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At least 937 records · Page 52

The new K, Pb-bearing uranyl-oxide mineral kroupaite: Crystal-chemical implications for the structures of uranyl-oxide hydroxy-hydrates

Abstract Kroupaite (IMA 2017-031), ideally KPb0.5[(UO2)8O4(OH)10]·10H2O, is a new uranyl-oxide hydroxylhydrate mineral found underground in the Svornost mine, Jáchymov, Czechia. Electron-probe micro-analysis (WDS) provided the empirical formula (K1.28Na0.07)Σ1.35(Pb0.23Cu0.14Ca0.05Bi0.03Co0.02Al0.01)Σ0.48 [(UO2)7.90(SO4)0.04O4.04(OH)10.00]·10H2O, on the basis of 40 O atoms apfu. Sheets in the crystal structure of kroupaite adopt the fourmarierite anion topology, and therefore kroupaite belongs to the schoepite-family of minerals with related structures differing in the interlayer composition and arrangement, and charge of the sheets. Uptake of dangerous radionuclides (90Sr or 135Cs) into the structure of kroupaite and other uranyl-oxide hydroxy-hydrate is evaluated based on crystal-chemical considerations and Voronoi-Dirichlet polyhedra measures. These calculations show the importance of these phases for the safe disposal of nuclear waste.

Geochemistry & Geophysics↗

Finchite, Sr(UO 2 ) 2 (V 2 O 8 )·5H 2 O, a new uranyl sorovanadate with the francevillite anion topology

Finchite (IMA2017-052), Sr(UO 2 ) 2 (V 2 O 8 )·5H 2 O, is the first uranium mineral known to contain essential Sr. The new mineral occurs as yellow-green blades up to ~10 µm in length in surface outcrops of the calcrete-type uranium deposit at Sulfur Springs Draw, Martin County, Texas, U.S.A. Crystals of finchite were subsequently discovered underground in the Pandora mine, La Sal, San Juan County, Utah, U.S.A., as diamond-shaped golden-yellow crystals reaching up to 1 mm. In this work, the crystal structure of finchite from both localities was determined using single-crystal X-ray diffraction and is orthorhombic, Pcan, with a = 10.363(6) Å, b = 8.498(5) Å, c = 16.250(9) Å, V = 1431.0(13) Å 3 , Z = 4 (R 1 = 0.0555) from Sulfur Springs Draw; and a = 10.3898(16), b = 8.5326(14), c = 16.3765(3) Å, V = 1451.8(4) Å 3 , Z = 4 (R 1 = 0.0600) from the Pandora mine. Electron-probe microanalysis provided the empirical formula (Sr 0.88 K 0.17 Ca 0.10 Mg 0.07 Al 0.03 Fe 0.02 ) Σ1.20 (UO 2 ) 2 (V 2.08 O 8 )·5H 2 O for crystals from Sulfur Springs Draw, and (Sr 0.50 Ca 0.28 Ba 0.22 K 0.05 ) Σ0.94 (U 0.99 O 2 ) 2 (V 2.01 O 8 )·5H 2 O for crystals from the Pandora mine, based on 17 O atoms per formula unit. The structure of finchite contains uranyl vanadate sheets based upon the francevillite topology. Finchite is a possible immobilization species for both uranium and the dangerous radionuclide 90 Sr because of the relative insolubility of uranyl vanadate minerals in water.

