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At least 523 records · Page 29
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Aerosol Physics Considerations for Using Cerium Oxide CeO 2 as a Surrogate for Plutonium Oxide PuO 2 in Airborne Release Fraction Measurements for Storage Container Investigations

Cerium oxide (CeO2) dust is recommended as a surrogate for plutonium oxide (PuO2) in airborne release fraction experiments. The total range of applicable particle sizes for PuO2 extends from 0.0032 μm (the diameter of a single PuO2 molecule) to 10 μm (the defined upper boundary for respirable particles). For particulates with a physical particle diameter of 1.0 μm, the corresponding aerodynamic diameters for CeO2 and PuO2 are 2.7 μm and 3.4 μm, respectively. Cascade impactor air samplers are capable of measuring the size distributions of CeO2 or PuO2 particulates. In this document, the aerodynamic diameters for CeO2 and PuO2 were calculated for seven different physical diameters (0.0032, 0.02, 0.11, 0.27, 1.0, 3.2, and 10 μm). For cascade impactor measurements, CeO2 and PuO2 particulates with the same physical diameter would be collected onto the same or adjacent collection substrates. The difference between the aerodynamic diameter of CeO2 and PuO2 particles (that have the same physical diameter) is 39% of the resolution of a twelve-stage MSP Inc. 125 cascade impactor, and 34% for an eight-stage Andersen impactor. An approach is given to calculate the committed effective dose (CED) coefficient for PuO2 aerosol particles, compared to a corresponding aerodynamic diameter of CeO2 particles. With this approach, use of CeO2 as a surrogate for PuO2 material would follow a direct conversion based on a molar equivalent. In addition to the analytical information developed for this document, several US national labs have published articles about the use of CeO2 as a PuO2 surrogate. Different physical and chemical aspects were considered by these investigators, including thermal properties, ceramic formulations, cold pressing, sintering, molecular reactions, and mass loss in high temperature gas flows. All of those US national lab studies recommended the use of CeO2 as a surrogate material for PuO2.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2017

Atmospheric Tracer Depletion Testing for Unfiltered Air In-Leakage Determination at the Wolf Creek Nuclear Power Plant

Atmospheric Tracer Depletion tests were conducted at the Wolf Creek Nuclear Power Plant to quantify the unfiltered in-leakage (UI) into the Control Room (CR), Control Building (CB), and Equipment Rooms (ER) at the Wolf Creek Nuclear Power Plant. Wolf Creek has two independent charcoal filter Emergency Ventilation Systems (EVS) that can be used to purify air entering the control building and control room. The Bravo System contains a filtration system in Room 1501 in the Auxiliary Building for the Control Room and another filtration system (FGK02B) on Elevation 2016 for the Control Building. The Alpha system contains a filtration system in Room 1512 in the Auxiliary Building for the Control Room and another filtration system (FGK02A) on Elevation 2016 for the Control Building. The Atmospheric Tracer Depletion (ATD) test is a technique to measure in-leakage using the concentration of perfluorocarbon compounds that have a constant atmospheric background. These levels are present in the Control Room and Control Building under normal operating conditions. When air is supplied by either of the EVS, most of the PFTS are removed by the charcoal filters. If the concentrations of the PFTs measured in protected areas are the same as the levels at the output of the EVS, the in-leakage of outside air into the protected area would be zero. If the concentration is higher in the protected area than at the output of the filter system, there is in-leakage and the in-leakage can be quantified by the difference. Sampling was performed using state-of-the-art Brookhaven Atmospheric Tracer Samplers (BATS) air sampling equipment and analysis performed on Brookhaven National Laboratory (BNL) dedicated PFT analytical systems. In the Alpha test two tracers PMCH and mcPDCH were used to determine in-leakage into the control building. The analytical system was tuned to maximize sensitivity after initial analysis of the Alpha test. The increased sensitivity permitted accurate quantification of five isomers of the PFT PDCH (mtPDCH, pcPDCH, otPDCH, mcPDCH, and ptPDCH). These isomers were quantified in the low concentration samples in the Alpha test and in all samples in the Bravo test. The best estimates of UI (Rui) for the four zones are provided in Table ES-1. For the CB, this estimate averages the four tracers at the four elevations. For the CR, this estimate uses the four sampling units located in the Control Room.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2017

