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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Estimating Aboveground Biomass in Tropical Forests: Field Methods and Error Analysis for the Calibration of Remote Sensing Observations

Mapping and monitoring of forest carbon stocks across large areas in the tropics will necessarily rely on remote sensing approaches, which in turn depend on field estimates of biomass for calibration and validation purposes. Here, we used field plot data collected in a tropical moist forest in the central Amazon to gain a better understanding of the uncertainty associated with plot-level biomass estimates obtained specifically for the calibration of remote sensing measurements. In addition to accounting for sources of error that would be normally expected in conventional biomass estimates (e.g., measurement and allometric errors), we examined two sources of uncertainty that are specific to the calibration process and should be taken into account in most remote sensing studies: the error resulting from spatial disagreement between field and remote sensing measurements (i.e., co-location error), and the error introduced when accounting for temporal differences in data acquisition. We found that the overall uncertainty in the field biomass was typically 25% for both secondary and primary forests, but ranged from 16 to 53%. Co-location and temporal errors accounted for a large fraction of the total variance (>65%) and were identified as important targets for reducing uncertainty in studies relating tropical forest biomass to remotely sensed data. Although measurement and allometric errors were relatively unimportant when considered alone, combined they accounted for roughly 30% of the total variance on average and should not be ignored. Lastly, our results suggest that a thorough understanding of the sources of error associated with field-measured plot-level biomass estimates in tropical forests is critical to determine confidence in remote sensing estimates of carbon stocks and fluxes, and to develop strategies for reducing the overall uncertainty of remote sensing approaches.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2022

Representativeness assessment of the pan-Arctic eddy covariance site network and optimized future enhancements

Abstract. Large changes in the Arctic carbon balance are expected as warming linked to climate change threatens to destabilize ancient permafrost carbon stocks. The eddy covariance (EC) method is an established technique to quantify net losses and gains of carbon between the biosphere and atmosphere at high spatiotemporal resolution. Over the past decades, a growing network of terrestrial EC tower sites has been established across the Arctic, but a comprehensive assessment of the network's representativeness within the heterogeneous Arctic region is still lacking. This creates additional uncertainties when integrating flux data across sites, for example when upscaling fluxes to constrain pan-Arctic carbon budgets and changes therein. This study provides an inventory of Arctic (here > = 60∘ N) EC sites, which has also been made available online (https://cosima.nceas.ucsb.edu/carbon-flux-sites/, last access: 25 January 2022). Our database currently comprises 120 EC sites, but only 83 are listed as active, and just 25 of these active sites remain operational throughout the winter. To map the representativeness of this EC network, we evaluated the similarity between environmental conditions observed at the tower locations and those within the larger Arctic study domain based on 18 bioclimatic and edaphic variables. This allows us to assess a general level of similarity between ecosystem conditions within the domain, while not necessarily reflecting changes in greenhouse gas flux rates directly. We define two metrics based on this representativeness score: one that measures whether a location is represented by an EC tower with similar characteristics (ER1) and a second for which we assess if a minimum level of representation for statistically rigorous extrapolation is met (ER4). We find that while half of the domain is represented by at least one tower, only a third has enough towers in similar locations to allow reliable extrapolation. When we consider methane measurements or year-round (including wintertime) measurements, the values drop to about 1/5 and 1/10 of the domain, respectively. With the majority of sites located in Fennoscandia and Alaska, these regions were assigned the highest level of network representativeness, while large parts of Siberia and patches of Canada were classified as underrepresented. Across the Arctic, mountainous regions were particularly poorly represented by the current EC observation network. We tested three different strategies to identify new site locations or upgrades of existing sites that optimally enhance the representativeness of the current EC network. While 15 new sites can improve the representativeness of the pan-Arctic network by 20 %, upgrading as few as 10 existing sites to capture methane fluxes or remain active during wintertime can improve their respective ER1 network coverage by 28 % to 33 %. This targeted network improvement could be shown to be clearly superior to an unguided selection of new sites, therefore leading to substantial improvements in network coverage based on relatively small investments.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2014

Use of the portable decay gamma source feature in the R2Smesh code system for the accurate estimation of photon radiation dose fields around activated ITER components - Paper 111

The R2Smesh ('Rigorous 2-step') code system, developed by KIT for the determination of shut-down dose rate distribution on mesh grid, can be utilized to provide photon radiation dose field around activated components which are moved from the irradiation site in fusion reactor to another external location. R2Smesh couples MCNP transport calculations and FISPACT inventory calculations through suitable interfaces utilizing MCNP mesh tally capability. Neutron flux and the decay gamma source are provided with high resolution on the mesh grid superimposed to the real geometry to take into account the spatial variations of the flux and the decay gamma source distributions. This feature enables the decay gamma source distribution to be transferred from the irradiation site in the reactor to any other external location for the determination of photon flux and dose rate distributions. In this work, we applied the feature of the portable decay gamma source in R2Smesh code system to a dose rate analysis for a tunnel connecting the ITER tokamak complex building (TCB) and the hot cell facility building (HCFB). This paper presents the use of the portability feature of the decay gamma source in R2Smesh interface code system developed by KIT for a dose rate analysis of the ITER tunnel connecting TCB and HCFB. This application shows that the decay gamma source of the activated components could be transferred successfully from one coordinate system to another coordinate system and superimposed to the geometry of the components. This is very useful when the activated components need to be transported from the reactor to somewhere else for maintenance, testing or storage and the dose rate analysis is requested. Another portable feature, which could convert the 3D decay gamma source distribution generated by R2Smesh system to the MCNP 'SDEF' card source description and avoid the use of source sampling subroutine, is under consideration. With this feature, one decay gamma source result for the specific reactor components could be shared by different institutions who need to use the same decay gamma source for different analysis. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2020