58 GEOSCIENCES↗

Analysis of DWPF Sludge Batch 6 (Macrobatch 7): Pour Stream Glass Samples

The Defense Waste Processing Facility (DWPF) began processing Sludge Batch 6 (SB6), also referred to as Macrobatch 7 (MB7), in June 2010. SB6 is a blend of the heel of Tank 40 from Sludge Batch 5 (SB5), H-Canyon Np transfers and SB6 that was transferred to Tank 40 from Tank 51.1 SB6 was processed using Frit 418. Sludge is received into the DWPF Chemical Processing Cell (CPC) and is processed through the Sludge Receipt and Adjustment Tank (SRAT) and Slurry Mix Evaporator Tank (SME). The treated sludge slurry is then transferred to the Melter Feed Tank (MFT) and fed to the melter. During processing of each sludge batch, the DWPF is required to take at least one glass sample to meet the objectives of the Glass Product Control Program (GPCP) and to complete the necessary Production Records so that the final glass product may be disposed of at a Federal Repository. The DWPF requested various analyses of radioactive glass samples obtained from the melter pour stream during processing of SB6 as well as reduction/oxidation (REDOX) analysis of MFT samples to determine the impact of Argon bubbling. Sample analysis followed the Task Technical and Quality Assurance Plan (TTQAP) and an Analytical Study Plan (ASP). Four Pour Stream (PS) glass samples and two MFT slurry samples were delivered to the Savannah River National Laboratory (SRNL) from the DWPF. Table 1-1 lists the sample information for each pour stream glass sample. SB6 PS3 (S03472) was selected as the official pour stream sample for SB6 and full analysis was requested. This report details the visual observations of the as-received SB6 PS No.3 glass sample as well as results for the chemical composition, Product Consistency Test (PCT), radionuclide content, noble metals, and glass density. REDOX results will be provided for all four pour stream samples and vitrified samples of MFT-558 and MFT-568A. Where appropriate, data from other pour stream samples will be provided.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Analysis Of The Sludge Batch 7b (Macrobatch 9) DWPF Pour Stream Glass Sample (Rev. 1)

The Defense Waste Processing Facility (DWPF) began processing Sludge Batch 7b (SB7b), also referred to as Macrobatch 9 (MB9), in January 2012. SB7b is a blend of the heel of Tank 40 from Sludge Batch 7a (SB7a) and the SB7b material that was transferred to Tank 40 from Tank 51. SB7b was processed using Frit 418. During processing of each sludge batch, the DWPF is required to take at least one glass sample to meet the objectives of the Glass Product Control Program (GPCP), which is governed by the DWPF Waste Form Compliance Plan, and to complete the necessary Production Records so that the final glass product may be disposed of at a Federal Repository. Two pour stream glass samples were collected while processing SB7b. The samples were transferred to the Savannah River National Laboratory (SRNL) where one was analyzed and the other was archived. The following conclusions were drawn from the analytical results provided in this report: The sum of oxides for the official SB7b pour stream glass is within the Product Composition Control System (PCCS) limits (95-105 wt%); The average calculated Waste Dilution Factor (WDF) for SB7b is 2.3. In general, the measured radionuclide content of the official SB7b pour stream glass is in good agreement with the calculated values from the Tank 40 dried sludge results from the SB7b Waste Acceptance Program Specification (WAPS) sample; As in previous pour stream samples, ruthenium and rhodium inclusions were detected by Scanning Electron Microscopy-Electron Dispersive Spectroscopy (SEM-EDS) in the SB7b pour stream sample; The Product Consistency Test (PCT) results indicate that the official SB7b pour stream glass meets the waste acceptance criteria for durability with a normalized boron release of 0.8 g/L, which is an order of magnitude less than the Environmental Assessment (EA) glass; The measured density of the SB7b pour stream glass was 2.70 g/cm 3 ; The Fe 2+ /ΣFe ratio of the SB7b pour stream samples was 0.07.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Summary Report For The Analysis Of The Sludge Batch 7b (Macrobatch 9) DWPF Pour Stream Glass Sample For Canister S04023 (Rev. 1)

In order to comply with the Defense Waste Processing Facility (DWPF) Waste Form Compliance Plan for Sluldge Batch 7b, Savannah River National Laboratory (SRNL) personnel characterized the Defense Waste Processing Facility (DWPF) pour stream (PS) glass sample collected while filling canister S04023. This report summarizes the results of the compositional analysis for reportable oxides and radionuclides and the normalized Product Consistency Test (PCT) results. The PCT responses indicate that the DWPF produced glass that is significantly more durable than the Environmental Assessment glass.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Initial Simulations of Empty Room Collapse and Reconsolidation at the Waste Isolation Pilot Plant

