SEARCH · Engineering Papers

Results for “radionuclide”

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

SourceResearch connectionFrom yearThrough year

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 415 records · Page 23

All sources NASA NTRS DOE OSTI

NASA NTRS1996

Measurement of proton production cross sections of (sup 10)Be and (sup 26)Al from elements found in lunar rocks

Cosmic rays penetrate the lunar surface and interact with the lunar rocks to produce both radionuclides and stable nuclides. Production depth profiles for long-lived radionuclides produce in lunar rocks are measured using Accelerator Mass Spectrometry (AMS). For a particular radionuclide these production depth profiles can be interpreted to give an estimate for the solar proton flux over a time period characterized by the half life of the radionuclide under study. This analysis is possible if and only if all the cross sections for the interactions of all cosmic ray particles with all elements found in lunar rocks are well known. In practice, the most important cross sections needed are the proton production cross sections, because 98% of solar cosmic rays and (similar to)87% of galactic cosmic rays are protons. The cross sections for the production of long-lived radionuclides were very difficult to measure before the development of AMS and only in recent years has significant progress been made in determining these essential cross sections. Oxygen and silicon are major constituents of lunar rocks. We have reported already C-14 production cross sections from O and Si for proton energies 25-500 MeV, and O(p,x)(sup 10)Be from 58 160 MeV[6]. Here we present new measurements for the cross sections O(p,x)Be-10,O(p,x)Be-7, Si(p,x)Be-7,Si(p,x)Al-26, and Si(p,x)Na-22 from approximately 30 - 500 MeV.

Sisterson, J. M.↗

NASA NTRS2021

On the Cycling of 231 Pa and 230 Th in Benthic Nepheloid Layers

The naturally-occurring radionuclides protactinium-231 ( 231 Pa) and thorium-230 ( 230 Th) are produced at approximately uniform rates in the ocean and thought to be removed from the water column through a reversible exchange with settling particles. Recent measurements along the U.S. GEOTRACES North Atlantic transect (GA03) revealed two features which are at odds with current understanding about 231 Pa and 230 Th cycling in the ocean: (i) a sharp decrease in dissolved 231 Pa ( 231 Pa d ) and 230 Th ( 230 Th d ) activities with depth below 2000-4000 m and (ii) very high particulate 231 Pa ( 231 Pa p ) and 230 Th ( 230 Th p ) activities near the bottom, at a number of stations between the New England continental shelf and Bermuda. Concomitant measurements of light attenuation from beam transmissometry showed that both features occur in benthic nepheloid layers (BNLs), which suggests that these features may stem, at least partly, from the presence of resuspended sediment in the deep water column. Here we explore the behaviour of 231 Pa and 230 Th in BNLs by using (i) radionuclide, optical, and hydrographic data from the western segment of GA03 (west of Bermuda) and (ii) a simplified model of particle and radionuclide cycling that includes a lateral particle source. First, the BNLs observed at GA03 stations are characterized from measurements of the beam attenuation coefficient converted to particle concentrations. At all stations, particle concentrations below the clear water minimum were the highest in the bottom mixed layer, whose thickness ranged from 95 to 320 m, and decreased generally with height above the bottom. The thickness of strong BNLs varied from 482 to 1358 m and the vertical integral of particle concentration in excess to that at the clear water minimum varied from 1 x 10 4 to 2 x 10 6 mg m −2 , among different stations. Second, the particle-radionuclide model is fitted to data from stations GT11-04 (New England continental rise) and GT11-08 (Hatteras abyssal plain), where samples for radionuclide analyses were collected in the BNL. The model can reproduce simultaneously the increase of particle concentration with depth, the low 231 Pa d and 230 Th d in the BNLs, and the high 231 Pa p and 230 Th p near the bottom. According to the model, at heights less than about 300 m above the seafloor, the dissolved phase was set primarily by a balance between adsorption and desorption, with vertical turbulent mixing playing a secondary role, whilst the particulate phase behaved largely as a non-reactive constituent supplied laterally and transported vertically by particle settling and turbulent mixing. Sensitivity tests with the model suggest that lateral particle sources near continental slopes and similar reliefs can produce significant biases both in the 230 Th normalization method and in the interpretation of sediment 231 Pa/ 230 Th records. Our findings yield insights into the influence of sediment resuspension and transport on 231 Pa and 230 Th in the deep ocean and highlight the need for considering these processes in paleoceanographic applications.