Atmospheric Tracer Depletion Testing for Unfiltered Air In-Leakage Determination at the Wolf Creek Nuclear Power Plant

Atmospheric Tracer Depletion tests were conducted at the Wolf Creek Nuclear Power Plant to quantify the unfiltered in-leakage (UI) into the Control Room (CR), Control Building (CB), and Equipment Rooms (ER) at the Wolf Creek Nuclear Power Plant. Wolf Creek has two independent charcoal filter Emergency Ventilation Systems (EVS) that can be used to purify air entering the control building and control room. The Bravo System contains a filtration system in Room 1501 in the Auxiliary Building for the Control Room and another filtration system (FGK02B) on Elevation 2016 for the Control Building. The Alpha system contains a filtration system in Room 1512 in the Auxiliary Building for the Control Room and another filtration system (FGK02A) on Elevation 2016 for the Control Building.The Atmospheric Tracer Depletion (ATD) test is a technique to measure in-leakage using the concentration of perfluorocarbon compounds that have a constant atmospheric background. These levels are present in the Control Room and Control Building under normal operating conditions. When air is supplied by either of the EVS, most of the PFTS are removed by the charcoal filters. If the concentrations of the PFTs measured in protected areas are the same as the levels at the output of the EVS, the in-leakage of outside air into the protected area would be zero. If the concentration is higher in the protected area than at the output of the filter system, there is in-leakage and the in-leakage can be quantified by the difference.Sampling was performed using state-of-the-art Brookhaven Atmospheric Tracer Samplers (BATS) air sampling equipment and analysis performed on Brookhaven National Laboratory (BNL) dedicated PFT analytical systems. In the Alpha test two tracers PMCH and mcPDCH were used to determine in-leakage into the control building. The analytical system was tuned to maximize sensitivity after initial analysis of the Alpha test. The increased sensitivity permitted accurate quantification of five isomers of the PFT PDCH (mtPDCH, pcPDCH, otPDCH, mcPDCH, and ptPDCH). These isomers were quantified in the low concentration samples in the Alpha test and in all samples in the Bravo test.The best estimates of UI (Rui) for the four zones are provided in Table ES-1. For the CB, this estimate averages the four tracers at the four elevations. For the CR, this estimate uses the four sampling units located in the Control Room.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2016

Status on the Development of a Modeling and Simulation Framework for the Economic Assessment of Nuclear Hybrid Energy

Continued effort to design and build a modeling and simulation framework to assess the economic viability of Nuclear Hybrid Energy Systems (NHES) was undertaken in fiscal year (FY) 2016. The purpose of this report is to document the various tasks associated with the development of such a framework and to provide a status of their progress. Several tasks have been accomplished. First, a synthetic time history generator has been developed in RAVEN, which consists of Fourier series and autoregressive moving average model. The former is used to capture the seasonal trend in historical data, while the latter is to characterize the autocorrelation in residue time series (e.g., measurements with seasonal trends subtracted). As demonstration, both synthetic wind speed and grid demand are generated, showing matching statistics with database. In order to build a design and operations optimizer in RAVEN, a new type of sampler has been developed with highly object-oriented design. In particular, simultaneous perturbation stochastic approximation algorithm is implemented. The optimizer is capable to drive the model to optimize a scalar objective function without constraint in the input space, while the constraints handling is a work in progress and will be implemented to improve the optimization capability. Furthermore, a simplified cash flow model of the performance of an NHES in the electric market has been developed in Python and used as external model in RAVEN to confirm expectations on the analysis capability of RAVEN to provide insight into system economics and to test the capability of RAVEN to identify limit surfaces. Finally, an example calculation is performed that shows the integration and proper data passing in RAVEN of the synthetic time history generator, the cash flow model and the optimizer. It has been shown that the developed Python models external to RAVEN are able to communicate with RAVEN and each other through the newly developed RAVEN capability called “EnsembleModel”.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2017

Particle integrity, sampling, and application of a DNA-tagged tracer for aerosol transport studies