The Waste Isolation Pilot Plant (WIPP) is a geologic repository for defense-related nuclear waste. If left undisturbed, the virtually impermeable rock salt surrounding the repository will isolate the nuclear waste from the biosphere. If humans accidentally intrude into the repository in the future, then the likelihood of a radionuclide release to the biosphere will depend significantly on the porosity and permeability of the repository itself. Room ceilings and walls at the WIPP tend to collapse over time, causing rubble piles to form on floors of empty rooms. The surrounding rock formation will gradually compact these rubble piles until they eventually become solid salt, but the length of time for a rubble pile to reach a certain porosity and permeability is unknown. This report details the first efforts to build models to predict the porosity and permeability evolution of an empty room as it closes. Conventional geomechanical numerical methods would struggle to model empty room collapse and rubble pile consolidation, so three different meshless methods, the Immersed Isogeometric Analysis Meshfree, Reproducing Kernel Particle Method (RKPM), and the Conformal Reproducing Kernel method, were assessed. First, the meshless methods and the finite element method each simulated gradual room closure, without ceiling or wall collapse. All three methods produced equivalent room closure predictions with comparable computational speed. Second, the Immersed Isogeometric Analysis Meshfree method and RKPM simulated two-dimensional empty room collapse and rubble pile consolidation. Both methods successfully simulated large viscoplastic deformations, fracture, and rubble pile rearrangement to produce qualitatively realistic results. In addition to geomechanical simulations, the flow channels in damaged salt and crushed salt were measured using micro-computed tomography, and input into a computational fluid dynamics simulation to predict the salt's permeability. Although room for improvement exists, the current simulation approaches appear promising.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Software Requirements for a Consequence Management Sample Data Simulator for Training and Drills

This document describes the requirements for a software tool that will enable FRMAC to simulate large sets of sample result data that is based realistically on simulated radionuclide deposition grids from NARAC. The user of this tool would be scientists involved in exercise and drill planning or part of the simulation cell of an exercise controller team. A key requirement is that this tool must be able to be run with a reasonable amount of training and job aids by any person within the Assessment, Laboratory Analysis, or Monitoring and Sampling divisions of the FRMAC to support any level of exercise from the small IPX to the national level full scale exercise. This tool should be relatively lean and stand-alone so that the user can run it in the field with limited IT resources. This document will describe the desired architecture, design characteristics, order of operations, and algorithms that can be given to a software development team to assist them in project scoping, costing, and eventually, development.

61 RADIATION PROTECTION AND DOSIMETRY↗

U.S. Effort Support to Examinations at Fukushima - Meeting Notes with Updated Information Requests (FY2020)

