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At least 1,099 records · Page 61

Depth of manual dismantling analysis: A cost–benefit approach

Highlights: ► A mathematical modeling tool for OEMs. ► The tool can be used by OEMs, recyclers of electr(on)ic equipment or WEEE management systems’ regulators. ► The tool makes use of cost–benefit analysis in order to determine the optimal depth of product disassembly. ► The reusable materials and the quantity of metals and plastics recycled can be quantified in an easy-to-comprehend manner. - Abstract: This paper presents a decision support tool for manufacturers and recyclers towards end-of-life strategies for waste electrical and electronic equipment. A mathematical formulation based on the cost benefit analysis concept is herein analytically described in order to determine the parts and/or components of an obsolete product that should be either non-destructively recovered for reuse or be recycled. The framework optimally determines the depth of disassembly for a given product, taking into account economic considerations. On this basis, it embeds all relevant cost elements to be included in the decision-making process, such as recovered materials and (depreciated) parts/components, labor costs, energy consumption, equipment depreciation, quality control and warehousing. This tool can be part of the strategic decision-making process in order to maximize profitability or minimize end-of-life management costs. A case study to demonstrate the models’ applicability is presented for a typical electronic product in terms of structure and material composition. Taking into account the market values of the pilot product’s components, the manual disassembly is proven profitable with the marginal revenues from recovered reusable materials to be estimated at 2.93–23.06 €, depending on the level of disassembly.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Thermal-Mechanical Cyclic Test of a Composite Cryogenic Tank for Reusable Launch Vehicles

This viewgraph presentation provides an overview of thermal-mechanical cyclic tests conducted on a composite cryogenic tank designed for reusable launch vehicles. Topics covered include: a structural analysis of the composite cryogenic tank, a description of Marshall Space Flight Center's Cryogenic Structure Test Facility, cyclic test plans and accomplishments, burst test and analysis and post-testing evaluation.

Messinger, Ross↗

Automated Particle Analysis of Hanford Tank Wastes 241-AN-106, 241-AN-101, and 241-AW-105

This effort is involved in developing particle size and density distribution (PSDD) for the major phases within Hanford tank waste sludge. The object of this work has been to use automated particle analysis (APA) methods on the Scanning Electron Microscope (SEM) to provide PSDDs. The PSDD can be used to quantify the proportion of gibbsite particles likely to settle quickly and enable blending of wastes that may be prone to pipeline plugging such as the uranium phase, clarkeite with other wastes. The Hanford tank waste solids were analyzed with SEM combined with x-ray Energy Dispersive Spectroscopy (EDS) using APA. This method can allow thousands of individual particles to be characterized and so may provide a more representative view of the samples and information PSDDs. We characterized several as-received sludge samples and developed PSDD with APA. Sample preparation methods were found to impact the collected results, as it was more difficult to collect representative samples of the larger particles. Much of the material was also dried and cemented together that further impacted results.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Composites Manufacturing Education and Technology Facility Expedites Manufacturing Innovation

The Composites Manufacturing Education and Technology facility (CoMET) at the National Wind Technology Center at the National Renewable Energy Laboratory (NREL) paves the way for innovative wind turbine components and accelerated manufacturing. Available for use by industry partners and university researchers, the 10,000-square-foot facility expands NREL's composite manufacturing research capabilities by enabling researchers to design, prototype, and test composite wind turbine blades and other components -- and then manufacture them onsite. Designed to work in conjunction with NREL's design, analysis, and structural testing capabilities, the CoMET facility expedites manufacturing innovation.