Aerosols are an ever-present part of our daily environment and have extensive effects on both human and environmental health. Particles in the inhalable range (1-10 μm diameter) are of particular concern because their deposition in the lung can lead to a variety of illnesses including allergic reactions, viral or bacterial infections, and cancer. Understanding the transport of inhalable aerosols across both short and long distances is necessary to predict human exposures to aerosols. To assess the transport of hazardous aerosols, surrogate tracer particles are required to measure their transport through occupied spaces. These tracer particles must not only possess similar transport characteristics to those of interest but also be easily distinguished from the background at low levels and survive the environmental conditions of the testing environment. A previously-developed DNA-tagged particle (DNATrax), composed of food-grade sugar and a DNA oligonucleotide as a “barcode” label, shows promise as a new aerosol tracer. Herein, the use of DNATrax material is validated for use in both indoor and outdoor environments. Utilizing passive samplers made of materials commonly found in indoor environments followed by quantitative polymerase chain reaction (qPCR) assay for endpoint particle detection, particles detection was achieved up to 90 m from the aerosolization location and across shorter distances with high spatial resolution. The unique DNA label and PCR assay specificity were leveraged to perform multiple simultaneous experiments. This allowed the assessment of experimental reproducibility, a rare occurrence among aerosol field tests. To transition to outdoor testing, the solid material provides some protection of the DNA label when exposed to ultraviolet (UV) radiation, with 60% of the DNA remaining intact after 60 minutes under a germicidal lamp and the rate of degradation declining with irradiation time. Additionally, exposure of the DNATrax material using formulations of two different food-grade sugars (maltodextrin and erythritol) to humidity as high as 66% had no significant effect on the DNA label’s degradation or the particle’s aerodynamic diameter, confirming particle stability under such conditions. In summary, confirmation of the DNATrax particles’ size and label integrity under variable conditions combined with experiment multiplexing and high resolution sampling provides a powerful experimental design for modeling aerosol transport through occupied indoor and outdoor locations.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2016

SCALE Code System

The SCALE Code System is a widely-used modeling and simulation suite for nuclear safety analysis and design that is developed, maintained, tested, and managed by the Reactor and Nuclear Systems Division (RNSD) of Oak Ridge National Laboratory (ORNL). SCALE provides a comprehensive, verified and validated, user-friendly tool set for criticality safety, reactor and lattice physics, radiation shielding, spent fuel and radioactive source term characterization, and sensitivity and uncertainty analysis. Since 1980, regulators, licensees, and research institutions around the world have used SCALE for safety analysis and design. SCALE provides an integrated framework with dozens of computational modules including three deterministic and three Monte Carlo radiation transport solvers that are selected based on the desired solution strategy. SCALE includes current nuclear data libraries and problem-dependent processing tools for continuous-energy (CE) and multigroup (MG) neutronics and coupled neutron-gamma calculations, as well as activation, depletion, and decay calculations. SCALE includes unique capabilities for automated variance reduction for shielding calculations, as well as sensitivity and uncertainty analysis. SCALE’s graphical user interfaces assist with accurate system modeling, visualization of nuclear data, and convenient access to desired results.SCALE 6.2 provides many new capabilities and significant improvements of existing features.New capabilities include:• ENDF/B-VII.1 nuclear data libraries CE and MG with enhanced group structures,• Neutron covariance data based on ENDF/B-VII.1 and supplemented with ORNL data,• Covariance data for fission product yields and decay constants,• Stochastic uncertainty and correlation quantification for any SCALE sequence with Sampler,• Parallel calculations with KENO,• Problem-dependent temperature corrections for CE calculations,• CE shielding and criticality accident alarm system analysis with MAVRIC,• CE depletion with TRITON (T5-DEPL/T6-DEPL),• CE sensitivity/uncertainty analysis with TSUNAMI-3D,• Simplified and efficient LWR lattice physics with Polaris,• Large scale detailed spent fuel characterization with ORIGAMI and ORIGAMI Automator,• Advanced fission source convergence acceleration capabilities with Sourcerer,• Nuclear data library generation with AMPX, and• Integrated user interface with Fulcrum.Enhanced capabilities include:• Accurate and efficient CE Monte Carlo methods for eigenvalue and fixed source calculations,• Improved MG resonance self-shielding methodologies and data,• Resonance self-shielding with modernized and efficient XSProc integrated into most sequences,• Accelerated calculations with TRITON/NEWT (generally 4x faster than SCALE 6.1),• Spent fuel characterization with 1470 new reactor-specific libraries for ORIGEN,• Modernization of ORIGEN (Chebyshev Rational Approximation Method [CRAM] solver, API for high-performance depletion, new keyword input format)• Extension of the maximum mixture number to values well beyond the previous limit of 2147 to ~2 billion,• Nuclear data formats enabling the use of more than 999 energy groups,• Updated standard composition library to provide more accurate use of natural abundances, andvi• Numerous other enhancements for improved usability and stability.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2018