Global methane emissions from coal mining to continue growing even with declining coal production

This paper presents projections of global methane emissions from coal mining under different coal extraction scenarios and with increasing mining depth through 2100. The paper proposes an updated methodology for calculating fugitive emissions from coal mining, which accounts for coal extraction method, coal rank, and mining depth and uses evidence-based emissions factors. A detailed assessment shows that coal mining-related methane emissions in 2010 were higher than previous studies show. This study also uses a novel methodology for calculating methane emissions from abandoned coal mines and represents the first estimate of future global methane emissions from those mines. The results show that emissions from abandoned mines increase faster than those from active ones. Using coal production data from six integrated assessment models, this study shows that by 2100 methane emissions from active underground mines increase by a factor of 4, while emissions from abandoned mines increase by a factor of 8. Abandoned mine methane emissions continue through the century even with aggressive mitigation actions.

coal mine methane, abandoned coal mine methane, em↗
DOE OSTI2023

A machine learning approach targeting parameter estimation for plant functional type coexistence modeling using ELM-FATES (v2.0)

Tropical forest dynamics play a crucial role in the global carbon, water, and energy cycles. However, realistically simulating the dynamics of competition and coexistence between different plant functional types (PFTs) in tropical forests remains a significant challenge. This study aims to improve the modeling of PFT coexistence in the Functionally Assembled Terrestrial Ecosystem Simulator (FATES), a vegetation demography model implemented in the Energy Exascale Earth System Model (E3SM) land model (ELM), ELM-FATES. Specifically, we explore (1) whether plant trait relationships established from field measurements can constrain ELM-FATES simulations and (2) whether machine learning (ML)-based surrogate models can emulate the complex ELM-FATES model and optimize parameter selections to improve PFT coexistence modeling. We conducted three ensembles of ELM-FATES experiments at a tropical forest site near Manaus, Brazil. By comparing the ensemble experiments without (Exp-CTR) and with (Exp-OBS) consideration of observed trait relationships, we found that accounting for these relationships slightly improves the simulations of water, energy, and carbon variables when compared to observations but degrades the simulation of PFT coexistence. Using ML-based surrogate models trained on Exp-CTR, we optimized the trait parameters in ELM-FATES and conducted another ensemble of experiments (Exp-ML) with these optimized parameters. The proportion of PFT coexistence experiments significantly increased from 21 % in Exp-CTR to 73 % in Exp-ML. After filtering the experiments that allow for PFT coexistence to agree with observations (within 15 % tolerance), 33 % of the Exp-ML experiments were retained, which is a significant improvement compared to the 1.4 % in Exp-CTR. Exp-ML also accurately reproduces the annual means and seasonal variations in water, energy, and carbon fluxes and the field inventory of aboveground biomass. This study represents a reproducible method that utilizes machine learning to identify parameter values that improve model fidelity against observations and PFT coexistence in vegetation demography models for diverse ecosystems. Our study also suggests the need for new mechanisms to enhance the robust simulation of coexisting plants in ELM-FATES and has significant implications for modeling the response and feedbacks of ecosystem dynamics to climate change.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2019

Aluminum Clad SNF: Technical Considerations and Challenges for Extended (>50 years) Dry Storage - 1900y

The U.S. Department of Energy (DOE) manages a significant inventory (>13 MTHM) of aluminum clad spent nuclear fuel (ASNF) using a combination of wet and dry storage at three primary sites; Hanford, Idaho National Laboratory (INL), and Savannah River Site (SRS). The majority of this ASNF is generated from U.S. and international research test reactors (RTR), examples include Advanced Test Reactor (ATR) at INL and High-Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). The Obama Administration suspension of the Yucca Mountain repository, issuance of the January 2013 DOE 'Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste,' delays in implementing the 'regionalization' SNF approach recommended in the 1995 'Department of Energy Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs Final Environmental Impact Statement,' and limited ASNF processing capability at SRS H-Canyon means that interim storage has become the de-facto strategy for managing ASNF. The 1995 Idaho Settlement Agreement requires the DOE to transfer all SNF stored at INL to interim dry storage prior to December 31, 2023 and wet storage capacity at the SRS is nearing capacity. This situation was recognized by the DOE Spent Nuclear Fuel Working Group (SNFWG) and at its June 2016 meeting formed a ASNF sub-working group comprised subject matter experts to evaluate this area and provide a report of findings. The ASNF sub-working group met many times over several month period to review the engineering and technical basis of existing and planned DOE ASNF dry storage facilities and additional studies on the behavior of ASNF under different conditions, including during reactor service. The ASNF sub-working group focused its review on RTR SNF. Based on these reviews, the ASNF sub-working group identified five knowledge gaps and technical data needs that should be addressed to ensure viability and safety of extended, greater than 50 years, sealed or vented interim dry storage of ASNF. These five areas include; behavior/chemistry of aluminum oxyhydroxide layer in dry storage, radiolytic gas generation properties of aluminum oxyhydroxide layers, effect of episodic breathing and radiolytic generation of corrosive gases during storage, performance of ASNF in existing dry storage systems, and effects of high temperature (>1000 C) drying on chemistry and behavior of aluminum oxyhydroxide layers. The ASNF sub-working group also recommended development of an action plan to identify technical and engineering activities needed to address the knowledge gaps and technical data needs. Results of the ASNF sub-working group are documented in 2017 report 'Aluminum Clad Spent Nuclear Fuel: Technical Considerations and Challenges for Extended (>50 years) Dry Storage,' DOE-ID/RPT-1575. The INL assembled a team comprised of twenty INL and Savannah River National Laboratory SNF management and subject matter experts to develop this action plan. Over the course of 2 months the team worked to develop the action plan. The plan identified and recommended six technical and engineering tasks needed to address the previously identified knowledge gaps and technical data needs. DOE Environmental Management Office of Technology Development is providing funding to the INL to execute these tasks. This paper will present and discuss the results of the ASNF sub-working group, tasks identified in the action plan, and experimental activities underway at INL and SRNL to address the action plan. (authors)