17 WIND ENERGY↗

Propagating neutronic uncertainties for FFTF LOFWOS Test #13

The safety evaluation conducted for licensing a Sodium-cooled Fast Reactor (SFR) may require modeling transients with best-estimate calculation tools that must first be validated against real-world measurements. To provide the community with a valuable benchmarking opportunity for validating SFR analysis tools and methods, the International Atomic Energy Agency (IAEA) initiated a coordinated research project (CRP) in 2018 for the analysis of the Fast Flux Test Facility (FFTF) Loss of Flow Without Scram (LOFWOS) Test #13. The impact of nuclear data uncertainties on neutronics parameters was previously investigated based on the COMMARA-2.0 covariance matrix. Since the transient simulation results are very sensitive to certain reactivity coefficients, it was decided to employ rigorous uncertainty propagation methods to quantify the impact of nuclear data uncertainties on the best-estimate predication of FFTF LOFWOS Test #13. The DAKOTA code is used to propagate neutronic uncertainties through SAS4A/SASSYS-1 transient simulations, while taking into account spatial and reaction-wise correlations within these uncertainties. This study shows that the remaining discrepancies observed between the Argonne National Laboratory (ANL) best-estimate results and the experimental measurements can be partly explained by the uncertainty associated with Gas Expansion Module (GEM) worth, which contributes the majority of the overall nuclear data uncertainty on the output from the FFTF LOFWOS Test #13 transient simulation. This study also confirmed the importance of including spatial and reaction-wise correlations of nuclear data uncertainties on feedback coefficients in the uncertainty propagation to avoid under-estimating their impact during the transient simulations.

42 ENGINEERING↗

A Method for the Assessment of Operational Safety of DGR Using Scenario Development - 18187

This paper treats the subject of the accident analysis for the operational phase of a deep geological repository (DGR) for radioactive waste. It describes the further development of the procedures used in the accident analysis. The procedures of the scenario development for the post-closure phase of DGRs are used as example for this further development. Nine international projects regarding scenario development were analyzed to create an advanced and more systematic method for the accident analysis. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Characterization of Tank 16 Annulus Samples

The closure of Tank 16H will require removal of material from the annulus of the tank. Samples from Tank 16H annulus were characterized and tested to provide information to evaluate various alternatives for removing the annulus waste. The analysis found all four annulus samples to be composed mainly of Si, Na, and Al and lesser amounts of other elements. The XRD data indicate quartz (SiO 2 ) and sodium aluminum nitrate silicate hydrate (Na 8 (Al 6 Si 6 O 24 )(NO 3 ) 2 .4H 2 O) as the predominant crystalline mineral phases in the samples. The XRD data also indicate the presence of crystalline sodium nitrate, sodium nitrite, gibbsite, hydrated sodium bicarbonate, and muscovite. Based on the weight of solids remaining at the end of the test, the water leaching test results indicate approximately 20-35% of the solids dissolved after three contacts with an approximately 3:1 volume of water at 45 C. The chemical analysis of the leachates and the XRD results of the remaining solids indicate sodium salts of nitrate, nitrite, sulfate, and possibly carbonate/bicarbonate make up the majority of the dissolved material. The majority of these salts were dissolved in the first water contact and simply diluted with each subsequent water contact. The water leaching removed large amounts of the uranium in two of the samples and {approx}1/3 of the {sup 99}Tc from all four samples. Most of the other radionuclides analyzed showed low solubility in the water leaching test. The preliminary data on the oxalic acid leaching test indicate the three acid contacts at 45 C dissolved from {approx}34-47% of the solids. The somewhat higher dissolution found in the oxalic acid leaching test versus the water leaching test might be offset by the tendency of the oxalic acid solutions to take on a gel-like consistency. The filtered solids left behind after three oxalic acid contacts were sticky and formed large clumps after drying. These two observations could indicate potential processing difficulties with solutions and solids from oxalic acid leaching. The gel formation might be avoided by using larger volumes of the acid. Further testing would be recommended before using oxalic acid to dissolve the Tank 16H annulus waste to ensure no processing difficulties are encountered in the full scale process.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Mixing with applications to inertial-confinement-fusion implosions