Integrated Component-Specific Measurements to Develop Emission Factors for Compressors and Gas Gathering Lines

The Department of Energy (DOE) National Energy Technology Laboratory (NETL) and other critical stakeholders (e.g., U.S. Environmental Protection Agency (USEPA), state regulators, and industry) seek to more accurately characterize and quantify methane emissions from compressors and gathering lines within the Gathering and Boosting (G&B) segment of the natural gas system network. These efforts will reduce uncertainties in emissions estimates in USEPA’s current Green House Gas Inventory (GHGI) under their Greenhouse Gas Reporting Program (GHGRP). For this study, four G&B stations in the Gulf Coast Area were visited four times over the course of one year. Components were classified according to equipment and component type and emissions from select components were measured using multiple measurement technologies. The following major equipment types were considered: i) compressors, ii) separators, iii) dehydrators, iv) coalescers, v) slug catchers, vi) yard piping, vii) tanks, and viii) gathering lines. Individual components on each equipment type were further divided into the following main categories: i) connectors, ii) valves, iii) pressure relief valves (PRV), iv) meters, v) gauges, vi) regulators, vii) pneumatic device vents, and viii) compressor vents (e.g., distance piece vents [DPVs] and rod packing vents [RPVs]). Components were screened for the presence of leaks and venting emissions with an optical gas imaging (OGI) camera, and methane emission rates were measured using a microdilution high volume sampler. Downwind emissions were measured using an Open Path Fourier Transform Infrared (OP-FTIR) Spectrometer. Data were evaluated to assess correlation with site, equipment, time, and operational variables. Key findings and recommendations from the study include: A) Different counting methodologies can result in vastly different component counts, thus influencing total emissions calculations using Subpart W population EFs. A standard methodology should be adopted for classifying and counting components in order to minimize inconsistencies across studies and instruct stakeholders responsible for GHGI reporting. B) Results support current component subcategories listed in USEPA Subpart W, including the subdivision of connectors (flanged and “other”) and pneumatic device vents (intermittent, continuous low bleed, and continuous high bleed). C) Emission rates vary across sites and a relationship between higher emissions and compressor age was observed for some components. D) Component emission rates varied between field campaigns. The type of components contributing to overall measured emissions also varied among field campaigns. However, the emission rate variability was larger between different components (of the same category) than the variability in one component over the four field campaigns. Therefore, when calculating emission factors, it is not necessary to take repeat measurements to capture component variability. E) Measuring the emission rate of a sample for an extended period of time is not necessary to capture the average component emission rate. F) For reporting purposes, excluding liquid lines from site-wide screening may have a negligible effect on component-related emissions. However, from a site management perspective, malfunctioning components were not uncommon on liquid lines and contributed substantially to fugitive emissions at some sites. G) Population EFs, whether including or excluding liquid lines, were a conservative estimate of fugitive emissions. Leaker EFs consistently underestimated emissions at the field sites. Measurements at more sites, particularity sites in other areas of the country, need to be completed before determining the accuracy of EFs.

