leads to decrease waste long-term toxicity. The stepwise implementation of these different processes defines a path towards sustainability for nuclear energy that are described in this paper from the fuel cycle viewpoint. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Microstructural development and mechanical properties of a near-eutectic directionally solidified Sn–Bi solder alloy

Sn–Bi solders may be applied for electronic applications where low-temperature soldering is required, i.e., sensitive components, step soldering and soldering LEDs. In spite of their potential to cover such applications, the mechanical response of soldered joints of Sn–Bi alloys in some cases does not meet the strength requirements due to inappropriate resulting microstructures. Hence, careful examination and control of as-soldered microstructures become necessary with a view to pre-programming reliable final properties. The present study aims to investigate the effects of solidification thermal parameters (growth rate — V{sub L} and cooling rate — T-dot{sub L}) on the microstructure of the Sn–52 wt.%Bi solder solidified under unsteady-state conditions. Samples were obtained by upward directional solidification (DS), followed by characterization through metallography and scanning electron microscopy (SEM). The microstructures are shown to be formed by Sn-rich dendrites decorated with Bi precipitates surrounded by a complex regular eutectic mixture, with alternated Bi-rich and Sn-rich phases. Experimental correlations of primary (λ{sub 1}), secondary (λ{sub 2}), tertiary (λ{sub 3}) dendritic and eutectic spacings (λ{sub coarse} and λ{sub fine}) with cooling rate and growth rate are established. Two ranges of lamellar eutectic sizes were determined, described by two experimental equations λ = 1.1 V{sub L}{sup −1/2} and λ = 0.67 V{sub L}{sup −1/2}. The onset of tertiary branches within the dendritic array along the Sn–52 wt.%Bi alloy DS casting is shown to occur for cooling rates lower than 1.5 °C/s. - Highlights: • The Sn–52 wt.%Bi solder was shown to have two eutectic sizes. • The fishbone eutectic is preferably located adjacent to the Bi-rich lamellar phases. • The onset of tertiary dendritic branches in Sn–Bi is associated with T-dot{sub L} < 1.5 °C/s. • Higher eutectic fraction and λ{sub 3} provoked a reverse increase in σ{sub u} and σ{sub y}.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗
DOE OSTI2019

Effect of Minerals on the Removal of Uranium (VI) in the Presence of Humic Acid and Colloidal Silica - 19406

Uranium is one of the key contaminants of concern in groundwater as a result of past nuclear processing activities at the Department of Energy (DOE) facilities such as the Savannah River Site (SRS) in Aiken, South Carolina. SRS was one of the major nuclear processing facilities during the Cold War where plutonium was produced. Beginning with the implementation of the environmental cleanup program in 1981, SRS has become a hazardous waste management facility responsible for the storage and remediation of contaminated soil and groundwater from radionuclides. Approximately 1.8 billion gallons of acidic waste containing radionuclides and dissolved heavy metals was disposed in the F/H Area seepage basins. This led to the unintentional creation of highly contaminated groundwater plumes, consisting of radionuclides with an acidic pH between 3 and 5.5. The acidity of the plumes contributed to the mobility of several constituents of concern (COC) such as tritium ({sup 3}H), uranium-238 ({sup 238}U), iodine-129 ({sup 129}I), and strontium-90 ({sup 90}Sr) for the F-Area plume and {sup 3}H, {sup 90}Sr and mercury for the H-Area plume. This investigation will focus on uranium in the oxidation state +6 (U(VI)), which is a key contaminant of concern in the groundwater plume of the FArea. Initially, the pump-treat-reinject system was implemented from 1997 to 2003 to remove contaminants from polluted groundwater. Downgrade groundwater within the system would be pumped to the water treatment facility and re-injected upgrade within the aquifer. The effectiveness and sustainability of this process diminished over time. Therefore, it was discontinued and replaced with a funnel-and-gate process in 2004. This new process injects sodium hydroxide directly into the gates of the F-Area groundwater to effectively raise pH levels. By raising the pH of the groundwater, a treatment zone is created by reversing the acidic nature of the contaminated sediments and producing a negative net charge on the surface of sediment particles. As a result, the adsorption of cationic contaminants is expected to be enhanced. So far, this process has resulted in a decrease of {sup 90}Sr and {sup 238}U concentrations, though the concentration of iodine has been unaffected by this treatment. The solution used for the injections contains a high carbonate alkalinity in order to overcome the surface acidic conditions and natural partitions in the groundwater system. To maintain the neutral pH in the treatment zone, systematic injections are required. The continuous use of high concentrations of a carbonate solution to raise pH could re-mobilize uranium previously adsorbed within the treatment zone, though this has not been observed in the monitoring data. Humic substances (HS) are major components of soil organic matter. HS are polyfunctional organic macromolecules formed by the chemo-microbiological decomposition of biomass or dead organic matter. These substances are usually divided into three main fractions: humin (insoluble at all pH values), humic acid (soluble at pH greater than 3.5), and fulvic acid (soluble at all pH values). Humic acid carries a large number of functional groups such as aromatic rings, carboxyl groups, phenols, aliphatic chains, etc. Humic acid is an important ion exchange and metal complexing ligand with a high complexation capacity. The ability of humic acid to chemically bind to metals influences in their migration behavior and fate in natural systems. Various studies have suggested that the retention of U(VI) via sorption in the presence of humic acid is a complex process. For instance, humic acid can form an organic coating at the surface of minerals, which can enhance the sequestration of metals. Ivanov et al. (2012) studied the sorption of U(VI) sorption onto bentonite in the presence and absence of humic acid. The study found that uranium sorption in the presence of humic acid was enhanced at low pH (below 3.8) while at moderate pH (3.8 and 6.5) uranium sorption was reduced compared to the absence of humic acid. At high pH (7 and 9), uranium sorption was reduced for both the presence and absence of humic acid. Also, Krepelova et al. (2007) not only found that U(VI) sorption onto kaolinite was influenced by pH, U(VI) concentration, the presence of inorganic carbon species, and humic acid but also that U(VI) preferred to be adsorbed onto kaolinite as a uranyl-humate complex. This investigation analyzed the mechanisms of the synergetic interactions of humic acid and colloidal silica on the removal of uranium in the presence of pure minerals such as quartz, kaolinite and goethite which are characteristic of the F/H area. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Electron Temperature Fluctuation Measurements and Transport Model Validation at Alcator C-Mod