Nepheloid layer↗

DOE OSTI2022

Rapid isolation of cyclotron-produced gallium-68

Methods for rapid isolation of radionuclides (e.g., 68Ga) produced using a cyclotron and methods for recycling of the parent isotope (e.g., 68Zn) are disclosed. In one version of the method, a solution including a radionuclide (e.g., 68Ga) is created from a target including cations (e.g., 68Zn). The solution including the radionuclide is passed through a first column including a sorbent comprising a hydroxamate resin to adsorb the radionuclide on the sorbent, and the radionuclide is eluted off the sorbent. The cations (e.g., 68Zn) are recovered from a recovery solution that has passed through the first column by passing the recovery solution through a second column including a second sorbent comprising a cation exchange resin.

62 RADIOLOGY AND NUCLEAR MEDICINE↗

DOE OSTI2020

Impact of UV-light and pH on the Fate of Tc, I, and U in Wetlands at Savannah River Site - 20230

The Savannah River Site (SRS) is one of the major nuclear facilities owned by the U.S Department of Energy. During the Cold War, these facilities produced large amounts of radioactive and hazardous waste. On site, three unlined seepage basins in the F-Area received approximately 1.8 billion gallons of low-level waste containing nitric acid, radionuclides, and dissolved metals due to plutonium and tritium production operations. The acidic nature of the waste created a source of relatively mobile radionuclides below the basins. Radionuclides previously disposed of within the F-Area, including uranium isotopes (U), technetium-99 (Tc-99), and iodine-129 (I-129), are moving with groundwater towards Four Mile Branch Wetland, where they are subsequently upwelling and interacting with natural organic matter (NOM). Many environmental factors including redox conditions, porewater composition, pH, light, and temperature affect the degradation of organic matter as well as interactions with Tc-99, U, and I-129. In particular, the high concentrations of nitrate from the acidic waste may increase the formation of reactive oxygen species (ROS) that impact both degradation of NOM and behavior of contaminants. In the presence of sunlight, I-129, Tc-99, and U speciation and complexation behavior may be affected by ROS and NOM degradation products in addition to the factors generally considered in subsurface systems in the absence of light. This research aims to determine whether the interactions between radionuclides, NOM, and nitrate affect the fate of I-129, Tc-99, and U and which processes are controlling their behavior. Experiments were conducted at variable pH in the presence of NaNO{sub 3} in order to determine the impact of light and pH on NOM degradation and to evaluate the impact on the fate of contaminants of concern. Soil samples high in NOM collected from two sites in the Southeastern United States (Savannah River Site and the Everglades) were studied. Batch experiments were conducted with NOM and the following aqueous contaminants, U, I-129, Tc-99 with exposure to ultraviolet (UV) light in an environmental chamber. This study was conducted in a sterile environment in order to exclude the potential for microbial degradation of organic matter. Results indicated photodegradation of NOM and significant interaction of radionuclides with NOM. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2024

An electrochemical generator for the continual supply of 213 Bi from 225 Ac for use in targeted alpha therapy applications

Bismuth-213 is a radionuclide of interest for targeted alpha therapy and is supplied via a radiochemical generator system through the decay of 225 Ac. Radionuclide generators employ longer lived “parent” radionuclides to routinely supply shorter-lived “daughter” radionuclides. The traditional 225 Ac/ 213 Bi radiochemical generator relies on an organic cation exchange resin where 225 Ac binds to the resin and 213 Bi is routinely eluted. These resins degrade when they absorb large doses of ionizing radiation (>1 × 10 6 Gy/mg), which has been observed when the loading activity of 225 Ac exceeds 2.59*10 9 Bq (70 mCi). Herein we report the development of an electrochemical generator for the supply of 213Bi that has the potential to overcome this limitation. Bismuth-213 spontaneously electrodeposits onto nickel foils in 0.1 M hydrochloric acid at 70 °C. Using this method, we were able to plate an average of 73 ± 4 % of the 213 Bi in solution and obtain a final 213 Bi recovery of 65 ± 8 % in 0.1 M citrate pH 4.5 via reverse electrolysis using titanium as the cathode. The recovered 213Bi had an average radiochemical purity of >99.8 % and was successfully used to radiolabel DOTATATE with an average radiochemical yield of 85.1 % (not optimized).