Approximate one-dimensional (1D) as well as 2D and 3D simulations are playing an important supporting role in the design and analysis of future experiments at National Ignition Facility. This work is mainly concerned with 1D simulations, used extensively in design and optimization. We couple a 1D buoyancy-drag mix model for the mixing zone edges with a 1D inertial confinement fusion simulation code. This analysis predicts that National Ignition Campaign (NIC) designs are located close to a performance cliff, so modeling errors, design features (fill tube and tent) and additional, unmodeled instabilities could lead to significant levels of mix. The performance cliff we identify is associated with multimode plastic ablator (CH) mix into the hot-spot deuterium and tritium (DT). The buoyancy-drag mix model is mode number independent and selects implicitly a range of maximum growth modes. Our main conclusion is that single effect instabilities are predicted not to lead to hot-spot mix, while combined mode mixing effects are predicted to affect hot-spot thermodynamics and possibly hot-spot mix. Combined with the stagnation Rayleigh-Taylor instability, we find the potential for mix effects in combination with the ice-to-gas DT boundary, numerical effects of Eulerian species CH concentration diffusion, and ablation-driven instabilities. With the help of a convenient package of plasma transport parameters developed here, we give an approximate determination of these quantities in the regime relevant to the NIC experiments, while ruling out a variety of mix possibilities. Plasma transport parameters affect the 1D buoyancy-drag mix model primarily through its phenomenological drag coefficient as well as the 1D hydro model to which the buoyancy-drag equation is coupled.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Influence of Cracks in Cementitious Engineered Barriers in a Near-Surface Disposal System: Assessment Analysis of the Belgian Case

In large cement-based structures such as a near surface disposal facility for radioactive waste voids and cracks are inevitable. However, the pattern and nature of cracks are very difficult to predict reliably. Cracks facilitate preferential water flow through the facility because their saturated hydraulic conductivity is generally higher than the conductivity of the cementitious matrix. Moreover, sorption within the crack is expected to be lower than in the matrix and hence cracks in engineered barriers can act as a bypass for radionuclides. Consequently, understanding the effects of crack characteristics on contaminant fluxes from the facility is of utmost importance in a safety assessment. In this paper we numerically studied radionuclide leaching from a crack-containing cementitious containment system. First, the effect of cracks on radionuclide fluxes is assessed for a single repository component which contains a radionuclide source (i.e. conditioned radwaste). These analyses reveal the influence of cracks on radionuclide release from the source. The second set of calculations deals with the safety assessment results for the planned near-surface disposal facility for low-level radioactive waste in Dessel (Belgium); our focus is on the analysis of total system behaviour in regards to release of radionuclide fluxes from the facility. Simulation results are interpreted through a complementary safety indicator (radiotoxicity flux). We discuss the possible consequences from different scenarios of cracks and voids. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Discovery: Under the Microscope at Kennedy Space Center

The National Aeronautics & Space Administration (NASA) is known for discovery, exploration, and advancement of knowledge. Since the days of Leeuwenhoek, microscopy has been at the forefront of discovery and knowledge. No truer is that statement than today at Kennedy Space Center (KSC), where microscopy plays a major role in contamination identification and is an integral part of failure analysis. Space exploration involves flight hardware undergoing rigorous "visually clean" inspections at every step of processing. The unknown contaminants that are discovered on these inspections can directly impact the mission by decreasing performance of sensors and scientific detectors on spacecraft and satellites, acting as micrometeorites, damaging critical sealing surfaces, and causing hazards to the crew of manned missions. This talk will discuss how microscopy has played a major role in all aspects of space port operations at KSC. Case studies will highlight years of analysis at the Materials Science Division including facility and payload contamination for the Navigation Signal Timing and Ranging Global Positioning Satellites (NA VST AR GPS) missions, quality control monitoring of monomethyl hydrazine fuel procurement for launch vehicle operations, Shuttle Solids Rocket Booster (SRB) foam processing failure analysis, and Space Shuttle Main Engine Cut-off (ECO) flight sensor anomaly analysis. What I hope to share with my fellow microscopists is some of the excitement of microscopy and how its discoveries has led to hardware processing, that has helped enable the successful launch of vehicles and space flight missions here at Kennedy Space Center.