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

Signature Analysis Utilizing a Dynamic Molten Salt Reactor Model for MC&A

Moving from traditional fixed fuel nuclear reactor systems to a mobile, dynamic fuel system that has dissolved special nuclear material in a molten salt is a paradigm shift in several respects. One major consideration is how to develop effective nuclear material controls and accounting for these novel reactor systems. The Molten Salt Reactor Experiment is one example of a critical molten salt system and provided a significant reference library based on the documented effort. But it was low thermal power that was not intended to reflect a commercial scale electricity production design. Therefore, to facilitate and assist vendors with domestic licensing considerations, research is underway to identify methods that could be used for domestic safeguards approaches for these novel reactor systems. This research presents the results of a signature analysis of data generated from three simulated scenarios using the molten salt demonstration reactor model defined in the Transient Simulation Framework of Reconfigurable Modules. Each scenario provides 1 hour isotope inventories over a 180 day period. The scenarios investigated provide test cases to examine if direct gamma-ray spectroscopy of the fuel can be used to identify changes comparing a base case (no reactivity control and fixed fission contribution) to an insertion of reactivity (10 pcm no change in fission composition) and a change in fission composition. The analysis demonstrates that monitoring the total count rate in a highly collimated high-resolution photon energy spectrum is sensitive to perturbations imposed into the reactor model. The total photon count rate changes ≈2% for the fission composition change and ≈4.5% for the reactivity insertion compared to the base case. However, both scenarios show an increase in the total photon count rate. The total photon count rate can be used to identify changes due to power (number of fissions) but not due to a change in the material undergoing fission. To distinguish between the two cases of increased power, the photon spectrum would require an intensive analysis technique. A photon peak count rate ratio analysis could be used to identify changes in the fissile material fission generation in the core through identification of a static peak that shows little variation to the source of fission and a highly varying peak. The photon peak strength will ultimately be determined by the isotope’s fission yield. A preliminary analysis investigating the coupling of the isotope’s fission yield and its concentration in the fuel salt derived from the modeling has been performed. A ratio analysis of the photon peak count rates of 140 La to 99 Mo demonstrated that the reactivity insertion creates a distinct difference in the ratio compared to the fission composition change scenario.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Review and Summary of Oxide Thickness Data for Aluminum-Clad Spent Nuclear Fuel

The U.S. Department of Energy (DOE) owns a large inventory of aluminum-clad spent nuclear fuel (ASNF) in interim storage pending ultimate disposition, with more being generated by currently operating research reactors. Dry storage in sealed DOE Standard Canisters is being investigated as an approach for long-term interim storage and/or disposition in a repository for ASNF. A primary challenge for ASNF storage is the presence of aluminum (oxy)hydroxide layers formed on the cladding surfaces during water exposure in the reactor and in wet storage, which forms a reservoir of chemisorbed water not readily removed at low (<100°C) drying temperatures. Free, physisorbed, and chemisorbed water are all susceptible to radiolytic breakdown under irradiation and can release hydrogen gas. Identifying the likely range of (oxy)hydroxide loadings on ASNF that may be placed in dry storage will help to ensure that the impact of the (oxy)hydroxide is adequately accounted for while avoiding over conservatism and enable mitigation strategies to be implemented where needed. The current report summarizes information on (oxy)hydroxide thicknesses and characteristics from the literature and from recent measurements taken in the present research campaign. In general, corrosion studies have indicated that the corrosion and oxide buildup on aluminum are affected by numerous conditions, including pH, temperature and heat flux, coolant flowrates, irradiation (in-reactor vs. unirradiated tests), duration of water exposure, and the amount of oxide already on the surface. Some of these factors are interrelated (e.g., the local temperature, heat flux, and coolant flowrate). Observations of the impact of irradiation (in-reactor versus ex-reactor measurements) on corrosion and oxide thickness are mixed, but multiple studies indicate that there is a significant difference between in-reactor and ex-reactor corrosion kinetics, even when other operating conditions such as pH, flowrate, heat flux, etc., simulate those in a reactor.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

The power of simplification: Operator interface with the AP1000{sup R} during design-basis and beyond design-basis events

