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At least 703 records · Page 39

Assessment of the 3420 Building Filtered Exhaust Stack Sampling Probe Location: Stack Verification Following Fan and Air Blender Additions

The Pacific Northwest National Laboratory (PNNL) 3420 Building, which is part of the Physical Sciences Facility (PSF), houses radiological capabilities that results in the requirement that emissions monitoring must be conducted for potential radionuclides in the exhaust air discharge of this building. The air monitoring system is required to conform to Title 40 of the Code of Federal Regulations part 61 (40 CFR 61) Subpart H, which in turns requires a sampling probe in the exhaust stream to conform to the criteria of American National Standards Institute / Health Physics Society (ANSI/HPS) N13.1-2011, Sampling and Monitoring Releases of Airborne Radioactive Substances from the Stack and Ducts of Nuclear Facilities. To support the air emissions permit for the 3420 Building on the PNNL campus, stack testing that used computational fluid dynamics (CFD) modeling as a surrogate stack and verification tests of velocity uniformity and flow angle on the retrofitted facility stack was performed. The ANSI/HPS N13.1-2011 criteria for the air monitoring probe location are that velocity uniformity, gaseous tracer uniformity, and particulate tracer uniformity must be less than or equal to 20%COV. Furthermore, no point in the sampling location may have a gaseous tracer concentration that varies from the mean concentration by more than 30%. Additionally, the flow angle at the sampling location must not be more than 20°. The CFD modeling of the stack, as reported by Recknagle et al. (2018) demonstrated that the stack meets the criteria at the probe location. The velocity uniformity and flow angle results from the 3420 stack verification tests, performed in October 2020, demonstrated that the CFD model results may be used to support the qualification of the stack sampling location. The measured velocity uniformity verification test result was 1.4%COV. This value is well within the uniformity criterion, which is that the velocity uniformity be =20%COV. Additionally, this value is well within the criterion that the actual stack measurement must be within 5% of the surrogate stack result of 2.1%COV when all four fans were operating. Additionally, the measured average flow angle at the 3420 stack monitor location was 15.5 degrees. Although this is higher than expected based on the CFD model, the result is =20 degrees, so the criterion is met. Based on these stack verification test results, the reconfigured 3420 Building filtered exhaust stack meets the qualification criteria given in the ANSI/HPS N13.1-2011 standard. Further changes to the system configuration or operating conditions that are outside the bounds described in this and the CFD report (Recknagle et al., 2018) may require additional tests and additional analysis to determine compliance with the standard.

3420 Building↗

NNSA-IAEC Science Area V Environmental ISR: Waste Management and Subsurface Science (Final Report FY18-FY20)

In August 2017, the U.S. National Nuclear Security Administration (NNSA) and the Israel Atomic Energy Commission (IAEC) launched a joint waste management program to evaluate the feasibility of siting a radioactive waste repository at intermediate depths in Israel. The bilateral collaboration involves three National Laboratories (LLNL, LANL, and Sandia), and the Nuclear Research Center, Negev, Geological Survey of Israel, and Ben Gurion University in Israel. This research is designed to assist with the evaluation of locations for an intermediate nuclear waste repository in Israel. This final report is a high-level summary of the results from the past three years by the team that worked on the WM-2: Radionuclide colloid-facilitated transport in fractured carbonate rock project for Science Area V, Environmental ISR: Subsurface Science and Waste Management. We have completed all of our tasks and met all of our milestones. MW-2 had three integrated tasks. Task 1 was laboratory experiments investigating colloid transport of radionuclides in natural fractured carbonate rocks from the Avdat formation, Israel. Task 2 was field experiments injecting radionuclide analogues pre-sorbed to clay colloids into fractured carbonate rock, and task 3 was numerical modeling of laboratory and field experimental results to assess the importance of colloids in the migration of relevant radionuclides. This task was incorporated into tasks 1 and 2. A 4 th task was initiated in FY20, to investigate the role of organics in facilitating radionuclide transport under the same conditions explored in task 1 due to the high organic content of the rocks being investigated for the immediate borehole locations. Preliminary results will be summarized here and the work will continue in FY21-23. The overall objective of this research is to evaluate the role of colloids (naturally occurring < 1 micron particles) in facilitating the transport of long-lived radionuclides from a nuclear waste repository situated in fractured carbonate rocks. Currently little data exists on radionuclide transport in carbonate rocks, but this is the main rock type available for siting a repository in Israel. Laboratory experiments were carried out in both Israel and the U.S., field experiments have taken place in Israel, and reactive transport modeling involved LLNL, LANL and Israel partner institutions. LLNL hosted a graduate student, Emily Tran, from Ben Gurion University in FY18 and FY19 and much of the work presented here was part of her PhD research.