Howard, Philip M.↗

Evaluation of Low-Level Waste Disposal Receipt Data for Los Alamos National Laboratory Technical Area 54, Area G Disposal Facility - Fiscal Year 2011

The Los Alamos National Laboratory (LANL or the Laboratory) generates radioactive waste as a result of various activities. Operational or institutional waste is generated from a wide variety of research and development activities including nuclear weapons development, energy production, and medical research. Environmental restoration (ER), and decontamination and decommissioning (D and D) waste is generated as contaminated sites and facilities at LANL undergo cleanup or remediation. The majority of this waste is low-level radioactive waste (LLW) and is disposed of at the Technical Area 54 (TA-54), Area G disposal facility. U.S. Department of Energy (DOE) Order 435.1 (DOE, 2001) requires that radioactive waste be managed in a manner that protects public health and safety, and the environment. To comply with this order, DOE field sites must prepare and maintain site-specific radiological performance assessments for LLW disposal facilities that accept waste after September 26, 1988. Furthermore, sites are required to conduct composite analyses that account for the cumulative impacts of all waste that has been (or will be) disposed of at the facilities and other sources of radioactive material that may interact with the facilities. Revision 4 of the Area G performance assessment and composite analysis was issued in 2008 (LANL, 2008). These analyses estimate rates of radionuclide release from the waste disposed of at the facility, simulate the movement of radionuclides through the environment, and project potential radiation doses to humans for several on-site and off-site exposure scenarios. The assessments are based on existing site and disposal facility data and on assumptions about future rates and methods of waste disposal. The accuracy of the performance assessment and composite analysis depends upon the validity of the data used and assumptions made in conducting the analyses. If changes in these data and assumptions are significant, they may invalidate or call into question certain aspects of the analyses. For example, if the volumes and activities of waste disposed of during the remainder of the disposal facility's lifetime differ significantly from those projected, the doses projected by the analyses may no longer apply. DOE field sites are required to implement a performance assessment and composite analysis maintenance program. The purpose of this program is to ensure the continued applicability of the analyses through incremental improvement of the level of understanding of the disposal site and facility. Site personnel are required to conduct field and experimental work to reduce the uncertainty in the data and models used in the assessments. Furthermore, they are required to conduct periodic reviews of waste receipts, comparing them to projected waste disposal rates. The radiological inventory for Area G was updated in conjunction with Revision 4 of the performance assessment and composite analysis (Shuman, 2008). That effort used disposal records and other sources of information to estimate the quantities of radioactive waste that have been disposed of at Area G from 1959, the year the facility started receiving waste on a routine basis, through 2007. It also estimated the quantities of LLW that will require disposal from 2008 through 2044, the year in which it is assumed that disposal operations at Area G will cease. This report documents the fourth review of Area G disposal receipts since the inventory was updated and examines information for waste placed in the ground during fiscal years (FY) 2008 through 2011. The primary objective of the disposal receipt review is to ensure that the future waste inventory projections developed for the performance assessment and composite analysis are consistent with the actual types and quantities of waste being disposed of at Area G. Toward this end, the disposal data that are the subject of this review are used to update the future waste inventory projections for the disposal facility. These projections are compared to the future inventory projections that were developed for Revision 4 of the performance assessment and composite analysis. The approach used to characterize the FY 2008 through 2011 waste is generally the same as that used to characterize the inventory for the Revision 4 analyses (Shuman, 2008). This methodology is described in Section 2. The results of the disposal receipt review are presented in Section 3 and discussed in terms of their significance to the Area G analyses.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Heavy ion linear accelerator for radiation damage studies of materials