The AP1000{sup R} plant is an 1100-MWe pressurized water reactor with passive safety features and extensive plant simplifications that enhance construction, operation, maintenance, safety and cost. The passive safety features are designed to function without safety-grade support systems such as component cooling water, service water, compressed air or HVAC. The AP1000 passive safety features achieve and maintain safe shutdown in case of a design-basis accident for 72 hours without need for operator action, meeting the expectations provided in the European Utility Requirements and the Utility Requirement Document for passive plants. Limited operator actions may be required to maintain safe conditions in the spent fuel pool (SFP) via passive means. This safety approach therefore minimizes the reliance on operator action for accident mitigation, and this paper examines the operator interaction with the Human-System Interface (HSI) as the severity of an accident increases from an anticipated transient to a design basis accident and finally, to a beyond-design-basis event. The AP1000 Control Room design provides an extremely effective environment for addressing the first 72 hours of design-basis events and transients, providing ease of information dissemination and minimal reliance upon operator actions. Symptom-based procedures including Emergency Operating Procedures (EOPs), Abnormal Operating Procedures (AOPs) and Alarm Response Procedures (ARPs) are used to mitigate design basis transients and accidents. Use of the Computerized Procedure System (CPS) aids the operators during mitigation of the event. The CPS provides cues and direction to the operators as the event progresses. If the event becomes progressively worse or lasts longer than 72 hours, and depending upon the nature of failures that may have occurred, minimal operator actions may be required outside of the control room in areas that have been designed to be accessible using components that have been designed to be reliable in these conditions. The primary goal of any such actions is to maintain or refill the passive inventory available to cool the core, containment and spent fuel pool in the safety-related and seismically qualified Passive Containment Cooling Water Storage Tank (PCCWST). The seismically-qualified, ground-mounted Passive Containment Cooling Ancillary Water Storage Tank (PCCAWST) is also available for this function as appropriate. The primary effect of these actions would be to increase the coping time for the AP1000 during design basis events, as well as events such as those described above, from 72 hours without operator intervention to 7 days with minimal operator actions. These Operator actions necessary to protect the health and safety of the public are addressed in the Post-72 Hour procedures, as well as some EOPs, AOPs, ARPs and the Severe Accident Management Guidelines (SAMGs). Should the event continue to become more severe and plant conditions degrade further with indications of inadequate core cooling, the SAMGs provide guidance for strategies to address these hypothetical severe accident conditions. The AP1000 SAMG diagnoses and actions are prioritized to first utilize the AP1000 features that are expected to retain a damaged core inside the reactor vessel. Only one strategy is undertaken at any time. This strategy will be followed and its effectiveness evaluated before other strategies are undertaken. This is a key feature of both the symptom-oriented AP1000 EOPs and the AP1000 SAMGs which maximizes the probability of retaining a damaged core inside the reactor vessel and containment while minimizing the chances for confusion and human errors during implementation. The AP1000 SAMGs are simple and straight-forward and have been developed with considerable input from human factors and plant operations experts. Most importantly, and different from severe accident management strategies for other plants, the AP1000 SAMGs do not require diagnosis of the location of the core (i.e., whether reactor vessel failure has occurred). This is a fundamental consequence of the AP1000 In-Vessel Retention approach, which allows severe accident management to be based on fundamental principles (e.g. provide coolant as close as possible to the core) that do not change during a specific event. This eliminates the need for one of the more difficult diagnostic requirements, since reactor vessel failure does not directly relate to any measurable plant parameter, and differs from other designs in that an engineered failure of the pressure vessel' (e.g. core catcher) is never required. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Identification and Quantification of Beta-Emitters in Waste Packages by Segmented Gamma-Scanning - 20114

A commonly applied method for the non-destructive characterization of the radioactive inventory of waste packages (e.g. 200-l drums) is segmented gamma-scanning. The gamma-emitting inventory is identified via their characteristic lines in the measured spectra. The corresponding net peak areas allow for quantification applying appropriate models. The identification and quantification of beta-emitters present in the waste package require further evaluation. While some of them emit gamma-rays, too, which can be used for identification and quantification, pure beta-emitters contribute to the gamma-spectra only by their Bremsstrahlung-spectra. {sup 32}P, {sup 90}Sr and {sup 90}Y are typical representatives of these isotopes. A method is developed for identification and quantification of these pure beta-emitters in radioactive waste packages. Origin of the procedure is the evaluation of the measured gamma-spectrum. After the identification of all lines present in the spectrum (characteristic gamma-lines, single- and double-escape-lines, X-ray-lines, annihilation peak etc.) this information is used together with calibration data of the detection system and further a-priori information like matrix composition, cross-section data from databases etc. to simulate the spectra without the Bremsstrahlung-components by an appropriate method. This may be Monte Carlo simulations or, like in our case, a computer code taking into account mathematical descriptions of the physical effects like photo effect, (multiple) Compton scattering, backscattering etc. and the attenuation effects of the waste matrix. In any case, the resulting spectrum must simulate the real spectrum without its Bremsstrahlung-component as good as possible. Subtracting the simulated from the measured spectrum then results in the Bremsstrahlung-components caused by the beta-emitters. Prior to evaluation, a set of Bremsstrahlung-spectra is calculated and stored for different beta-emitters in different matrices using the Bethe-Heitler equation. This data set can be extended for new conditions, if necessary, and is allocated in a data base for future application. Applying a non-linear optimization using the generalized reduced gradient method for comparing the calculated with the derived Bremsstrahlung-spectra results in the identification of the beta-emitter and a factor, which is proportional to its activity. The proportionality is determined by previous calibration measurements. The application of the method is demonstrated by measurements on a 200-l mock-up drum and a real waste package. (authors)

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2018

Inventory and Waste Characterization Status Report and OWL Update.

This report represents completion of milestone deliverable M2SF-18SNO10309013 "Inventory and Waste Characterization Status Report and OWL Update that reports on FY2018 activities for the work package (WP) SF-18SNO1030901. This report provides the detailed final information for completed FY2018 work activities for WP SF-18SN01030901, and a summary of priorities for FY2019. This status report on FY2018 activities includes evaluations of waste form characteristics and waste form performance models, updates to the OWL development, and descriptions of the two planned management processes for the OWL. Updates to the OWL include an updated user's guide, additions to the OWL database content for wastes and waste forms, results of the Beta testing and changes implemented from it. There are two processes being planned in FY2018, which will be implemented in FY2019. One process covers methods for interfacing with the DOE SNF DB (DOE 2007) at INL on the numerous entries for DOE managed SNF, and the other process covers the management of updates to, and version control/archiving of, the OWL database. In FY2018, we have pursued three studies to evaluate/redefine waste form characteristics and/or performance models. First characteristic isotopic ratios for various waste forms included in postclosure performance studies are being evaluated to delineate isotope ratio tags that quantitatively identify each particular waste form. This evaluation arose due to questions regarding the relative contributions of radionuclides from disparate waste forms in GDSA results, particularly, radionuclide contributions of DOE-managed SNF vs HLW glass. In our second study we are evaluating the bases of glass waste degradation rate models to the HIP calcine waste form. The HIP calcine may likely be a ceramic matrix material, with multiple ceramic phases with/without a glass phase. The ceramic phases are likely to have different degradation performance from the glass portion. The distribution of radionuclides among those various phases may also be a factor in the radionuclide release rates. Additionally, we have an ongoing investigation of the performance behavior of TRISO particle fuels and are developing a stochastic model for the degradation of those fuels that accounts for simultaneous corrosion of the silicon carbide (SiC) layer and radionuclide diffusion through it. The detailed model of the TRISO particles themselves, will be merged with models of the degradation behavior(s) of the graphite matrix (either prismatic compacts or spherical "pebbles") containing the particles and the hexagonal graphite elements holding the compacts.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Handling Radioactive Waste from the Proton Accelerator Facility at the Paul Scherrer Institut (PSI) - Always Surprising? - 13320