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Complete Round-Robin Hydrogen Gas Analysis Capability Comparison (Milestone 2.6)

The Department of Energy (DOE) is currently evaluating strategies for the extended dry storage of aluminum-clad spent nuclear fuel (ASNF). Part of this assessment concerns the extent of radiolytic molecular hydrogen (H2) generation from the aluminum cladding’s oxyhydroxide corrosion layers. Understanding this radiation-induced process and the factors effecting it (e.g., system conditions such as temperature and gaseous environment) are essential for the development of predictive computer models to support the Technical Considerations and Challenges for Extended (> 50 yrs) Dry Storage of ASNF program. To achieve this goal and ensure that the experimental data gathered by Task 2 (Oxyhydroxide Layer Radiolytic Gas Generation Resolution) research groups (Idaho National Laboratory and Savannah River National Laboratory) are consistent, a round-robin H2 analysis capability comparison was initiated. Here we present the results from said round-robin and conclude that despite differences in sample preparation, irradiation parameters, and analytical procedures, the measured data are sufficiently consistent between the two laboratories (= 15%).

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Small to Full Height Scale Comparisons of Cesium Ion Exchange Performance with Crystalline Silicotitanate

The U.S. Department of Energy’s (DOE) Hanford Site houses 56 million gallons of high-level radioactive waste generated from plutonium production from 1944 to 1988. The supernatant waste, currently stored in underground tanks, is intended to be vitrified following filtration and 137 Cs removal at the Hanford Waste Treatment and Immobilization Plant (WTP) Low-Activity Waste (LAW) Vitrification Facility. The WTP Pretreatment Facility will not be operational for several years. The Tank Side Cesium Removal (TSCR) system is a technology demonstration that will remove cesium from tank waste supernate to support directly feeding LAW to the vitrification facility. The 137 Cs removal is important to meet the WTP LAW contract specification and ultimately for creating a contact-handled waste form. The waste acceptance criteria (WAC) limit for the WTP LAW Facility is <3.18E-5 Ci 137 Cs per mole Na. The TSCR system is skid mounted and employs two key technologies: dead end filtration for solids removal and ion exchange (IX) for cesium removal. Filtration is necessary to protect the functionality of the IX columns. The IX process utilizes sodium form crystalline silicotitanate (CST) IX media produced by UOP Honeywell, LLC (Des Plaines, IL) under the product name IONSIV R9140-B, 18 x 50 mesh, in a lead-lag-polish column configuration. Each column contains a CST bed height of 92 inches and a volume of approximately 157 gallons of CST IX media. The full-scale TSCR operation will run at a residence time of 1.9 bed volumes (BV)/h, which results in a superficial velocity of 7.3 cm/min and a flowrate of ~5 gal/min. Column testing at small (2.5% of the full bed height), medium (12% of the full bed height), and full-height scales has been previously conducted to evaluate process variables and scale up performance of Cs exchange onto the CST. Cesium load performances from various sieve cuts at the small scale indicated that a 30-mesh sieve cut be tested to determine if it better reflects the 12% and full-height column performances at the small scale. Two process flowrates were tested in the small-column configuration with <30-mesh CST and simulant solutions. These tests were compared to full-height column tests at the same residence times to assess CST particle size effect on column scaling. Table S.1 summarizes the observed column performance determined for the two flowrates juxtaposed to the previous work with 5.6 M Na simulant at higher scales (used as benchmarks). The WAC breakthroughs between the small and full-height tests at 1.3 BV/h differed by ~59 BVs. The WAC breakthroughs for small, medium, and full-height tests at 1.8 BV/h were consistent at 240 BVs. The 50% Cs breakthroughs were nominally equivalent for all column tests. The common transition zones and onsets of Cs breakthrough at the 1.8 BV/h tests indicated that the Cs mass transfers were equivalent and thus the <30-mesh CST Cs load performance at the small scale successfully modeled that of the full-height system. It is recommended that the <30-mesh CST be used in subsequent 10-mL CST bed tests.