Much is still not known about the end-state of core materials in each unit that was operating on March 11, 2011 at the Fukushima Daiichi Nuclear Power Station (Daiichi). Information obtained from Daiichi is required to inform Decontamination and Decommissioning (D&D) activities, improving the ability of the Tokyo Electric Power Company Holdings, Incorporated (TEPCO Holdings) to characterize potential hazards and to ensure the safety of workers involved with cleanup activities. This document, which has been updated to include Fiscal Year 2020 (FY2020) information, summarizes results from a U.S. effort to review Daiichi information and extract insights to enhance the safety of existing and future nuclear power plant designs. This U.S. effort, which was initiated in 2014 by the Department of Energy Office of Nuclear Energy (DOE-NE), is completed by a group of experts in reactor safety and plant operations that identify examination needs and evaluate recent Daiichi examination data to address these needs. Since its inception, comprehensive annual reports were issued that document significant safety insights being obtained in areas of special emphasis: system and component performance, radionuclide surveys and sampling, debris end-state location, combustible gas effects, and plant operations and maintenance. In addition to reducing uncertainties related to severe accident modeling progression, these insights have and continue to be used to update guidance for severe accident prevention, mitigation, and emergency planning. Reduced uncertainties in modeling the events at Daiichi improve the realism of reactor safety evaluations that inform future D&D activities. For FY2020, it was decided that the program would gain more benefit from a more concise report that emphasizes new information and insights that affect prior findings and recommendations from the U.S. experts participating in this effort. A key aspect of prior U.S. efforts, the updated list of information requests, is included in this letter report to ensure that they are transmitted to organizations within Japan. In addition, findings and associated recommendations are provided regarding information presented by TEPCO Holdings, Nuclear Damage Compensation and Decommissioning Facilitation Corporation (NDF), and the Japan Atomic Energy Agency (JAEA). This letter report also continues to emphasize how information obtained from the affected reactors at Daiichi has been and will continue to be used to update severe accident management strategies, improve maintenance activities (especially in areas of radiation protection) and reduce uncertainties in systems analysis code models. In addition, recommendations are included that would expand the use of this information to provide insights regarding maintenance, radiation protection, design, and siting activities of existing and new reactors.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Using Radioiodine Speciation to Address Environmental Remediation and Waste Stream Sequestration Problems at the Fukushima Daiichi Nuclear Power Plant and a DOE Site

Iodine-129 ( 129 I), with a half-life of half-life of 16 million years, is commonly considered the single greatest risk driver in high-level and low-level nuclear repositories. This risk stems from several basic properties of 129 I, and under many geochemical conditions, it can move as an anion at nearly the rate of water through the subsurface environment. 129 I is also extremely radiologically toxic because over 90% of body burden accumulates in the thyroid, which weighs only about 14g in an adult. There is also a large worldwide inventory of radioiodine as a result of its high fission yield and this inventory is rapidly increasing as a result of nuclear energy production. Radioiodine is produced at a rate of 40 GBq (1 Ci) per gigawatt of electricity produced by nuclear power. To illustrate how the properties of 129 I magnify its risk, 129 I accounts for only 0.00002% of the radiation released from the Savannah River Site in Aiken, South Carolina, but contributes 13% of the population dose, a six orders of magnitude magnification of risk with respect to its radioactivity. The currently favored solid phase for LLW immobilization is cement, while HLW immobilization is the incorporation of waste into glass (vitrification). However, so far, the incorporation into cement and subsequent leaching of only iodide has been seriously investigated. The major problem with this approach is that it ignores the complex speciation of iodine, i.e., it ignores iodate and organo-iodine which have different chemistries. Most of the past research was devoted to the mechanism of iodide uptake in cement hydrate phases that is sorption and/or incorporation. Very few data exist on iodate and organo-iodine incorporation in cement, even though large quantities of liquid waste containing also radioiodine have already been solidified in cement Iodine-129 from low-level waste is commonly disposed of in cementitious materials. Grout, a dense cementitious fluid, mixed with a reducing slag, is often used to immobilize radionuclides. However, the reducing environment might not be conducive to immobilize iodine. For example, the silver based immobilization technologies (e.g., AgCl, Ag-impregnated granular activated carbon, Ag-mordenite) remove iodine from the aqueous phase by promoting the formation of Ag-iodide precipitates. The solubility of AgI is eight orders of magnitude lower than it is for AgIO 3 . Similarly, coprecipitation of iodine into calcium carbonate phases occurs only with IO 3 - and not with I - and org-I. If one would want to immobilize iodine more effectively, different engineering approaches would need to be used to promote binding of I - , IO 3 - , or organo-I. Using laboratory experiments with grout, slag, and silver-based adsorbents, and GC-MS and I K-edge XANES and EXAFS and C K-edge XANES spectroscopy for identifying iodine speciation, the major problems with these methods have been identified as focused too much on just one of the iodine species for immobilization, while others, especially organo-I, remained mobile. Finally, we established that most of the adsorbents that are used contain sufficient amounts of organic matter to create organo-I . It is anticipated that increased attention directed at understanding and quantifying the speciation of radioiodine, as opposed to simply total radioiodine, will lead to improved remediation results to be used for long-term radioiodine disposal in cementitious waste forms.