The Paul Scherrer Institut (PSI) is the largest national research centre in Switzerland. Its multidisciplinary research is dedicated to a wide field in natural science and technology as well as particle physics. In this context, PSI is operating, amongst others, a large proton accelerator facility since more than 30 years. In two cyclotrons, protons are accelerated to high speeds and then guided along roughly 100 m of beam line to three different target stations to produce secondary particles like mesons and neutrons for experiments and a separately beam line for UCN. The protons induce spallation processes in the target materials, and also at other beam loss points along the way, with emission of protons, neutrons, hydrogen, tritium, helium, heavier fragments and fission processes. In particular the produced neutrons, due to their large penetration depth, will then interact also with the surrounding materials. These interactions of radiation with matter lead to activation and partly to contamination of machine components and the surrounding infrastructures. Maintenance, operation and decommissioning of installations generate inevitably substantial amounts of radioactive operational and dismantling waste like targets, magnets, collimators, shielding (concrete, steel) and of course secondary waste. To achieve an optimal waste management strategy for interim storage or final disposal, radioactive waste has to be characterized, sorted and treated. This strategy is based on radiation protection demands, raw waste properties (size, material, etc.), and requirements to reduce the volume of waste, mainly for legal and economical reasons. In addition, the radiological limitations for transportation of the waste packages to a future disposal site have to be taken into account, as well as special regulatory demands. The characterization is a task of the waste producer. The conditioning processes and quality checks for radioactive waste packages are part of an accredited waste management process of PSI, especially of the Section Dismantling and Waste Management. Strictly proven and accepted methods needed to be developed and enhanced for safe treatment, transport, conditioning and storage. But in the field of waste from research activities, individual and new solutions have to be found in an increasingly growing administrative environment. Furthermore, a wide variety of components, with a really large inventory of radioactive nuclides, has to be handled. And there are always surprising challenges concerning the unusual materials or the nuclide inventory. In case of the operational and dismantling radioactive accelerator waste, the existing conditioning methods are in the process of a continuous enhancement - technically and administratively. The existing authorized specifications of conditioning processes have to be extended to optimize and fully describe the treatment of the inevitably occurring radioactive waste from the accelerator facility. Additional challenges are the changes with time concerning the legal and regulatory requirements - or do we have to consider it as business as usual? This paper gives an overview of the current practices in radioactive waste management and decommissioning of the existing operational accelerator waste. (authors)

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

Final Scientific/Technical Report: Characterizing the Response of the Cascadia Margin Gas Hydrate Reservoir to Bottom Water Warming Along the Upper Continental Slope