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Ion Exchange Processing of AP-105 Hanford Tank Waste through Crystalline Silicotitanate in a Staged 2- then 3-Column System

The U.S. Department of Energy (DOE) is working to expedite processing of Hanford tank waste supernate at the Hanford Waste Treatment and Immobilization Plant (WTP). To support this goal, Washington River Protection Solutions, LLC (WRPS, Richland, WA) is designing a system for suspended solids and cesium (Cs/ 137 Cs) removal from Hanford tank waste supernate. The effluent will then be sent to the WTP Low-Activity Waste (LAW) Facility for vitrification. The Cs removal is critical for eliminating the high dose rate associated with 137 Cs and facilitating a contact maintenance philosophy for the LAW Facility. The maximum 137 Cs concentration in the LAW sent to the WTP is targeted to be below the 3.18E-5 Ci 137 Cs/mole of Na waste acceptance criteria (WAC) limit. The filtration and ion exchange systems will be placed near the Hanford tanks and are collectively termed the Tank Side Cesium Removal (TSCR) system.

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Crystalline Silicotitanate Batch Contact Testing with Ba, Ca, Pb, and Sr

Washington River Protection Solutions is working to support initial production of immobilized low-activity waste (LAW) by feeding Hanford tank supernate from tank farms to the Hanford Waste Treatment and Immobilization Plant (WTP) LAW Facility. This goal incorporates the design of a Tank Side Cesium Removal system, which filters tank waste supernate to remove suspended solids and then removes Cs by processing it through crystalline silicotitanate (CST) ion exchange media manufactured by Honeywell UOP, LLC. The 137 Cs-depleted product is intended to be sent to the WTP for vitrification. Processing of actual tank waste supernate showed effectively complete uptake of Sr and Ba by CST and significant uptake of Ca. Further, Campbell et al. (2019) analyzed CST post-column testing and found significant (>1E-2 mmoles/g) uptake of Ca and Pb along with some Ba, Cd, Fe, Sr, and U. This led to concern that selected metals, particularly the +2 cations, Ca, Sr, Ba, and Pb may be consuming Cs exchange sites and possibly reducing CST capacity for Cs. Exchange of +2 cations was assumed to be associated with the M(OH) + species for the metal (M) ion in the caustic solution. A series of batch contact testing was conducted to evaluate the exchange behavior of Ba, Ca, Pb, and Sr onto CST. The Cs exchange behavior was also tested as a benchmark for direct comparisons. The CST was provided in the sodium form by Honeywell UOP, as IONSIV TM R9140-B, Lot 2002009604, 18 x 50 mesh. A <30-mesh aliquot was collected to match the sieve fraction expected for use in upcoming small column test configurations. Kinetic exchange rate and isotherms were measured at metal concentrations benchmarked from the AP-107 tank waste feed composition and as limited by the metal solubility in the alkaline solution. Two simplified matrices were tested: 1) 1.0 M NaOH/4.6 M NaNO 3 and 2) 0.1 M NaOH/5.5 M NaNO 3 ; these matrices represented the expected 5.6 M Na concentration of process feed and served to address the hydroxide concentration effect on exchange behavior.