129-Iodine↗

Performance Assessment for the Environmental Restoration Disposal Facility (FY 2019 Annual Status Report)

DOE O 435.1 and DOE M 435.1-1 require that a determination of continued adequacy of the performance assessment (PA), composite analysis, and disposal authorization statement (DAS) be annually, and it must consider the results of data collection and analysis from research, field studies, and monitoring as well as update any Radioactive Waste Management Basis documents. Beginning in 1996, the Environmental Restoration Disposal Facility (ERDF) has accepted low-level radioactive, hazardous, and mixed wastes that were generated during cleanup activities at the Hanford Site. ERDF is composed of a series of cells (currently 10) or disposal areas and can accommodate future design expansions as needed. During this reporting period (fiscal year 2019, October 1, 2018, to September 30, 2019). approximately 139,721 U.S. tons of waste was disposed at ERDF. From inception through September 30, 2019, ERDF has received approximately 18.5 million U.S. tons of waste, which equates to consumption of approximately 88% of the disposal volume. As a condition of the DAS, disposal operations within ERDF must be in accordance with the waste acceptance criteria (WAC) that provide specific radionuclide disposal limits, waste from restrictions, and descriptions of acceptable waste packages in compliance with DOE M 435-1.1 requirements. The ERDF WAC stipulates that waste destined for disposal at ERDF be controlled based on source, physical form, and containment concentration and activity levels

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Performance Assessment for the Disposal of Low Level Waste in the 200 East Area Burial Grounds (FY2019 Annual Status Report)

This annual review provides the projected dose estimates of radionuclide inventories disposed in the 200 East Area Low-Level Waste Burial Grounds (LLBGs) since September 26, 1988. These estimates are calculated using the original dose methodology developed in the performance assessment (PA) analysis (WHC-SD-WM-TI-730) and are compared with U.S. Department of Energy requirements (DOE O 435.1 and performance objectives defined in companion documents DOE M 435.1 and DOE-STD-5002-2017). All performance objectives are currently satisfied, and operational waste acceptance criteria (HNF-EP-0063) and waste acceptance practices continue to be sufficient to maintain compliance with performance objectives. Inventory estimates and associated dose estimates from future waste disposal actions are unchanged from previous years’ evaluations, indicating that potential impacts are well below performance objectives; therefore, future compliance with DOE O 435.1 is expected. A new PA study has been initiated in this fiscal year (FY) 2019 for evaluation of active disposal sites within the 200 West and East Areas (Trenches 31 and 34 in 200 West; Trench 94 in 200 East) because extended time has elapsed between the current annual status report and the original PA (WHC-SD-WM-TI-370) for the active disposal sites. The new PA for the active disposal sites is expected to be completed in FY 2021.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Performance Assessment for the Disposal of Low Level Waste in the 200 West Area Burial Grounds (FY2019 Annual Statius Report)

This annual review provides the projected dose estimates of radionuclide inventories disposed in the active 200 West Area Low-Level Waste Burial Grounds (LLBGs) since September 26, 1988. These estimates are calculated using the original dose methodology developed in the performance assessment (PA) analysis (WHC-EP-0645) and are compared with performance objectives defined in U.S. Department of Energy requirements (DOE O 435.1 and its companion documents DOE M 435.1-1 and DOE-STD-5002-2017). All performance objectives are currently satisfied, and operational waste acceptance criteria (HNF-EP-0063) and waste acceptance practices continue to be sufficient to maintain compliance with performance objectives. Inventory estimates and associated dose estimates from future waste disposal actions are unchanged from previous years’ evaluations, indicating that potential impacts are well below performance objectives; therefore, future compliance with DOE O 435.1 is expected.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Solidification of REDC Organics and Airborne Release Fraction Measurements