The tokamak is a type of toroidal device used to confine a fusion plasma using large magnetic fields. Tokamaks and stellarators the leading devices for confining plasmas for fusion, and the capability to predict performance in these magnetically confined plasmas is essential for developing a sustainable fusion energy source. The magnetic configuration of tokamaks and stellarators does not exist in Nature, yet, the fundamental processes governing transport in fusion plasmas are universal – turbulence and instabilities, driven by inhomogeneity and asymmetry in the plasma, conspire to transport heat and particles across magnetic field lines and can play critical roles in impurity confinement and generation of intrinsic rotation. Turbulence exists in all plasmas, and in neutral fluids as well. The study of turbulence is essential to developing a fundamental understanding of the nature of the fourth state of matter, plasmas. Experimental studies of turbulence in tokamaks date back to early scattering observations from the late 1970s. Since that time, great advances in turbulence diagnostics have been made, all of which have significantly enhanced our knowledge and understanding of turbulence in tokamaks. Through comparisons with advanced gyrokinetic theory and turbulent-transport models a great deal of evidence exists to implicate turbulent-driven transport as an important mechanism determining transport in all channels: heat, particle and momentum However, prediction and control of turbulent-driven transport remains elusive. Key to development of predictive transport models for magnetically confined fusion plasmas is validation of the nonlinear gyrokinetic transport model, which describes transport due to turbulence. Validation of gyrokinetic codes must include detailed and quantitative comparisons with measured turbulence characteristics, in addition to comparisons with inferred transport levels and equilibrium profiles. For this reason, advanced plasma diagnostics for studying core turbulence are needed in order to assess the accuracy of gyrokinetic models for turbulent-driven particle, heat and momentum transport. New core turbulence diagnostics at the world-class tokamaks Alcator C-Mod at MIT and ASDEX Upgrade at the Max Planck Institute for Plasma Physics have been designed, developed, and operated over the course of this project. These new instruments are capable of measuring electron temperature fluctuations and the phase angle between density and temperature fluctuations locally and quantitatively. These new data sets from Alcator C-Mod and ASDEX Upgrade are being used to fill key gaps in our understanding of turbulent transport in tokamaks. In particular, this project has results in new results on the topics of the Transport Shortfall, the role of ETG turbulence in tokamak plasmas, profile stiffness, the LOC/SOC transition, and intrinsic rotation reversals. These data are used in a rigorous process of “Transport model validation”, and this group is a world-leader on using turbulence models to design new hardware and new experiments at tokamaks. A correlation electron cyclotron emission (CECE) diagnostic is an instrument used to measure micro-scale fluctuations (mm-scale, compared to the machine size of meters) of electron temperature in magnetically confined fusion plasmas, such as those in tokamaks and stellarators. These micro-scale fluctuations are associated with drift-wave type turbulence, which leads to enhanced cooling and mixing of particles in fusion plasmas and limits achieving the required temperatures and densities for self-sustained fusion reactions. A CECE system can also be coupled with a reflectometer system that measured micro-scale density fluctuations, and from these simultaneous measurements, one can extract the phase between the density (n) and temperature (T) fluctuations, creating an nT phase diagnostic. Measurements of the fluctuations and the phase angle between them are extremely useful for testing and validating predictive models for the transport of heat and particles in fusion plasmas due to turbulence. Once validated, the models are used to predict performance in ITER and other burning plasmas, such as the MIT ARC design. Most recently, data from the newly developed, so-called “CECE diagnostic” [Cima 1995, White 2008] and “nT phase angle measurements” [Haese 1999, White 2010] ]will be combined with data from density fluctuation diagnostics at ASDEX Upgrade to support a long-term program of physics research in turbulence and transport that will allow for more stringent testing and validation of gyrokinetic turbulent-transport codes. This work directly impacts the development of predictive transport models in the U.S. FES program, such as TGLF, developed by General Atomics, which are used to predict performance in ITER and other burning plasma devices as part of advancing the development of fusion energy sciences.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2018