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Spent Crystalline Silicotitanate Storage Study—Post AP-105DF Processing

A Tank-Side Cesium Removal (TSCR) system is under development by Washington River Protection Solutions to support initial production of immobilized low-activity waste (LAW) by feeding Hanford tank supernate from tank farms to the Hanford Waste Treatment and Immobilization Plant (WTP) LAW Facility. Tank waste supernate is filtered to remove suspended solids and then Cs is removed by processing it through crystalline silicotitanate (CST) ion exchange media manufactured by Honeywell UOP, LLC. The Cs-loaded CST columns will be stored indefinitely, with a goal of eventual CST removal and treatment. Thus, the spent CST needs to be recoverable after undetermined storage time. Previous testing with AP-105 simulant showed that rinsing the CST bed with 3 bed volumes (1.4 apparatus volumes [AVs]) of 0.1 M NaOH resulted in a dried bed that maintained flow characteristics indicative of ease of recovery. This study explored the intermediate conditions (between feed dried in place and the 1.4 AVs of 0.1 M NaOH rinse) to evaluate CST bed properties.

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Round Robin Summary for ASTM Method C1733

This report was prepared to summarize the results of an inter- and intra-laboratory round robin of ASTM Method C1733-17 (now C1733-20a) Distribution Coefficients of Inorganic Species by the Batch Method. The testing used the sorption of cesium (Cs) to a standard reference soil from the San Joaquin Valley (Standard Reference Material 2709a). The data collected from the nine participants was used for a statistical analysis to assess the precision of the method and generate a precision and bias statement to update the C1733-20a standard.

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Large Surface Explosion Coupling Experiment - SNL Remote Optical.

Two surface chemical explosive tests were observed for the Large Surface Explosion Coupling Experiment (LSECE) at the Nevada National Security Site in October 2020. The tests consisted of two one-ton explosions, one occurring before dawn and one occurring mid- afternoon. LSECE was performed in the same location as previous underground tests and aimed to explore the relationship between surface and underground explosions in support of global nonproliferation efforts. Several pieces of remote sensing equipment were deployed from a trailer 2.02 km from ground zero including high-speed cameras, radiometers and a spectrometer. The data collected from these tests will increase the knowledge of large surface chemical explosive signatures.

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Uranium Mass Attribute System (UMAS) Study: Evaluation of Options Against the Arms Control Evaluation Criteria

In this study, eleven possible radiation and other non-destructive measurement systems for confirming U-235 mass attribute were identified. The systems were reviewed for their applicability in a potential future arms control agreement, and a subset of the eight most plausible systems were selected for futher evaluation against the Arms Control Evaluation Criteria (ACEC). The evaluation results were then used to identify a primary and secondary recommended UMAS for confirming a U-235 mass attribute on warheads, nuclear components of warheads, and/or materials as part of a hypothetical arms control agreement.

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Physical Sciences Vistas, Issue 1 2021 [Newsletter]

Issue 1 highlights of the Los Alamos Physical Sciences Vistas newsletter include: Finding and aiding success through the Lab's employee scholarship fund; Volunteer vignettes: helping others throughout Northern New Mexico; X-ray topography instrument reimagined for use at Oregon State University; Potentially activated metal recycle project triples its goal; Adaptive 3D machine-learning method for 3D coherent diffraction imaging; and, Sigma team aims for excellence in safety, quality, productivity.

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Recalculation of Soil Bulk Density Used in the Scorpius MCNP6 Model

The soil composition used in the Scorpius MCNP6.2 (Ref. 1) model was recently recalculated. Originally, it was “determined using the United States Geological Survey (USGS) Mercury Core Library and the Nevada National Security Site U.S. Geological Survey Databases (NNSS USGS) and Nevada National Security Site Petrographic, Geochemical, and Geophysical Database (NNSS PGG),” but the details had been lost. Reference 2 documents the updated soil composition determination. Reference 2 estimated the grain density of the soil instead of the bulk density. This report corrects that error. I am indebted to Garrett Euler (EES-17) for reading Ref. 2, pointing out its errors, and guiding me through the correct calculations. This report is organized as follows. For completeness, the composition calculations from Ref. 2 are repeated: Sec. II discusses the data that are reported in the PGG Access database, and Sec. III uses the data in the PGG Access database to compute the composition of the soil. Section IV estimates the bulk density of the soil. Section V evaluates the soil wall thickness in the Scorpius model. Section VI estimates the effect of the soil density change on previously calculated results. Section VII presents a summary and conclusions. Input files and output files are listed in appendices.