The Radiochemical Engineering Development Center (REDC) generates various liquid organic wastes from processing irradiated targets to recover heavy elements. In the past, these organics were discharged to the Oak Ridge National Laboratory (ORNL) liquid low-level waste (LLLW) system, along with the aqueous waste. Because of a reduction in aqueous LLLW from other ORNL generators and an increase in the radionuclide concentration, particularly 238 Pu, in the organic waste stream, the organics can no longer be discharged to the LLLW system. The plan is to solidify the liquid organic waste for disposal as solid debris waste. The documented safety analysis for REDC uses an airborne release fraction of 5 x 10 -4 with a respirable fraction of 0.2 for plutonium and other actinides from packaged solid debris waste during a fire scenario, for an overall airborne release fraction/respirable fraction of 10 -4 . Solidifying the organic liquids, either by using PM-199 Organoclay ® or by grouting, would produce solid waste forms with no free liquid, which should qualify for disposal at the Waste Isolation Pilot Plant. The use of Organoclay would be much simpler to implement inside a hot cell. Granular Organoclay can be added to the organic waste solutions until there is a dry layer on top of the liquid and then allowed to cure for a few hours. Grouting the liquids would require using multiple ingredients and mixing steps. Both methods produced waste forms that trapped cerium, as a surrogate for plutonium, inside the waste, with airborne release fractions during fire scenario tests of well below the 10 -4 limit.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

SRS Composite Analysis Monitoring Plan (FY2019 Model Validation)

Using a projected end-state date of 2065 (SRNS 2015b), the Savannah River Site (SRS) Composite Analysis (CA) modeling for each facility and waste site began on the inventory year assigned to it so that source depletion and radionuclide transport out of the system could be appropriately captured. Some SRS waste sites that have already achieved their end states (i.e. end-state inventories and end-state configuration) are currently contributing to the potential off-site public dose through source release, groundwater transport, discharge to on-site surface streams, and stream transport to the CA point of assessments (POAs). The inventory year assigned to these waste sites is 2002 or before. This means that SRS CA results from 2002 and beyond are a reasonable representation for these waste sites that have already achieved their end states and are currently contributing to the potential off-site public dose. The SRS Annual Environmental Report (AER) monitoring can differentiate and separate liquid pathway data allowing the data representing only waste sites at their end state to be produced. Because the SRS CA has projected reasonable end-state impacts from 2002 and beyond, and the AER monitoring can differentiate and separate operating and end-state contributions to annual liquid pathway release, an opportunity exists to use the AER monitoring data to validate the SRS CA model. The CA model validation program uses a graded and systematic approach for taking corrective action, starting with an SRS established administrative dose limit of 15 mrem/yr, below which no action is required. Based on the location of the 2010 SRS CA POAs, the only potential exposure pathway for the public is through surface water. The completion of the FY2019 CA model validation indicates that the SRS CA projected dose, while generally conservative, provides a reasonable representation of the maximum annual doses. These doses are well below the administrative limit; therefore, no additional action is required.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Evaluation of The NSUF Reactor Activation and Damage (RAD) Calculator: Assessing the Accuracy of the Activation Calculator

Nuclear Science User Facilities developed an activation calculator as part of the Combined Material Experiment Toolbox project. The activation calculator estimates the radionuclide concentration in an irradiated sample to indicate when and where the samples may best be examined to only help with scoping the experiment. To verify the accuracy of the calculator, the results from the specific gamma dose rate [mrem/hr/g] at various cooldown times were compared to equivalent results generated in ORIGEN. Most elements were tested as a sample composed solely of that element. These samples were exposed to thirteen different lengths of irradiation in all reactor positions included in the calculator. For elements of interest (i.e., ones currently included in the Nuclear Fuels and Materials Library), 83% of the test cases fell within the allowed tolerance for this scoping tool of ±50%. The overall error is logarithmic with irradiation length and is almost completely independent of reactor position. The primary source of error is missing nuclear data and reactions. NSUF believes that the RAD Calculator can be deployed to the user community because it is designed to provide estimates to aid researchers in planning experiments, and users are cautioned about the limitations of the calculator.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Monitoring Potential Transport of Radioactive Contaminants in Shallow Ephemeral Channels: FY2019