Dynamics of Organophosphate‐Induced Structural Changes in Acetylcholinesterase Revealed by Time‐Resolved Small‐Angle X‐Ray Scattering and Inelastic Neutron Scattering

Organophosphates (OPs), the active compounds in nerve agents and insecticides such as paraoxon (POX), exert their toxic effects by irreversibly inhibiting acetylcholinesterase (AChE). Currently available antidotes for OP toxicity can reactivate AChE, but reactivation rates are slow which can lead to irreversible nervous system damage or death. We hypothesize that better antidotes with accelerated reactivation rates can be developed by understanding the structural dynamics of OP‐AChE conjugates. Previous steady‐state small‐angle X‐ray scattering (SAXS) experiments showed that POX and certain other OPs induce quaternary structural changes in human AChE (hAChE) that were reversible by oxime antidotes. To characterize the dynamics of these OP‐induced structural changes we have used time‐resolved SAXS (TR‐SAXS) and inelastic neutron scattering (INS). Stopped‐flow TR‐SAXS experiments were performed at BL‐4 at the Stanford Synchrotron Radiation Lightsource by mixing a molar excess of POX (500 μM) with hAChE (50 μM). Scattering data were collected every 20 msec during the first 1000 msec after mixing. Changes in the solution structure of hAChE during this period occurred with k obs = 0.13(±0.01) s −1 . This rate of structural change is comparable to the rate of hAChE inactivation observed using biochemical assays. INS data were collected at the Cold Neutron Chopper Spectrometer at the Spallation Neutron Source at Oakridge National Laboratory using 100 mg samples of hAChE and POX‐hAChE that were lyophilized and rehydrated in D 2 O. INS data collected at 120 K (to prevent translational motions) were used to compare the vibrational density states of native and POX‐conjugated hAChE. The POX‐conjugated hAChE exhibited an increased number of vibrational states, particularly those associated with regional motions such as loop dynamics, whereas global collective motions remained unchanged. Molecular dynamics simulations were used to simulate the vibrational spectra and identify the region(s) in hAChE that contributed to the differences in the INS experimental data. In aggregate, these studies provide experimental evidence for OP conjugate‐induced changes in hAChE dynamics that has to be considered in the design of antidotes for OP poisoning that have improved efficacy and patient outcomes. Support or Funding Information This work was supported by the CounterACT Program, National Institutes of Health Office of the Director, Grant 1U01NS083451 from NINDS, and by grants to SSRL and ORNL from the U.S. Department of Energy, and the National Institutes of Health. This abstract is from the Experimental Biology 2018 Meeting. There is no full text article associated with this abstract published in The FASEB Journal .

Blumenthal, Donald K.↗
DOE OSTI2025

Separate, separated, and together: the transcriptional program of the Clostridium acetobutylicum-Clostridium ljungdahlii syntrophy leading to interspecies cell fusion

ABSTRACT Syntrophic cocultures (hitherto assumed to be commensalistic) of Clostridium acetobutylicum and Clostridium ljungdahlii , whereby CO 2 and H 2 produced by the former feed the latter, result in interspecies cell fusion involving large-scale exchange of protein, RNA, and DNA between the two organisms. Although mammalian cell fusion is mechanistically dissected, the mechanism for such microbial-cell fusions is unknown. To start exploring this mechanism, we used RNA sequencing to identify genes differentially expressed in this coculture using two types of comparisons. One type compared coculture to the two monocultures, capturing the combined impact of interactions through soluble signals in the medium and through direct cell-to-cell interactions. The second type compared membrane-separated versus -unseparated cocultures, isolating the impact of interspecies physical contact. While we could not firmly identify specific genes that might drive cell fusion, consistent with our hypothesized model for this interspecies microbial cell fusion, we observed differential regulation of genes involved in C. ljungdahlii’s autotrophic Wood-Ljungdahl pathway metabolism and genes of the motility machinery. Unexpectedly, we also identified differential regulation of biosynthetic genes of several amino acids, and notably of arginine and histidine. We verified that they are produced by C. acetobutylicum and are metabolized by C. ljungdahlii to its growth advantage. These and other findings, and notably upregulation of C. acetobutylicum ribosomal-protein genes, paint a more complex syntrophic picture and suggest a mutualistic relationship, whereby beyond CO 2 and H 2 , C. acetobutylicum feeds C. ljungdahlii with growth-boosting amino acids, while benefiting from the H 2 utilization by C. ljungdahlii . IMPORTANCE The construction and study of synthetic microbial cocultures is a growing research area due to the untapped potential of defined multi-species industrial bioprocesses and the utility of defined cocultures for generating insight into complex, undefined, natural microbial consortia. Our previous work showed that coculturing C. acetobutylicum and C. ljungdahlii leads to a unique metabolic phenotype (production of isopropanol) and heterologous cell fusion events. Here, we used RNAseq to explore genes involved in and impacted by these fusions. First, we compared gene expression in coculture to each monoculture. Second, we utilized a transwell system to compare gene expression in mixed cocultures to cocultures with both species physically separated by a permeable membrane, isolating the impact of interspecies “touching” on the transcriptome. This study deepens our mechanistic understanding of the C. acetobutylicum-C. ljungdahlii coculture phenotype, laying the groundwork for reverse genetic studies of heterologous cell fusion in Clostridium cocultures.