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Requirements and Conceptual Design of Off-gas Systems for the Reprocessing of Metallic Fuels

An assessment has been conducted to determine how key regulations regarding volatile radionuclide emissions to the atmosphere may apply to the off-gas streams associated with electrochemical reprocessing. The scope of this assessment was based upon a generic electrochemical reprocessing scheme with a throughput rate of 200 MTIHM/y applied to metallic fuel discharged from a sodium fast reactor (SFR), but the findings are able to be translated to other advanced nuclear scenarios as merited. Air dispersion modeling was performed using the EPA CAP-88 model and evaluated the uncontrolled decontamination factors (DFs) that would be required to achieve regulatory compliance with the dose-based limits set forth by EPA regulation 40 CFR 190.10(a). These DFs were compared to those required by fuel cycle–based limits set forth by EPA regulation 40 CFR 190.10(b). Two theoretical sites with disparate climatological conditions were selected for air dispersion modeling (Idaho and Tennessee). The radionuclides modeled included 3 H, 85 Kr, 129 I, and selected alpha-emitting transuranic isotopes (referred to here as 239 Pu-TRU <1y ). It was found that the fuel cycle-based limits in 40 CFR 190.10(b) are most restrictive for 85 Kr and 239 Pu-TRU <1y , with DFs of 3 and 6.1E+09, respectively. The dose-based limit as derived from 40 CFR 190.10(a) could require mitigation of tritium in some scenarios, with an estimated DF of about 3 for the reference scenarios. The fuel cycle-based limit for 129 I resulted in a DF of about 240 for the reference scenario. The need for iodine mitigation based on dose to the public depended upon the physical form of iodine as either particulate or vapor-phase species. Emission of iodine from the facility as a vapor necessitated DFs of about 2 but emission as a particulate would require DFs >6,000 to meet thyroid dose-based limits. Effects of physical form on needed iodine mitigation are significant, but the understanding of speciation of iodine both during electrochemical reprocessing and after release to the atmosphere is limited. The electrochemical processing unit operations were evaluated to identify potential release points for the volatile radionuclides and to assess the potential for retention of the radionuclides within the process (thus decreasing the need for mitigation). Mitigation strategies for 3 H, 85 Kr, 129 I, and 239 Pu-TRU <1y were identified. In all cases, there are reasonably achievable pathways to regulatory compliance, although in some cases additional R&D is merited to verify the chemical speciation of these isotopes and to develop and demonstrate potential treatment technologies for this application. Whether or not additional off-gas controls (beyond common operations such as HEPA filtration and oxygen and moisture control) are needed for any of these regulated or volatile radionuclides depends on the (a) type of facility (NRC-regulated or DOE), (b) used fuel process rate, (c) used fuel burnup and composition, (d) speciation and retention of volatile radionuclides in the process and in the cell gas cleanup system, (e) site-specific parameters such as location, meteorology, stack height, and site boundaries, and (f) levels of conservatism and safety factors used in assessing compliance to air emissions regulations. Performance of this assessment revealed several areas where information is lacking or additional research is required in order to better determine if or what kinds of off-gas control might be needed. First, and most significantly, the understanding of the chemical speciation and physical form and partitioning of iodine during electrochemical processing operations is lacking and prevents the ability to accurately assess the potential iodine mitigation requirements. Future research in this area should be multifaceted and include thermodynamic modeling of iodine speciation in different process steps, experiments to quantify the kinetics of vapor-phase and melt-phase transitions, bench-scale experiments to determine the potential chemical and physical form of iodine emissions from the electrorefining process, and verification of iodine behavior with experiments utilizing operational facilities. Similarly, an improved understanding of iodine behavior in the environment after release from the facility stack will be required to refine dose estimations, as particulate and vapor-phase emissions result in significantly different doses to the MEI.

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C-NR Countroom Overview [Slides]

Perform qualitative and quantitative assay of gamma, beta, and alpha-emitting radionuclides in a variety of matrices and over a wide range of activity levels.