The objective of this project is to understand the response of the WA margin gas hydrate system to contemporary warming of bottom water along the upper continental slope. Through pre-cruise analysis and modeling of archive and recent geophysical and oceanographic data, we (1) inventoried bottom simulating reflectors along the WA margin and defined the upper limit of gas hydrate stability, (2) refined margin-wide estimates of heat flow and geothermal gradients, (3) characterized decadal scale temporal variations of bottom water temperatures at the upper continental slope of the Washington margin, and (4) used numerical simulations to provide quantitative estimates of how the shallow boundary of methane hydrate stability responds to modern environmental change. These pre-cruise results provided the context for a systematic geophysical and geochemical survey of methane seepage along the upper continental slope from 48° to 46°N during a 10-day field program on the R/V Thompson from October 10-19, 2014. This systematic inventory of methane emissions along this climate-sensitive margin corridor and comprehensive sediment and water column sampling program provided data and samples for Phase 3 of this project that focused on determining fluid and methane sources (deep-source vs. shallow; microbial, thermogenic, gas hydrate dissociation) within the sediment, and how they relate to contemporary intermediate water warming. During the 2014 research expedition, we sampled nine seep sites between ~470 and 520 m water depth, within the zone of predicted methane hydrate retreat over the past 40 years. We imaged 22 bubble plumes with heights commonly rising to ~300 meters below sea level with one reaching near the sea surface. We collected 22 gravity cores and 20 CTD/hydrocasts from the 9 seeps and at background locations (no acoustic evidence of seepage) within the depth interval of predicted downslope retreat of the methane hydrate stability zone. Approximately 300 pore water samples were extracted from the gravity cores, and the pore water was analyzed for a comprehensive suite of solutes, gases, and stable isotope ratios. This comprehensive geochemical dataset was used to characterize the fluid and gas source(s) at each of the seep sites surveyed. The primary results of this project are: 1) Bottom simulating reflector-derived heat flow values decrease from 95 mW/m2 10 km east of the deformation front to ~60 mW/m2 60 km landward of the deformation front, with anomalously low values of ~25 mW/m2 on a prominent mid-margin terrace off central Washington. 2) The temperature of the incoming sediment/ocean crust interface at the deformation front ranges between 164-179 oC off central Washington, and the 350 oC isotherm at the top of the subducting ocean crust occurs 95 km landward of the deformation front. Differences between BSR-derived heat flow and modeled conductive heat flow suggest mean upward fluid flow rates of 0.4 cm/yr across the margin, with local regions (e.g. fault zones) exhibiting fluid flow rates up to 3.5 cm/yr. 3) A compilation of 2122 high-resolution CTD, glider, and Argo float temperature profiles spanning the upper continental slope of the Washington margin from the years 1968 to 2013 show a long-term warming trend that ranges from 0.006-0.008 oC/yr. Based on this long-term bottom water warming, we developed a 2-D thermal model to simulate the change in sediment temperature distribution over this period, along with the downslope retreat of the methane hydrate stability field. Over the 43 years of the simulation, the thermal disturbance propagated 30 m into the sediment column, causing the base of the methane hydrate stability field to shoal ~13 m and to move ~1 km downslope. 4) A preliminary analysis of seafloor observations and mid-water column acoustic data to detect bubble plumes was used to characterize the depth distribution of seeps along the Cascadia margin. These results indicate high bubble plume densities along the continental shelf at water depths <180 m and at the upper limit of methane hydrate stability along the Washington margin. 5) The majority of the seeps cored during the 2014 research expedition on the R/V Thompson contained abundant authigenic carbonate indicating that they are locations of long-lived seepage rather than emergent seep systems related to methane hydrate dissociation. Despite the evidence for enhanced methane seepage at the upper limit of methane hydrate stability along the Washington margin, we found no unequivocal evidence for active methane hydrate dissociation as a source of fluid and gas at the seeps surveyed. The pore fluid and bottom water chemistry shows that the seeps are fed by a variety of fluid and methane sources, but that methane hydrate dissociation, if occurring, is not widespread and is only a minor source (below the detection limit of our methods). Collectively, these results provide a significant advance in our understanding of the thermal structure of the Cascadia subduction zone and the fluid and methane sources feeding seeps along the upper continental slope of the Washington-sector of the Cascadia margin. Though we did not find unequivocal evidence for methane hydrate dissociation as a source of water and methane at the upper pressure-temperature limit of methane hydrate stability at present, continued warming of North Pacific Intermediate Water in the future has the potential to impact the methane hydrate reservoir in sediments at greater depths along the slope. Thus, this study provides a strong foundation and the necessary characterization of the background state of seepage at the upper limit of methane hydrate stability for future investigations of this important process.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2024

Estimating Total Methane Emissions from the Denver-Julesburg Basin Using Bottom-Up Approaches

Methane is a powerful greenhouse gas with a 25 times higher 100-year warming potential than carbon dioxide and is a target for mitigation to achieve climate goals. To control and curb methane emissions, estimates are required from the sources and sectors which are typically generated using bottom-up methods. However, recent studies have shown that national and international bottom-up approaches can significantly underestimate emissions. In this study, we present three bottom-up approaches used to estimate methane emissions from all emission sectors in the Denver-Julesburg basin, CO, USA. Our data show emissions generated from all three methods are lower than historic measurements. A Tier 1/2 approach using IPCC emission factors estimated 2022 methane emissions of 358 Gg (0.8% of produced methane lost by the energy sector), while a Tier 3 EPA-based approach estimated emissions of 269 Gg (0.2%). Using emission factors informed by contemporary and region-specific measurement studies, emissions of 212 Gg (0.2%) were calculated. The largest difference in emissions estimates were a result of using the Mechanistic Air Emissions Simulator (MAES) for the production and transport of oil and gas in the DJ basin. The MAES accounts for changes to regulatory practice in the DJ basin, which include comprehensive requirements for compressors, pneumatics, equipment leaks, and fugitive emissions, which were implemented to reduce emissions starting in 2014. The measurement revealed that normalized gas loss is predicted to have been reduced by a factor of 20 when compared to 10-year-old normalization loss measurements and a factor of 10 less than a nearby oil and production area (Delaware basin, TX); however, we suggest that more measurements should be made to ensure that the long-tail emission distribution has been captured by the modeling. This study suggests that regulations implemented by the Colorado Department of Public Health and Environment could have reduced emissions by a factor of 20, but contemporary regional measurements should be made to ensure these bottom-up calculations are realistic.

03 NATURAL GAS↗
DOE OSTI2011

Advanced Nanostructured Molecular Sieves for Energy Efficient Industrial Separations