03 NATURAL GAS↗
DOE OSTI2018

Multi-Physics Signature Field Experiment at LLNL's Site 300 (S300 FE): Atmospheric Sampling Technical Report

As a part of LLNL’s Local Venture’s field experiment (S300 FE), Site 300 was instrumented to determine the feasibility and methodology needed to detect an “atmospheric chemical signature” from an explosion. This objective was just one within a broader field campaign to collect multi-physics data for the Local Venture. LLNL leveraged two planned, above-ground shot events at B851’s open firing table in April 2017 which served as the explosive devices. Gas samples were taken using continuous in situ sensors and timed bag collection sampling. In situ collection of carbon monoxide (at the ppm level) was done with a Miran SapphIRe portable ambient air analyzer. The bag samples were taken using a “blue box” gas sampler. Collected bags were analyzed for multiple species using LLNL laboratory facilities. Observations of wind speed, direction, air pressure, air temperature, and relative humidity were taken at the Site 300 52 m tall meteorological tower.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2018

sUAS Support for Mars 2020

The release of airborne radioactive material presents a health risk hazard to many individuals, including emergency responders and the public. Characterizing the unknown radioactive dangers produced during such incidents is essential. An advantage to utilizing unmanned aerial systems in this effort is the reduction of risk to personnel otherwise tasked with characterization, as well as more quickly obtaining data points during cloud immersion. By pairing a continuous air monitor (CAM) and detector with a drone, quantification of unknown material in the plume can be determined, and thus the extent of dose to people in the area can be estimated. A small, low-flow air sampler with a Geiger-Mueller counter may be characterized using experimental measurements of point sources representing a cloud of material and modelled using the Monte Carlo N-Particle Transport Code (MCNP), a software package for simulating nuclear processes, to find the usefulness and advantages of the system for detection of nuclides of concern in various nuclear plume scenarios. The flying time of small-unmanned aircraft systems (sUASs) is limited; thus, the minimum detectable activity and concentration rates possible must be determined over the system’s window of operation. While the activity concentrations corresponding to external and internal doses of concern can be detected with certainty for 90 Sr and 137 Cs in a plume, those corresponding to 238 Pu are more difficult to detect. The central principle of employing the sUAS is that the cockpit does not have an operator onboard the aircraft; therefore, control of the aircraft must take place by other means. With a sUAS, command and control can be separated into three distinct methods: (1) ground control or remote piloting, (2) semi-autonomous control, and (3) autonomous control. The isotope 238 Pu is used in the space industry, providing long-lived heat generation for fuel in outer space (NASA 2014). Due to the inherent risks with rocket launches to get objects into space, monitoring for release of these hazardous substances is a major undertaking. The current means is through employment of environmental CAMs. It is conceivable, however, that drone monitoring for release of 238 Pu in the event of a space launch mishap would give a greater chance of detection of plutonium particulate, because the particulate’s diffusion in air upon settling to the ground would make the probability of collecting a particle on a filter in a CAM very low.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

IAEA Coordinated Research Project on HTGR Physics, Thermal Hydraulics, and Depletion Uncertainty Analysis: PHISICS/RELAP5-3D Results for the Phase III Coupled Core Exercises