A new eXtreme MATerial (XMAT) research facility is being proposed at Argonne National Laboratory to enable rapid in situ mesoscale bulk analysis of ion radiation damage in advanced materials and nuclear fuels. This facility combines a new heavy-ion accelerator with the existing high-energy X-ray analysis capability of the Argonne Advanced Photon Source. The heavy-ion accelerator and target complex will enable experimenters to emulate the environment of a nuclear reactor making possible the study of fission fragment damage in materials. Material scientists will be able to use the measured material parameters to validate computer simulation codes and extrapolate the response of the material in a nuclear reactor environment. Utilizing a new heavy-ion accelerator will provide the appropriate energies and intensities to study these effects with beam intensities which allow experiments to run over hours or days instead of years. The XMAT facility will use a CW heavy-ion accelerator capable of providing beams of any stable isotope with adjustable energy up to 1.2 MeV/u for 238 U 50+ and 1.7 MeV for protons. This energy is crucial to the design since it well mimics fission fragments that provide the major portion of the damage in nuclear fuels. The energy also allows damage to be created far from the surface of the material allowing bulk radiation damage effects to be investigated. Furthermore, the XMAT ion linac includes an electron cyclotron resonance ion source, a normal-conducting radio-frequency quadrupole and four normal-conducting multi-gap quarter-wave resonators operating at 60.625 MHz. This paper presents the 3D multi-physics design and analysis of the accelerating structures and beam dynamics studies of the linac.

47 OTHER INSTRUMENTATION↗

Toward Unified Autonomous Scattering Experiments: A Cross-Facility Case Study at ALS and PETRA III

Autonomous experiments rely on the integration of control, data acquisition, analysis, and decision-making frameworks. While such systems have been demonstrated at individual facilities, adapting them to additional instruments remains challenging due to differences in local infrastructure. We present a modular workflow that connects existing open-source tools for data access (Tiled), workflow orchestration (Prefect), analysis and visualization (pyFAI, Plotly Dash), and Gaussian-process-based adaptive sampling (gpCAM) into a unified framework for autonomous scattering experiments. The same configuration operates across two synchrotron beamlines (ALS 7.3.3 and PETRA III P03) with only minimal facility-specific adjustments, as shown in proof-of-concept demonstrations. This validates that a consistent design emphasizing modularity and shared interfaces can ease deployment across diverse experimental environments. The resulting framework provides a flexible foundation for extending autonomous control and analysis capabilities beyond a single beamline or instrument.

47 OTHER INSTRUMENTATION↗

A quarantine protocol for analysis of returned extraterrestrial samples

A protocol is presented for the analysis at an earth-orbiting quarantine facility of return samples of extraterrestrial material that might contain (nonterrestrial) life forms. The protocol consists of a series of tests designed to determine whether the sample, conceptualized as a 1-kg sample of Martian soil, is free from nonterrestrial biologically active agents and so may safely be sent to a terrestrial containment facility, or it exhibits biological activity requiring further (second-order) testing outside the biosphere. The first-order testing procedure seeks to detect the presence of any replicating organisms or toxic substances through a series of experiments including gas sampling, analysis of radioactivity, stereomicroscopic inspection, chemical analysis, microscopic examination, the search for metabolic products under growth conditions, microbiologicl assays, and the challenge of cultured cells with any agents found or with the extraterrestrial material as is. Detailed plans for the second-order testing would be developed in response to the actual data received from primary testing.