Due to the very small relative volatility difference between propane and propylene, current propane/propylene separation by distillation requires very tall distillation towers (150-250 theoretical plates) and large reflux ratios (up to 15), which is considered to be the most energy consuming large-scale separation process. Adsorptive separation processes are widely considered to be more energy-efficient alternatives to distillation. However, slow diffusion kinetics/mass transport rate through the adsorbent bed often limits the performance of such processes, so further improvements are possible if intra-particle mass transfer rates can be improved. Rive Technology, Inc. is developing and commercializing its proprietary mesoporous zeolite technology for catalysis and separation. With well-controlled intracrystalline mesoporosity, diffusion kinetics through such mesoporous zeolite based catalysts is much improved relative to conventional zeolites, leading to significantly better product selectivity. This 'proof-of-principle' project (DE-EE0003470) is intended to demonstrate that Rive mesoporous zeolite technology can be extended and applied in adsorptive propane/propylene separation and lead to significant energy saving compared to the current distillation process. In this project, the mesoporous zeolite Y synthesis technology was successfully extended to X and A zeolites that are more relevant to adsorbent applications. Mesoporosity was introduced to zeolite X and A for the first time while maintaining adequate adsorption capacity. Zeolite adsorbents were tested for liquid phase separation performance using a pulse flow test unit and the test results show that the separation selectivity of the mesoporous zeolite adsorbent is much closer to optimal for a Simulated Moving Bed (SMB) separation process and the enhanced mesoporosity lead to >100% increase of overall mass transport rate for propane and propylene. These improvements will significantly improve the performance of an adsorptive separation unit for propane/propylene separation compared with traditional zeolite adsorbents. The enhanced transport will allow for more efficient utilization of a given adsorbent inventory by reducing process cycle time, allowing a faster production rate with a fixed amount of adsorbent or smaller adsorbent inventory at a fixed production rate. Smaller adsorbent inventory would also lead to significant savings in the capital cost due to smaller footprint of the equipment. Energy consumption calculation, based on the pulse test results for rived NaX zeolite adsorbent, of a hypothetical moderate-scale SMB propane/propylene separation plant that processes 6000 BPSD refinery grade propylene (70% propylene) will consume about 60-80% less energy (both re-boiler and condenser duties) compared to a C3 splitter that process the same amount of feed. This energy saving also translates to a reduction of 30,000-35,000 tons of CO 2 emission per year at this moderate processing rate. The enhancement of mass transport achievable by introduction of controlled mesoporosity to the zeolite also opens the door for the technology to be applied to several other adsorption separation processes such as the separation of xylene isomers by SMB, small- and large scale production of O 2 /N 2 from air by pressure swing adsorption, the separation of CO 2 from natural gas at natural gas wellheads, and the purification of ultra-high purity H 2 from the off gas produced by steam-methane-reforming.

03 NATURAL GAS↗
DOE OSTI2019

Probability of Pinching Failure in PWR SNF after Long-Term Storage

A probabilistic assessment was performed on pinching failure of PWR SNF rod cladding under accidental SNF cask transport conditions. a set of 3,000 rod pinching scenarios was analyzed, with different geometrical, material, and loading parameters, using a Latin Hypercube Sampling (LHS). The investigated accidents include train derailments leading to cask collisions or cask drops. Pinching failure was related to cladding offset strain, defined as the plastic deformation interval in radial cladding direction normalized with respect to the outer cladding diameter. Finite element (FE) models were used to compute SNF rod deformations and the offset strain demand OSD in each of the assessed pinching scenarios. The numerical models account for the contribution of the degraded fuel pellets to rod pinching stiffness. The offset strain capacity OSC was computed using regression and interpolation models based on ring compression (RCT) test results, and account for hydride-related effects on cladding ductility. The results show a low probability of pinching failure. Hydride-related cladding embrittlement affects the structural integrity of PWR SNF rods after long-term dry storage of up to 300 years. Radial hydride patterns or large amounts of circumferentially oriented hydrides reduce the cladding ductile deformation capability when pinched at low material temperatures. The probability of hydride-related pinching failure of PWR cladding is estimated after long-term dry storage to address: How do PWR SNF rod conditions like hydrogen content, vacuum drying peak cladding hoop stress σ{sub θ}, and cladding material temperature affect cladding ductility? Is hydride-related cladding embrittlement a material degradation mechanism that affects the structural integrity of PWR SNF rods under pinching loading? What are the controlling parameters for deformation demands in PWR SNF rods under pinching loading? Is the pinching load amplitude an important parameter that controls the probability of pinching failure? How do parameters like pellet-cladding gap or fuel pellet degradation affect the pinching loading capacity of PWR SNF rods? Statistical modelling of PWR SNF rod conditions in U.S. inventory: A pseudo rod database was generated based on EIA database. Analysis of train accident scenarios: a probability density function (pdf) of a pinching load distribution P{sub 9m} of a 9-m cask drop is linearly scaled to generate a more representative: distribution P{sub AS}. The new distribution considers train collision velocities, cask drops for on- and off-bridge accidents, and cask impact orientation. LHS generates 3,000 rod pinching scenarios based on nine independent geometrical, material, and loading parameters. Probabilistic FE Analysis: The LHS realizations are used to generate 3,000 FE models. The FE analysis is conducted for P{sub 9m}, and P{sub AS}. The workflow is automatized. The OSC is computed with regression and interpolation models, as a function of the hydrogen content C{sub H}, vacuum drying peak σ{sub θ}, and SNF temperature T{sub FR}: The regression data was obtained in RCTs on PWR cladding. The OSD is derived from the cladding deformations in the FE models. Cladding failure was identified via a comparison of OSC and OSD. A 300 and a 30 years-of-dry-storage duration is investigated: The dry-storage period affects the expected SNF temperature and consequently, the OSC and the OSD. The material temperature LHS population is adapted to account for a SNF temperature variation. The probability of hydride-related, ductile cladding failure during a severe transportation accident is not significantly affected by the pinching load amplitude: The probability is slightly reduced from a range of less than 0.0333% to 0.1667% for P{sub 9m}, to a range of less than 0.0333% to 0.1333% for P{sub AS}, depending on the vacuum drying peak cladding hoop stress modelling assumptions. Further, the SNF dry-storage duration does not significantly affect the probability of ductile pinching failure. Based on available train accidents data, the pinching failure probability for a 5,000 km long transportation route ranges from less than 1.62 E-7 to a value of 6.47 E-7, depending on the vacuum drying peak cladding hoop stress modelling assumptions. The pinching load amplitude is not a controlling parameter for the probability of ductile cladding pinching failure, and does not control the cladding deformations once pellet-cladding contact is established, The pellet-cladding gap width, however, plays a key role. Fuel pellet degradation only affects the deformation capacity, if the pellet stiffness is significantly reduced (more than 90%). Only highly embrittled cladding is likely to fail. The deformation demands on the cladding under pinching loading are small.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2011