This report details the Parallel and Highly Innovative Simulation for INL Code System (PHISICS)/Reactor Excursions and Leak Analysis Program (RELAP5)-3D results obtained for the neutronics stand-alone core exercises defined for Phase II of the International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP) on high-temperature gas cooled reactor (HTGR) uncertainty analysis in modeling (UAM). The Phase II models and results are linked to the earlier Standardized Computer Analyses for Licensing Evaluation (SCALE)/Sampler/New ESC-based Weighting Transport (NEWT) data generated for the lattice physics (lattice) stage Phase I of the CRP. The focus of this report is the propagation of cross-section uncertainties from the lattice to the core calculation (core) phase of the prismatic 350 megawatt (MW) General Atomics (GA) Modular HTGR (MHTGR). The Phase II neutronics exercises consist of two sub-cases (II-2a and II-2b) specified for the fresh and mixed core loading of the MHTGR-350 prismatic high-temperature reactor design. Exercises II-2a and II-2b are performed without temperature feedback using isothermal temperature profiles. For the nominal lattice physics (i.e., best-estimate) model, a series of fresh and depleted single block and super cell SCALE/KENO and SCALE/NEWT models were developed at Idaho National Laboratory (INL) to assess the impact of spectral effects on the few-group cross sections. A combination of these lattice cells were used in a super cell calculation to construct 8-group cross-section libraries for use in the INL-developed PHISICS/RELAP5-3D code suite. The block-level homogenized libraries were assigned to the 220 fuel zones utilized in the MHTGR-350 PHISICS core model, according to the fresh and mixed fresh/depleted core loading pattern. In order to evaluate the effect of uncertainty propagation, several sets consisting of 1,000 perturbed NEWT lattice and PHISICS/RELAP5-3D core calculations were performed using the SCALE 6.2 ENDF/B-VII.1 cross-section co-variance data and the stochastic sampling module Sampler in SCALE 6.2.0. The Figures of Merit (FOM) investigated include the mean and standard deviation uncertainty values obtained for the eigenvalues, control rod worth values, block-averaged power profiles, power axial offsets, and isothermal temperature feedback coefficients. A sensitivity study on the main nuclide reaction contributors to the FOM uncertainties has not yet been performed since the SCALE/PHISICS/RAVEN coupling with ENDF/B-VII co-variance data is still being developed. It was found that uncertainties in the cross sections lead to uncertainties in the local power values of up to 1.5%. The variances between the fresh and mixed/depleted core models are not as large as the axial variations between the top and bottom within each core model. The core eigenvalue uncertainty is approximately 0.5%, which is significant in comparison to typical eigenvalue difference targets (less than ~0.1% for LWRs). The power peaking and axial offset uncertainties are smaller (0.10% – 0.15%) and probably below the level of significance in typical core design metrics. The use of an 8-group energy structure slightly under-estimates some of the uncertainties, but in absolute terms, the uncertainty results are reasonably close to the 26-group results. The 26- and 8-group mean and nominal results differ by a larger degree, and if it can be afforded in terms of calculation resources, the 26-group structure should be used for HTGR calculations. The 8-group structure can, however, still be useful for reasonably accurate uncertainty assessments of transient cases where computational time is important. It was furthermore found that the use of a super cell during the cross-section generation process for better representation of the spectral environment at the core periphery did not lead to significant changes in the core eigenvalues, axial offset, and power peaking uncertainties. The impact of the super cell cross sections on the nominal and mean eigenvalues is, however, important enough (0.6% – 1.1% lower) to justify the use of these super cells for HTGR core simulation to represent the softer spectral environment in the core peripheral regions for nominal best-estimate calculations. The use of super cells for HTGR uncertainty studies is, therefore, not strictly necessary, but the improvements in nominal results could justify their use.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Moore ME Feb 18 2019 aerosol class for Northern NM College [Slides)

This presentation was prepared for the request from Northern New Mexico College. "For your talk I’m thinking a bit more detail about the specific sampler you’re bringing, and the types in use at LANL, then setup and ops check, loading and removing a filter, cutting out the right size circle, reading the sample (we have a couple Eberline SAC-4s and Ludlum 2929s here), and radon/daughters corrections."

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2020

Measuring aerosol collection efficiency for the Bladewerx “New Speclon TM 5” and the older “Speclon TM 5” filter

Bladewerx TM LLC (Rio Rancho, NM) manufactures instrumentation, neutron shielding and activation foils for the radiation protection industry. Specializing in portable alpha/beta air monitors and sample counters, Bladewerx is the source of Speclon TM PTFE filter media that they recommend for high-resolution alpha spectroscopy. Los Alamos National Laboratory (LANL) utilizes Speclon TM filter material in CAM (Continuous Air Monitor) samplers for workplace air monitoring. The LANL Aerosol Engineering Facility received air filter material from Bladewerx, referred to as “New Speclon 5” in this document, in order to distinguish from filter material that was previously received (referred to as “Speclon 5” in this document). In this document, the aerosol collection efficiency and airflow resistance (pressure drop) were measured for the New Speclon 5 filter material.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2020

Use of Remote Sensing and In-Situ Observations to Develop and Evaluate Improved Representations of Convection and Clouds for the ACME Model