Bagby, J. R.↗

Integral Reactor Containment Condensation Model and Experimental Validation

This NEUP funded project, NEUP 12-3630, is for experimental, numerical and analytical studies on high-pressure steam condensation phenomena in a steel containment vessel connected to a water cooling tank, carried out at Oregon State University (OrSU) and the University of Wisconsin at Madison (UW-Madison). In the three years of investigation duration, following the original proposal, the planned tasks have been completed: (1) Performed a scaling study for the full pressure test facility applicable to the reference design for the condensation heat transfer process during design basis accidents (DBAs), modified the existing test facility to route the steady-state secondary steam flow into the high pressure containment for controllable condensation tests, and extended the operations at negative gage pressure conditions (OrSU). (2) Conducted a series of DBA and quasi-steady experiments using the full pressure test facility to provide a reliable high pressure condensation database (OrSU). (3) Analyzed experimental data and evaluated condensation model for the experimental conditions, and predicted the prototypic containment performance under accidental conditions (UW-Madison). A film flow model was developed for the scaling analysis, and the results suggest that the 1/3 scaled test facility covers large portion of laminar film flow, leading to a lower average heat transfer coefficient comparing to the prototypic value. Although it is conservative in reactor safety analysis, the significant reduction of heat transfer coefficient (50%) could under estimate the prototypic condensation heat transfer rate, resulting in inaccurate prediction of the decay heat removal capability. Further investigation is thus needed to quantify the scaling distortion for safety analysis code validation. Experimental investigations were performed in the existing MASLWR test facility at OrST with minor modifications. A total of 13 containment condensation tests were conducted for pressure ranging from 4 to 21 bar with three different static inventories of non-condensable gas. Condensation and heat transfer rates were evaluated employing several methods, notably from measured temperature gradients in the HTP as well as measured condensate formation rates. A detailed mass and energy accounting was used to assess the various measurement methods and to support simplifying assumptions required for the analysis. Condensation heat fluxes and heat transfer coefficients are calculated and presented as a function of pressure to satisfy the objectives of this investigation. The major conclusions for those tests are summarized below: (1) In the steam blow-down tests, the initial condensation heat transfer process involves the heating-up of the containment heat transfer plate. An inverse heat conduction model was developed to capture the rapid transient transfer characteristics, and the analysis method is applicable to SMR safety analysis. (2) The average condensation heat transfer coefficients for different pressure conditions and non-condensable gas mass fractions were obtained from the integral test facility, through the measurements of the heat conduction rate across the containment heat transfer plate, and from the water condensation rates measurement based on the total energy balance equation. 15 (3) The test results using the measured HTP wall temperatures are considerably lower than popular condensation models would predict mainly due to the side wall conduction effects in the existing MASLWR integral test facility. The data revealed the detailed heat transfer characteristics of the model containment, important to the SMR safety analysis and the validation of associated evaluation model. However this approach, unlike separate effect tests, cannot isolate the condensation heat transfer coefficient over the containment wall, and therefore is not suitable for the assessment of the condensation heat transfer coefficient against system pressure and noncondensable gas mass fraction. (4) The average condensation heat transfer coefficients measured from the water condensation rates through energy balance analysis are appropriate, however, with considerable uncertainties due to the heat loss and temperature distribution on the containment wall. With the consideration of the side wall conduction effects, the results indicate that the measured heat transfer coefficients in the tests is about 20% lower than the prediction of Dehbi’s correlation, mainly due to the side wall conduction effects. The investigation also indicates an increase in the condensation heat transfer coefficient at high containment pressure conditions, but the uncertainties invoked with this method appear to be substantial. (5) Non-condensable gas in the tests has little effects on the condensation heat transfer at high elevation measurement ports. It does affect the bottom measurements near the water level position. The results suggest that the heavier non-condensable gas is accumulated in the lower portion of the containment due to stratification in the narrow containment space. The overall effects of the non-condensable gas on the heat transfer process should thus be negligible for tall containments of narrow condensation spaces in most SMR designs. Therefore, the previous correlations with noncondensable gas effects are not appropriate to those small SMR containments due to the very poor mixing of steam and non-condensable gas. The MELCOR simulation results agree with the experimental data reasonably well. However, it is observed that the MELCOR overpredicts the heat flux for all analyzed tests. The MELCOR predicts that the heat fluxes for CCT’s approximately range from 30 to 45 kW/m2 whereas the experimental data (averaged) ranges from about 25 to 40 kW/m2. This may be due to the limited availability of liquid film models included in MELCOR. Also, it is believed that due to complex test geometry, measured temperature gradients across the heat transfer plate may have been underestimated and thus the heat flux had been underestimated. The MELCOR model predicts a film thickness on the order of 100 microns, which agrees very well with film flow model developed in this study for scaling analysis. However, the expected differences in film thicknesses for near vacuum and near atmospheric test conditions are not significant. Further study on the behavior of condensate film is expected to refine the simulation results. Possible refinements include but are not limited to, the followings: CFD simulation focusing on the liquid film behavior and benchmarking with experimental analyses for simpler geometries. 16 1 INTRODUCTION This NEUP funded project, NEUP 12-3630, is for experimental, numerical and analytical studies on high-pressure steam condensation phenomena in a steel containment vessel connected to a water cooling tank, carried out at Oregon State University (OrSU) and the University of Wisconsin at Madison (UW-Madison). The experimental results are employed to validate the containment condensation model in reactor containment system safety analysis code for integral SMRs. Such a containment condensation model is important to demonstrate the adequate cooling. In the three years of investigation, following the original proposal, the following planned tasks have been completed: (1) Performed a scaling study for the full pressure test facility applicable to the reference design for the condensation heat transfer process during design basis accidents (DBAs), modified the existing test facility to route the steady-state secondary steam flow into the high pressure containment for controllable condensation tests, and extended the operations at negative gage pressure conditions (OrSU). (2) Conducted a series of DBA and quasi-steady experiments using the full pressure test facility to provide a reliable high pressure condensation database (OrSU). (3) Analyzed experimental data and evaluated condensation model for the experimental conditions, and predicted the prototypic containment performance under accidental conditions (UW-Madison). The results are applicable to integral Small Modular Reactor (SMR) designs, including NuScale, mPower, Westinghouse SMR, Holtec-160 and other integral reactors with small containments of relatively high pressures under accidental conditions. Testing has been conducted at the OrSU laboratory in the existing MASLWR (Multi-Application Small Light Water Reactor) integral test facility sponsored by the US Department of Energy. Its highpressure stainless steel containment model (~2 MPa) is scaled to the NuScale SMR currently under development at NuScale Power, Inc.. Minor modifications to the model containment have been made to control the non-condensable gas fraction and to utilize the secondary loop stable steam flow for condensation testing. UW-Madison has developed a containment condensation model, which leveraged previous validated containment heat transfer work carried out at UW-Madison, and extended the range of applicability of the model to integral SMR designs that utilize containment vessels of high heat transfer efficiencies. In this final report, the research background and literature survey are presented in Chapter 2 and 3, respectively. The test facility description and modifications are summarized in Chapter 4, and the scaling analysis is introduced in Chapter 5. The tests description, procedures, and data analysis are presented in Chapter 6, while the numerical modeling is presented in Chapter 7, followed by a conclusion section in Chapter 8.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Evaluation of Air Quality Impacts from Addition of Boiler 5 at Naval Reactors Facility