A Greenhouse-Gas Information System: Monitoring and Validating Emissions Reporting and Mitigation

Current GHG-mitigating regimes, whether internationally agreed or self-imposed, rely on the aggregation of self-reported data, with limited checks for consistency and accuracy, for monitoring. As nations commit to more stringent GHG emissions-mitigation actions and as economic rewards or penalties are attached to emission levels, self-reported data will require independent confirmation that they are accurate and reliable, if they are to provide the basis for critical choices and actions that may be required. Supporting emissions-mitigation efforts and agreements, as well as monitoring energy- and fossil-fuel intensive national and global activities would be best achieved by a process of: (1) monitoring of emissions and emission-mitigation actions, based, in part, on, (2) (self-) reporting of pertinent bottom-up inventory data, (3) verification that reported data derive from and are consistent with agreed-upon processes and procedures, and (4) validation that reported emissions and emissions-mitigation action data are correct, based on independent measurements (top-down) derived from a suite of sensors in space, air, land, and, possibly, sea, used to deduce and attribute anthropogenic emissions. These data would be assessed and used to deduce and attribute measured GHG concentrations to anthropogenic emissions, attributed geographically and, to the extent possible, by economic sector. The validation element is needed to provide independent assurance that emissions are in accord with reported values, and should be considered as an important addition to the accepted MRV process, leading to a MRV&V process. This study and report focus on attributes of a greenhouse-gas information system (GHGIS) needed to support MRV&V needs. These needs set the function of such a system apart from scientific/research monitoring of GHGs and carbon-cycle systems, and include (not exclusively): the need for a GHGIS that is operational, as required for decision-support; the need for a system that meets specifications derived from imposed requirements; the need for rigorous calibration, verification, and validation (CV&V) standards, processes, and records for all measurement and modeling/data-inversion data; the need to develop and adopt an uncertainty-quantification (UQ) regimen for all measurement and modeling data; and the requirement that GHGIS products can be subjected to third-party questioning and scientific scrutiny. This report examines and assesses presently available capabilities that could contribute to a future GHGIS. These capabilities include sensors and measurement technologies; data analysis and data uncertainty quantification (UQ) practices and methods; and model-based data-inversion practices, methods, and their associated UQ. The report further examines the need for traceable calibration, verification, and validation processes and attached metadata; differences between present science-/research-oriented needs and those that would be required for an operational GHGIS; the development, operation, and maintenance of a GHGIS missions-operations center (GMOC); and the complex systems engineering and integration that would be required to develop, operate, and evolve a future GHGIS. Present monitoring systems would be heavily relied on in any GHGIS implementation at the outset and would likely continue to provide valuable future contributions to GHGIS. However, present monitoring systems were developed to serve science/research purposes. This study concludes that no component or capability presently available is at the level of technological maturity and readiness required for implementation in an operational GHGIS today. However, purpose-designed and -built components could be developed and implemented in support of a future GHGIS. The study concludes that it is possible to develop and provide a capability-driven prototype GHGIS, as part of a Phase-1 effort, within three years from project-funding start, that would make use of and integrate existing sensing and system capabilities. As part of a Phase-2 effort, a requirements-driven, operational GHGIS could be developed, within ten years from project funding start. That schedule is driven by the development and long lead-times for some system components. The two efforts would be focused on different deliverables but could commence concurrently, to save time, if that was deemed desirable. We note that, developing and supporting an operational GHGIS will require a new approach and management, sustained funding and other support, as well as technical advances and development of purpose-built components that meet the requisite specifications. A functioning GHGIS will provide the basis for reasoned choices on how best to respond to rising GHG levels, especially when proposed U.S. actions are compared with or conditioned on the actions of other nations.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2016

The resolve survey atomic gas census and environmental influences on galaxy gas reservoirs

We present the H i mass inventory for the REsolved Spectroscopy Of a Local VolumE (RESOLVE) survey, a volume-limited, multi-wavelength census of >1500 z = 0 galaxies spanning diverse environments and complete in baryonic mass down to dwarfs of ∼10{sup 9} M{sub ⊙}. This first 21 cm data release provides robust detections or strong upper limits (1.4M {sub H i} < 5%–10% of stellar mass M {sub *}) for ∼94% of RESOLVE. We examine global atomic gas-to-stellar mass ratios (G/S) in relation to galaxy environment using several metrics: group dark matter halo mass M {sub h}, central/satellite designation, relative mass density of the cosmic web, and distance to the nearest massive group. We find that at fixed M {sub *}, satellites have decreasing G/S with increasing M {sub h} starting clearly at M {sub h} ∼ 10{sup 12} M{sub ⊙}, suggesting the presence of starvation and/or stripping mechanisms associated with halo gas heating in intermediate-mass groups. The analogous relationship for centrals is uncertain because halo abundance matching builds in relationships between central G/S, stellar mass, and halo mass, which depend on the integrated group property used as a proxy for halo mass (stellar or baryonic mass). On larger scales G/S trends are less sensitive to the abundance matching method. At fixed M {sub h} ≤ 10{sup 12} M{sub ⊙}, the fraction of gas-poor centrals increases with large-scale structure density. In overdense regions, we identify a rare population of gas-poor centrals in low-mass (M {sub h} < 10{sup 11.4} M{sub ⊙}) halos primarily located within ∼1.5× the virial radius of more massive (M {sub h} > 10{sup 12} M{sub ⊙}) halos, suggesting that gas stripping and/or starvation may be induced by interactions with larger halos or the surrounding cosmic web. We find that the detailed relationship between G/S and environment varies when we examine different subvolumes of RESOLVE independently, which we suggest may be a signature of assembly bias.

79 ASTRONOMY AND ASTROPHYSICS↗