The overachieving goal of the whole CMDV-MCS project is to improve understanding of warm season continental convection and to develop treatments of convection and microphysics capable of representing mesoscale convective systems (MCSs) features in large-scale models. Our tasks for this project contributing to the overachieving goal include: (1) Improve the ice nucleation formulation for MG2 and P3 cloud microphysics schemes; (2) Improve the treatment of subgrid dynamics and thermodynamics driving the ice nucleation in E3SM; and (3) Test the performance of improved ice microphysics in E3SM with observation data. In this project, we have (1) Improved the ice nucleation parameterization for MG2 and P3 in E3SM by implementing two advanced empirical parameterizations with connection to aerosols. The two deterministic heterogeneous ice nucleation parameterizations (i.e., DeMott et al., 2015; Niemand et al., 2012) were merged with the MG2 and P3 cloud microphysics schemes in E3SM. Long-term simulations were conducted to examine the impacts of these new parameterizations on simulated cloud properties; (2) Improved the treatment of subgrid dynamics and thermodynamics driving the ice nucleation in E3SM. We evaluated the double Gaussian PDF of vertical velocity simulated by the Cloud Layers Unified By Binormals (CLUBB) and the sub-column vertical velocity sampled from the Subgrid Importance Latin Hypercube Sampler (SILHS) in E3SM. We introduced the vertical velocity variance induced by topographic gravity waves for ice nucleation and droplet activation; and (3) Tested the performance of improved ice microphysics in E3SM with observation data. We tested the new treatments of ice nucleation in the single column model (SCM) mode for the stratiform mixed-phase clouds observed during 9-10 October 2004 in the DOE ARM Mixed-Phase Arctic Cloud Experiment (M-PACE) and for the convective clouds observed on 20 May 2011 in the Midlatitude Continental Convective Clouds Experiment (MC3E). Modeled ice nucleating particles (INPs) concentrations were compared against observations collected around the globe.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2020

Oak Ridge National Laboratory Shutdown Dose Rate Code Suite

The Oak Ridge National Laboratory (ORNL) shutdown dose rate (SDDR) code suite calculates gamma dose rates at a location of interest due to activation of materials. The code suite is based on the rigorous two-step (R2S) method that involves two radiation transport calculations: (1) neutron radiation transport to determine neutron fluxes in the material for which activation calculations will be performed and (2) gamma radiation transport to determine dose rates at a location of interest, due to material activation. The ORNL SDDR code suite is used to determine an importance function that characterizes the neutron’s importance to the final SDDR when the Monte Carlo radiation transport method is used. This report demonstrates how the ORNL SDDR code suite is implemented through the following codes: Monte Carlo N-Particle (MCNP) code Version 5-1.60, Oak Ridge Isotope Generator (ORIGEN), MSX suite of utilities, Neutron Activation Gamma Source Sampler (NAGSS), and Automated Variance Reduction Generator (ADVANTG). For this demonstration, the International Thermonuclear Experimental Reactor (ITER) SDDR benchmark problem is used and the dose rate result is compared to previously published work.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2020

Advanced Mixing Condensation Particle Counter (aMCPC) Instrument Handbook

The U.S. Department of Energy Atmospheric Radiation Measurement (ARM) user facility’s aerosol counting, composition, extinction, and sizing system (ACCESS) includes a base module (9400), a filter sampler (9401), an advanced mixing condensation particle counter (aMCPC, 9403), a miniaturized optical practical counter (MOPC, 9405), and a single-channel tricolor absorption photometer (STAP, 9406). This handbook focuses on the aMCPC, which measures the ultra-fast total aerosol particle number concentration at 180 ms response rate with a detection diameter (D 50 = 7 nm).

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2020

Miniaturized Optical Particle Counter (MOPC) Instrument Handbook

The U.S. Department of Energy Atmospheric Radiation Measurement (ARM) aerosol counting, composition, extinction, and sizing system (ACCESS) includes a base module (9400), a filter sampler (9401), an advanced mixing condensation particle counter (MCPC, 9403), a miniaturized optical particle counter (mOPC, 9405), and a single-channel tricolor absorption photometer (STAP, 9406). This handbook focuses on the mOPC, which measures the aerosol particle size distribution between 0.19 and 3 μm.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2020

Single-Channel Tricolor Absorption Photometer (STAP) Instrument Handbook

The U.S. Department of Energy Atmospheric Radiation Measurement (ARM) user facility’s aerosol counting, composition, extinction, and sizing system (ACCESS) includes a base module (9400), a filter sampler (9401), an advanced mixing condensation particle counter (MCPC, 9403), a miniaturized optical particle counter (MOPC, 9405), and a single-channel tricolor absorption photometer (STAP, 9406). This handbook describes the principles and operations of the single-channel tricolor absorption photometer, Model 9406.

47 OTHER INSTRUMENTATION↗