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Milestone 2.8: Preliminary Radiolytic Gas Generation Measurements from Helium-Backfilled Samples

The Department of Energy (DOE) is currently managing nearly 13 metric tons of aluminum-clad spent nuclear fuel (ASNF) with the intention of extended (> 50 years) dry storage in helium-backfilled canisters. Due to in-reactor and cooling pond conditions, oxyhydroxide corrosion layers have formed on the surface of the ASNF elements. These corrosion layers are susceptible to radiolysis and the formation of molecular hydrogen gas (H 2 ) due to the fuel’s inherent radiation field. Consequently, a rigorous evaluation of the effect of helium gas on radiolytic H 2 production is necessary to support the Technical Considerations and Challenges for Extended (> 50 yrs) Dry Storage of ASNF program, especially as previous Task 2 - Oxyhydroxide Layer Radiolytic Gas Generation Resolution work demonstrated a significant effect of gas composition on the radiolytic yield (G-value) of H 2 . Here we report preliminary G-values for the radiolytic formation of H 2 from the gamma irradiation of pre-corroded aluminum alloy 1100 coupons flame sealed in helium environments. Irradiations yielded G(H 2 ) values of (5.1 ± 0.5) x 10 –4 and (9.4 ± 0.9) x 10 –4 µmol J –1 for pristine coupons, and (10.1 ± 0.4) x 10 –4 and (15.1 ± 1.2) x 10 –4 µmol J –1 for pre-corroded coupons for 0% and 50% relative humidity, respectively. These helium environment G(H 2 ) values are between 28% and 58% higher than previously reported values for argon environments. This enhancement is attributed to the significant difference in first ionization energy between helium (24.59 eV) and argon (15.76) facilitating additional processes, e.g., Penning ionization. These new preliminary helium environment G(H 2 ) values will be employed by Task 3 - Sealed and Vented System Episodic Breathing and Gas Generation Prediction to model the effect of radiolytic H 2 accumulation in helium environments to evaluate the practicality of the extended storage standard canister design.

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NBL Pu CRM (Phase 1 Progress Report)

Several 100 μg aliquots of NBL CRM 137 Pu sulfate tetrahydrate were prepared in 20 mL glass vials at some time in the past and are expected to have similar chemical composition and similar concentrations of contaminants as the bulk 250 mg CRM 137 Pu sulfate tetrahydrate units. Thus, these units were characterized and used to test Pu purification chemistry to be employed on the larger unit. For preliminary characterization, the residue of bottle 6 was dissolved in 4 M HNO3/0.01 M HF and transferred quantitatively to a pre-weighed 30 mL Savillex Teflon vial with a total solution mass of approximately 18 g. A gravimetric aliquot was removed from this solution for characterization and the remaining solution was dried down and reconstituted in 8 M HNO 3 to test the anion exchange purification column.

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SCEPTRE 2.2 Quick Start Guide

This report provides a summary of notes for building and running the Sandia Computational Engine for Particle Transport for Radiation Effects (SCEPTRE) code. SCEPTRE is a general- purpose C++ code for solving the li near Boltzmann transport equation in serial or parallel using unstructured spatial finite elements, multigroup energy treatment, and a variety of angular treatments including discrete ordinates and spherical harmonics. Either the first-order form of the Boltzmann equation or one of the second-order forms may be solved. SCEPTRE requires a small number of open-source Third Part y Libraries (TPL) to be available, and example scripts for building these TPLs are provided. The TPLs needed by SCEPTRE are Trilinos, boost, and netcdf. SCEPTRE uses an autotools build system , and a sample configure script is provided. Running the SCEPTRE code requires that the user provide a spatial finite-elements mesh in Exodus format and a cross section library in a format that will be described. SCEPTRE uses an xml-based input, and several examples will be provided.

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SCEPTRE 2.2 Angular Quadrature Sets

This document includes details of the angular quadrature sets available in SCEPTRE for performing numerical integrations in the angular phase space. The angular dependence of the boundary and fixed-source terms an d initial angular flux are specified by angular index rather than by direction. It is, therefore, necessary to know the mapping from a specific direction to a direction index. This document includes angular quadrature weights and direction cosines for most of the quadrature sets available in SCEPTRE.

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