Willis, Noah B. (ORCID:0009000689365955)↗
DOE OSTI2014

Laboratory Evaporation Testing Of Hanford Waste Treatment Plant Low Activity Waste Off-Gas Condensate Simulant

The Hanford Waste Treatment and Immobilization Plant (WTP) Low Activity Waste (LAW) vitrification facility will generate an aqueous condensate recycle stream, LAW Off-Gas Condensate, from the off-gas system. The baseline plan for disposition of this stream is to send it to the WTP Pretreatment Facility, where it will be blended with LAW, concentrated by evaporation and recycled to the LAW vitrification facility again. Alternate disposition of this stream would eliminate recycling of problematic components, and would enable de-coupled operation of the LAW melter and the Pretreatment Facilities. Eliminating this stream from recycling within WTP would also decrease the LAW vitrification mission duration and quantity of canistered glass waste forms. This LAW Off-Gas Condensate stream contains components that are volatile at melter temperatures and are problematic for the glass waste form. Because this stream recycles within WTP, these components accumulate in the Condensate stream, exacerbating their impact on the number of LAW glass containers that must be produced. Approximately 32% of the sodium in Supplemental LAW comes from glass formers used to make the extra glass to dilute the halides to be within acceptable concentration ranges in the LAW glass. Diverting the stream reduces the halides in the recycled Condensate and is a key outcome of this work. Additionally, under possible scenarios where the LAW vitrification facility commences operation prior to the WTP Pretreatment facility, identifying a disposition path becomes vitally important. This task examines the impact of potential future disposition of this stream in the Hanford tank farms, and investigates auxiliary evaporation to enable another disposition path. Unless an auxiliary evaporator is used, returning the stream to the tank farms would require evaporation in the 242-A evaporator. This stream is expected to be unusual because it will be very high in corrosive species that are volatile in the melter (chloride, fluoride, sulfur), will have high ammonia, and will contain carryover particulates of glass-former chemicals. These species have potential to cause corrosion of tanks and equipment, precipitation of solids, release of ammonia gas vapors, and scale in the tank farm evaporator. Routing this stream to the tank farms does not permanently divert it from recycling into the WTP, only temporarily stores it prior to reprocessing. Testing is normally performed to demonstrate acceptable conditions and limits for these compounds in wastes sent to the tank farms. The primary parameter of this phase of the test program was measuring the formation of solids during evaporation in order to assess the compatibility of the stream with the evaporator and transfer and storage equipment. The origin of this LAW Off-Gas Condensate stream will be the liquids from the Submerged Bed Scrubber (SBS) and the Wet Electrostatic Precipitator (WESP) from the LAW facility melter offgas system. The stream is expected to be a dilute salt solution with near neutral pH, and will likely contain some insoluble solids from melter carryover. The soluble components are expected to be mostly sodium and ammonium salts of nitrate, chloride, and fluoride. This stream has not been generated yet, and, thus, the composition will not be available until the WTP begins operation, but a simulant has been produced based on models, calculations, and comparison with pilot-scale tests. This report discusses results of evaporation testing of the simulant. Two conditions were tested, one with the simulant at near neutral pH, and a second at alkaline pH. The neutral pH test is comparable to the conditions in the Hanford Effluent Treatment Facility (ETF) evaporator, although that evaporator operates at near atmospheric pressure and tests were done under vacuum. For the alkaline test, the target pH was based on the tank farm corrosion control program requirements, and the test protocol and equipment was comparable to that used for routine evaluation of feed compatibility studies for the 242-A evaporator. One of the radionuclides that is volatile in the melter and expected to be in high concentration in this LAW Off-Gas Condensate stream is Technetium-99 (99Tc). Technetium will not be removed from the aqueous waste in the Hanford WTP, and will primarily end up immobilized in the LAW glass by repeated recycle of the off-gas condensate into the LAW melter. Other radionuclides that are also expected to be in appreciable concentrations in the LAW Off-Gas Condensate are 129I, 90Sr, 137Cs, and 241Am. The concentrations of these radionuclides in this stream will be much lower than in the LAW, but they will still be higher than limits for some of the other disposition pathways currently available. At this time, these scoping tests did not evaluate the partitioning of the radionuclides to the evaporator condensate, since ample data are available separately from other experience in the DOE complex. Results from the evaporation testing show that the neutral SBS simulant first forms turbidity at ~7.5X concentration, while the alkaline-adjusted simulant became turbid at ~3X concentration. The major solid in both cases was Kogarkoite, Na3FSO4. Sodium and lithium fluorides were also detected. Minimal solids were formed in the evaporator bottoms until a substantial fraction of liquid was removed, indicating that evaporation could minimize storage volume issues. Achievable concentration factors without significant insoluble solids were 17X at alkaline pH, and 23X at neutral pH. In both runs, significant ammonia carried over and was captured in the condenser with the water condensate. Results also indicate that with low insoluble solids formation in the initial testing at neutral pH, the use of Reverse Osmosis is a potential alternate method for concentrating the solution, although an evaluation is needed to identify equipment that can tolerate insoluble solids. Most of the ammonia remains in the evaporator bottoms during the neutral pH evaporation, but partitions to the condensate during alkaline evaporation. Disposition of both streams needs to consider the management of ammonia vapor and its release. Since this is an initial phase of testing, additional tasks related to evaporation methods are expected to be identified for development. These tasks likely include evaluation and testing of composition variability testing and evaluations, corrosion and erosion testing, slurry storage and immobilization investigations, and evaporator condensate disposition.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Anticonvulsant drug polytherapy stops status epilepticus and prevents neuronal loss in soman‐exposed rats