This report documents an air-quality impact analysis of criteria air-pollutant emissions from the addition of Boiler 5 at NRF to support the permit application. Maximum increases in criteria air-pollutant concentrations resulting from Boiler 5 emissions were modeled and compared to significant contribution levels, as defined in IDAPA 58.01.01. Concentrations were calculated at each of 1352 potential public receptor locations on and around the INL Site, and reported at 1) receptor locations where the maximum concentrations occurred as a result of operating only Boiler 5, and 2) receptor locations where the maximum concentrations occurred as a result of baseline actual emissions (BAEs) from all significant INL sources considered in the INL Application for a Synthetic Minor Sitewide Air Quality Permit to Construct with Facility Emission Cap Component (DOE-ID 2017). Increases in criteria air pollutant concentrations resulting from operation of NRF Boiler 5 are considerably less than significant contribution levels at all INL public receptor locations.

99 GENERAL AND MISCELLANEOUS↗

Ultrasmall Zeolite L Crystals Prepared from Highly Interdispersed Alkali‐Silicate Precursors

Abstract The preparation of nanosized zeolites is critical for applications where mass‐transport limitations within microporous networks hinder their performance. Often the ability to generate ultrasmall zeolite crystals is dependent upon the use of expensive organics with limited commercial relevance. Herein, we report the generation of zeolite L crystals with uniform sizes less than 30 nm using a facile, organic‐free method. Time‐resolved analysis of precursor assembly and evolution during nonclassical crystallization highlights key differences among silicon sources. Our findings reveal that a homogenous dispersion of potassium ions throughout silicate precursors leads to the formation of a metastable nonporous phase, which undergoes an intercrystalline transformation to zeolite L. The generation of highly interdispersed alkali‐silicate precursors is seemingly critical to enhancing the rate of nucleation and facilitating the formation of ultrasmall crystal.

Li, Rui↗