Desert Research Institute (DRI) conducted a field assessment of the potential for contaminated soil to be transported from the Smoky Site Contamination Area (CA) because of storm runoff. This activity supported U.S. Department of Energy (DOE) Environmental Management Nevada Program (EM NV) efforts to establish post-closure monitoring plans for the Smoky Site Soils Corrective Action Unit (CAU) 550. The work was intended to confirm the likely mechanism of transport and determine the meteorological conditions that might cause the movement of contaminated soils, as well as determine the particle size fraction most closely associated with transported radionuclide-contaminated soils. These data will facilitate the design of the appropriate post-closure monitoring program. In 2011, DRI installed a meteorological monitoring station on the west side of the Smoky Site CA and a hydrologic (runoff) monitoring station within the CA, near the east side. The meteorological station collected air temperature, wind speed, wind direction, relative humidity, precipitation, solar radiation, barometric pressure, soil temperature, and soil water content data. The maximum, minimum, and average or total values (as appropriate) for each of these parameters were recorded for each 10-minute interval. The maximum, minimum, and average water depth in the flume installed at the hydrologic station were also recorded for each 10-minute interval. This report presents the data collected from these stations during fiscal year (FY) 2019. During the FY2019 reporting period, the warmest months were June, July, and August and the coldest were December and February. The highest monthly solar radiation (i.e., sunshine) values were recorded in June, July, and August, whereas December through February had the lowest solar radiation values. Monthly mean wind speeds were highest in the spring (April) and early summer (June). Winds were predominantly from the north, except in February, July, August, and September when winds were mostly from the west. The monthly average relative humidity ranged from 17 percent to approximately 60 percent. Humidity was lowest in the summer and highest in the winter. Monthly total precipitation ranged from 0.0 during August and September to 2.22 inches (in) (56.39 millimeters [mm]) in February. Total precipitation for FY2019 was 7.83 in (198.88 mm). During the reporting period, a single, nonzero, flow event was recorded at the flume. This occurred over two days between March 6 and 7, 2019, in response to 1.16 in (29.46 mm) of precipitation falling within 9.3 hours on March 6, 2019. This event produced a peak water depth of 0.39 in (0.99 centimeters [cm]) in the flume, and runoff was measured for approximately 23.5 hours. However, the maximum recorded flow depth was less than the limit for flow discharge and velocity that can be calculated for the six-inch (15 cm) throat of a Parshall flume. The runoff event produced insufficient flow velocities through the natural channel to cause local erosion and transport of bedload materials. No bedload samples were collected for radiological analysis. Observing meteorological and environmental conditions that lead to storm runoff will help identify the parameters and threshold conditions that should be incorporated into post-closure monitoring plans for this and similar Soils CAU sites. Routine monitoring of meteorological and hydrologic parameters at the Smoky Site CA has been discontinued at the request of EM NV, effective at the conclusion of FY2019. Environmental data acquisition was completed and the meteorological and runoff monitoring stations were removed from the Smoky Site on October 15, 2019.

54 ENVIRONMENTAL SCIENCES↗

Calculation for T Plant Estimated PTE with Receipt of K Basin Sludge and Filter Media

The purpose of this environmental calculation is to: (1) show compliance with the WAC 173-400-110, New Source Review (NSR) for Sources and Portable Sources requirements and (2) provide the methodology, assumptions and calculation used to estimate the potential to emit (PTE) and Total Effective Dose Equivalent to the Maximally Exposed Individual (TEDE to the MEI) radionuclide airborne emissions associated with the receipt of K Basin sludge and filter media for storage at T Plant.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