Introduction When treatment of status epilepticus (SE) is delayed, seizures become self‐sustaining and refractory to benzodiazepine therapy. Chemical warfare nerve agents (CWNAs), such as soman(GD), increase acetylcholine through inhibition of acetylcholine sterase and can lead to SE if seizures are not treated quickly and controlled. Prolonged seizures may lead to extensive neuropathology, spontaneous recurrent seizures and long‐term performance deficits. To identify a better treatment against pharmacoresistant seizures caused by GD exposure, we are using combinations of drugs aimed at reversing the effects of maladaptive receptor trafficking that follows CWNA exposure. Methods Rats were implanted with telemetry transmitters for continuous monitoring of EEG, body temperature and activity. After surgical recovery, rats were exposed to 1.2 LD 50 GD and treated 1 min later with atropine sulfate and the oxime HI‐6 and then 40 min after seizure onset with the anticonvulsant phenobarbital with or without the NMDA antagonist ketamine (PHE/KET), the benzodiazepine midazolam (PHE/MDZ) or a combination of all three drugs (PHE/KET/MDZ).Two weeks after exposure, rat brains were sectioned and stained for neuropathology assessments. Results Triple therapy with PHE/KET/MDZ reduced total time spent in seizures (SE +early recurrent seizures) in the first 72 h and reduced the number of spontaneous recurrent seizures compared to midazolam monotherapy. Tripletherapy also reduced the duration of toxic signs in the hours following GD‐exposure and prevented the development of hyperactivity that occurs in the weeks following exposure compared to GD/MDZ monotherapy. In addition, tripletherapy reduced loss of neurons in the piriform cortex and thalamus compared to GD/MDZ monotherapy. Conclusion Triple therapy with PHE/KET/MDZ hadanticonvulsant, antiepileptic, and neuroprotective effects compared with midazolammonotherapy in GD‐exposed rats. Triple therapy may be a highly effective approach against pharmacoresistant seizures, such as those caused by GD exposure, and may allow for administration of lower doses, resulting in fewer side effects than are often seen with large doses of individual drug therapies. Disclaimer: The views expressed in this abstract are those of the authors and do not reflect the official policy of the Department of the Army, Department of Defense, or the U.S. Government. The experimental protocol was approved by the Animal Care and Use Committee at the United States Army Medical Research Institute of Chemical Defense and all procedures were conducted in accordance with the principles stated in the Guide for the Care and Use of Laboratory Animals and the Animal Welfare Act of 1966 (P.L. 89‐544), as amended. Support or Funding Information KatieWalker, Erica Kundrick and Sean O'Brien were supported in part by an appointment to the Research Participation Program for the U.S. Army Medical Research and Materiel Command administered by the Oak Ridge Institute for Science and Education through an agreement between the U.S. Department of Energy and U.S. Army Medical Research and Materiel Command. This research was supported by the NIH U01 to Dr. Claude Wasterlain.

Stone, Michael↗
DOE OSTI2015

A User Guide to PARET/ANL (V.7.6) (Rev.1)

PARET was originally created in 1969 at what is now Idaho National Laboratory (INL), to analyze reactivity insertion events in research and test reactor cores cooled by light or heavy water, with fuel composed of either plates or pins. The use of PARET is also appropriate for fuel assemblies with curved fuel plates when their radii of curvatures are large with respect to the fuel plate thickness. The PARET/ANL version of the code has been developed at Argonne National Laboratory (ANL) under the sponsorship of the U.S. Department of Energy/NNSA, and has been used by the Reactor Conversion Program to determine the expected transient behavior of a large number of reactors. PARET/ANL models the various fueled regions of a reactor core as channels. Each of these channels consists of a single flat fuel plate/pin (including cladding and, optionally, a gap) with water coolant on each side. In slab geometry the coolant channels for a given fuel plate are of identical dimensions (mirror symmetry), but they can be of different thickness in each channel. There can be many channels, but each channel is independent and coupled only through reactivity feedback effects to the whole core. The time-dependent differential equations that represent the system are replaced by an equivalent set of finite-difference equations in space and time, which are integrated numerically. PARET/ANL uses fundamentally the same numerical scheme as RELAP5 for the time-integration of the point-kinetics equations. The one-dimensional thermal-hydraulic model includes temperature-dependent thermal properties of the solid materials, such as heat capacity and thermal conductivity, as well as the transient heat production and heat transfer from the fuel meat to the coolant. Temperature- and pressure-dependent thermal properties of the coolant such as enthalpy, density, thermal conductivity, and viscosity are also used in determining parameters such as friction factors and heat transfer coefficients. The code first determines the steady-state solution for the initial state. Then the solution of the transient is obtained by integration in time and space. Multiple heat transfer, DNB and flow instability correlations are available. The code was originally developed to model reactors cooled by an open loop, which was adequate for rapid transients in pool-type cores. An external loop model appropriate for Miniature Neutron Source Reactors (MNSR’s) was also added to PARET/ANL to model natural circulation within the vessel, heat transfer from the vessel to pool and heat loss by evaporation from the pool. PARET/ANL also contains models for decay heat after shutdown, control rod reactivity versus time or position, time-dependent pump flow, and loss-of-flow event with flow reversal as well as logic for trips on period, power, and flow. Feedback reactivity effects from coolant density changes and temperature changes are represented by tables. Feedback reactivity from fuel heat-up (Doppler Effect) is represented by a four-term polynomial in powers of fuel temperature. Photo-neutrons produced in beryllium or in heavy water may be included in the point-kinetics equations by using additional delayed neutron groups.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

A User Guide to PARET/ANL

PARET was originally created in 1969 at what is now Idaho National Laboratory (INL), to analyze reactivity insertion events in research and test reactor cores cooled by light or heavy water, with fuel composed of either plates or pins. The use of PARET is also appropriate for fuel assemblies with curved fuel plates when their radii of curvatures are large with respect to the fuel plate thickness. The PARET/ANL version of the code has been developed at Argonne National Laboratory (ANL) under the sponsorship of the U.S. Department of Energy/NNSA since the inception of the Reactor Conversion Program. Since Reduced Enrichment for Research and Test Reactors (RERTR) began in 1978, PARET/ANL has been benchmarked to experimental data including SPERT testing, and used to determine the expected transient behavior of a large number of reactors regardless of enrichment both inside and outside the Reactor Conversion Program. This document provides the pertinent information for the use of PARET/ANL Version 7.6. PARET/ANL models the various fueled regions of a reactor core as channels. Each of these channels consists of a single flat fuel plate/pin (including cladding and, optionally, a gap) with water coolant on each side. In slab geometry the coolant channels for a given fuel plate are of identical dimensions (mirror symmetry), but they can be of different thickness in each channel. There can be many channels, but each channel is independent and coupled only through reactivity feedback effects to the whole core. The time-dependent differential equations that represent the system are replaced by an equivalent set of finite-difference equations in space and time, which are integrated numerically. PARET/ANL uses fundamentally the same numerical scheme as RELAP5 for the time-integration of the point-kinetics equations. The one-dimensional thermal-hydraulic model includes temperature-dependent thermal properties of the solid materials, such as heat capacity and thermal conductivity, as well as the transient heat production and heat transfer from the fuel meat to the coolant. Temperature- and pressure-dependent thermal properties of the coolant such as enthalpy, density, thermal conductivity, and viscosity are also used in determining parameters such as friction factors and heat transfer coefficients. The code first determines the steady-state solution for the initial state. Then the solution of the transient is obtained by integration in time and space. Multiple heat transfer, DNB and flow instability correlations are available. The code was originally developed to model reactors cooled by an open loop, which was adequate for rapid transients in pool-type cores. An external loop model appropriate for Miniature Neutron Source Reactors (MNSR’s) was also added to PARET/ANL to model natural circulation within the vessel, heat transfer from the vessel to pool and heat loss by evaporation from the pool. PARET/ANL also contains models for decay heat after shutdown, control rod reactivity versus time or position, time-dependent pump flow, and loss-of-flow event with flow reversal as well as logic for trips on period, power, and flow. Feedback reactivity effects from coolant density changes and temperature changes are represented by tables. Feedback reactivity from fuel heat-up (Doppler Effect) is represented by a four-term polynomial in powers of fuel temperature. Photo-neutrons produced in beryllium or in heavy water may be included in the point-kinetics equations by using additional delayed neutron groups.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

A User Guide to PARET/ANL

PARET was originally created in 1969 at what is now Idaho National Laboratory (INL), to analyze reactivity insertion events in research and test reactor cores cooled by light or heavy water, with fuel composed of either plates or pins. The use of PARET is also appropriate for fuel assemblies with curved fuel plates when their radii of curvatures are large with respect to the fuel plate thickness. The PARET/ANL version of the code has been developed at Argonne National Laboratory (ANL) under the sponsorship of the U.S. Department of Energy/NNSA since the inception of the Reactor Conversion Program. Since Reduced Enrichment for Research and Test Reactors (RERTR) began in 1978, PARET/ANL has been benchmarked to experimental data including SPERT testing, and used to determine the expected transient behavior of a large number of reactors regardless of enrichment both inside and outside the Reactor Conversion Program. This document provides the pertinent information for the use of PARET/ANL Version 7.6. PARET/ANL models the various fueled regions of a reactor core as channels. Each of these channels consists of a single flat fuel plate/pin (including cladding and, optionally, a gap) with water coolant on each side. In slab geometry the coolant channels for a given fuel plate are of identical dimensions (mirror symmetry), but they can be of different thickness in each channel. There can be many channels, but each channel is independent and coupled only through reactivity feedback effects to the whole core. The time-dependent differential equations that represent the system are replaced by an equivalent set of finite-difference equations in space and time, which are integrated numerically. PARET/ANL uses fundamentally the same numerical scheme as RELAP5 for the time-integration of the point-kinetics equations. The one-dimensional thermal-hydraulic model includes temperature-dependent thermal properties of the solid materials, such as heat capacity and thermal conductivity, as well as the transient heat production and heat transfer from the fuel meat to the coolant. Temperature- and pressure-dependent thermal properties of the coolant such as enthalpy, density, thermal conductivity, and viscosity are also used in determining parameters such as friction factors and heat transfer coefficients. The code first determines the steady-state solution for the initial state. Then the solution of the transient is obtained by integration in time and space. Multiple heat transfer, DNB and flow instability correlations are available. The code was originally developed to model reactors cooled by an open loop, which was adequate for rapid transients in pool-type cores. An external loop model appropriate for Miniature Neutron Source Reactors (MNSR’s) was also added to PARET/ANL to model natural circulation within the vessel, heat transfer from the vessel to pool and heat loss by evaporation from the pool. PARET/ANL also contains models for decay heat after shutdown, control rod reactivity versus time or position, time-dependent pump flow, and loss-of-flow event with flow reversal as well as logic for trips on period, power, and flow. Feedback reactivity effects from coolant density changes and temperature changes are represented by tables. Feedback reactivity from fuel heat-up (Doppler Effect) is represented by a four-term polynomial in powers of fuel temperature. Photo-neutrons produced in beryllium or in heavy water may be included in the point-kinetics equations by using additional delayed neutron groups.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Modeling and Simulation of Xe-100-type Pebble Bed Gas-Cooled Reactor with SCALE

The US Department of Energy (DOE) announced the Advanced Reactor Demonstration Program (ARDP) to accelerate the deployment of advanced reactor concepts. Awardees of ARDP funds are expected to demonstrate the operation of an advanced reactor within 7 years of receiving the award. X-Energy’s advanced reactor concept, the Xe-100, was selected as one of two advanced reactor concepts to receive funding to demonstrate the operation of its high-temperature gas-cooled pebble-bed reactor before the end of this decade. As a result of this push to bring advanced reactors to maturation and commercialization, transition and deployment scenario studies are being performed under the Systems Analysis and Integration (SA&I) campaign within the DOE Office of Nuclear Energy (DOE-NE) to evaluate the transition of the current US commercial fleet of light-water reactors (LWRs) to a future fleet of advanced reactors consisting of a mix of ARDP type reactor concepts and advanced LWRs. To accurately evaluate the front- and back-end resource requirements, it is important to perform reactor physics calculations to determine the discharge burnup and isotopic content, fuel residence time, as well as other parameters. For this purpose, a summer project funded by the SA&I campaign allowed for the setup of SCALE models for full-core Xe-100 type high-temperature gas-cooled pebble-bed reactor and a Xe-100 type slice using publicly available information. The core-averaged equilibrium compositions and zone-wise equilibrium compositions for the slice and 3D models, respectively, were obtained following an iterative depletion method developed by Bostelmann et al. using SCALE’s reactor physics sequence TRITON. The slice model was used with TRITON to generate burnup-dependent cross section libraries at different temperatures which can be used with SCALE’s ORIGAMI code to rapidly determine fuel inventory and therefore to perform quick sensitivity studies on parameters such as the pebble location in the core. The SCALE/TRITON transport and depletion calculation for the Xe-100 type slice model indicates that the isotopic concentrations are in good agreement at 1,300 effective full power days (EFPD) for 235 U. An analysis of 236 U results match 239 Pu results would seem to indicate a typographical error in Mulder and Boyes wherein the reported results of 236 U and 239 Pu are reversed. In addition to SCALE/TRITON calculations, a new capability within SCALE/ORIGAMI for the simulation of pebble-bed reactors was used to study the burnup sensitivity with respect to the pebble pathway through the core. The SCALE/ORIGAMI results show that pebbles that travel closer to the reflector for the entire depletion history have a higher burnup than pebbles that travel through the middle of the core because of the higher thermal to fast flux ratio near the reflector. Consequently, a pebble’s burnup is strongly affected by the pebble’s pathway for each pass. Additional phenomena such as temperature distributions in the core and different travel times of the pebbles in the individual radial zones further affect the burnup distribution. The sensitivity of the discharge vector to the pebble pathways taken during each pass can be evaluated in the future using SCALE/ORIGAMI now that the SCALE inputs have been established.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Reverse-mode differentiation in arbitrary tensor network format: with application to supervised learning.

This paper describes an efficient reverse-mode differentiation algorithm for contraction operations of tensor networks that may have arbitrary and unconventional network topologies. The approach leverages the tensor contraction tree of Evenbly and Pfeifer (2014), which provides an instruction set for the contraction sequence of a network. We show that this tree can be efficiently leveraged for differentiation of a full tensor network contraction using a recursive scheme that exploits (1) the bilinear property of contraction and (2) the property that trees have single path from root to leaves. While differentiation of tensor-tensor contraction is already possible in most automatic differentiation packages, we show that exploiting these two additional properties in the specific context of contraction sequences can improve efficiency. Following a description of the algorithm and computational complexity analysis, we investigate its utility for gradient-based supervised learning for low-rank function recovery and for fitting real-world unstructured datasets. We demonstrate improved performance over alternating least-squares optimization approaches and the capability to handle heterogeneous and arbitrary tensor network formats. When compared to alternating minimization algorithms, we find that the gradient-based approach requires a smaller oversampling ratio (number of samples compared to number model parameters) for recovery. This increased efficiency extends to fitting unstructured data of varying dimensionality and when employing a variety of tensor network formats. Here, we show improved learning using the hierarchical Tucker method over the tensor-train in high-dimensional settings on a number of benchmark problems.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2024

Preliminary Design of Nuclear Reactor Heat Delivery Systems: Integration with Reference Oil Refinery, Methanol Synthesis, and Hydrogen Production

The Department of Energy’s (DOE) Integrated Energy Systems (IES) program is generating comprehensive analyses validating the opportunity for using nuclear energy in a variety of applications including future clean grids, providing heat for direct use, and providing heat to help reduce emissions in chemical commodities. This work focuses on the preliminary designs of thermal delivery systems that can integrate heat produced from a nuclear core to industrial processes. The key research question that needs to be answered is: what is the prospective method for integrating nuclear generated heat energy into non-electric applications that can facilitate combined heat and power operations by advanced nuclear reactor systems? This research is a composition of case studies showing preliminary conceptual designs for thermal delivery systems integrating advanced nuclear systems with a few industrial systems including high temperature steam electrolysis, a reference oil refinery, and potential future methanol systems that supplant some natural gas use with nuclear energy. Piping and instrumentation diagrams have been developed to show the conceptual integration of nuclear systems with representative industrial systems. Different features of the configurations are dependent on the specific integration requirements including energy source conditions, demand quantity, and require energy application conditions. Design concepts are validated using thermodynamic balance calculations to verify system performance including calculating system losses during transport. Key components: pumps and compressors, heat exchangers, and network piping are reported with key design information and sizing.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Advancing Fusion Research and Development at TAE Technologies Through INFUSE Program

The U.S. Department of Energy’s Innovation Network for Fusion Energy (INFUSE) program serves as a crucial catalyst by fostering public-private partnership that accelerates technological innovation for fusion energy research and development (R&D) in the private sector. Further, this article provides a comprehensive overview of the technical goals and accomplishments of projects awarded to TAE Technologies through the INFUSE program since 2019. We offer high-level perspectives on how these projects have contributed to fusion energy R&D, and we address key challenges encountered during these collaborations.

field-reversed configuration↗