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2020

Development of a Conditioned System-Level Groundwater Model to Evaluate Long-Term Groundwater Impacts within a Performance Assessment - 20115

A preliminary performance assessment (PA) of single-shell tank Waste Management Area (WMA) A-AX located at the U.S. Department of Energy (DOE)'s Hanford Site in southeastern Washington is being conducted to satisfy the requirements of DOE Order 435.1 [1] as it relates to closure of the single-shell radioactive waste tanks in the WMA A-AX tank farms. A PA assesses the fate, transport, and impacts of radionuclides within a low-level radioactive waste disposal facility in its assumed closure configuration and the subsequent potential doses to humans over a 1,000-year compliance period and a 10,000-year performance evaluation period. The WMA A-AX preliminary PA evaluation is structured around the complementary use of process-level and system-level models to calculate the facility performance against established DOE Order 435.1 [1] performance objectives. Process-level models are those that represent a detailed phenomenological representation of processes of concern in the PA. Process models typically only represent one or a few of the components of the PA, such as groundwater flow and transport, and must be integrated with other modeling elements to perform PA calculations. System-level models are those that are abstracted from the process models, retaining the essential features of the process model, while allowing integration of all aspects of the PA in a single modeling framework. System-level models are often characterized by coarser numerical discretization, lower dimensionality, or other similar simplifications compared to the process-level model. Traditionally, a three-dimensional (3-D) process-level model is utilized primarily to evaluate the long-term impact on groundwater and the potential doses to individuals who consume contaminated groundwater. System-level models are also utilized to evaluate the groundwater pathway in PAs. These models typically have reduced dimensionality (1-D) and are conditioned utilizing flow fields (Darcy fluxes) and moisture content distributions that are abstracted from the process level models. The abstraction approach assures that the flow field in both models is consistent for a specific set of input parameters for flow, differing only in the discretization and dimensionality of the two models. The preliminary WMA A-AX PA incorporates a detailed representation of the geological system and hydraulic properties within the 3-D model STOMP{sup C} numerical code so that the effects of relevant features and processes on water flow and radionuclide transport in the subsurface can be evaluated. The complementary system-level model is developed utilizing the GoldSim{sup C} code to implement a simplified, 1-D equivalent model to represent the groundwater pathway. Contaminant transport through the vadose zone and unconfined aquifer for Tc-99 and I-129 were evaluated in each of the models. Adjustments to the saturated portion of the GoldSim{sup C} 1-D model were required to mimic the dispersion effect captured with the 3-D STOMP{sup C} model. Once this conditioning was conducted, highly similar results for the transport of Tc-99 and I-129 were achieved at the point of calculation, located 100 m downgradient from the WMA A-AX fenceline. These radionuclides represent elements that are regarded as primary dose drivers in the PA groundwater pathway analysis. The high degree of conformance between the two models suggests that the use of the equivalent 1-D system model is suitable for evaluating the full suite of radionuclides that will be released from the tank sources within WMA A-AX over the 1,000-year compliance period and over the 10,000-year evaluation period. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2024

Investigating sol–gel matrix loading capacity toward producing surrogate nuclear explosive debris with realistic composition

Post detonation nuclear forensic materials which resemble the size, color, elemental composition, and radionuclide content of real nuclear debris would be valuable for developing and validating new nuclear forensic techniques. As nuclear fallout types vary significantly, the ability to tailor each of these parameters accurately is desired to produce materials capable of testing analytical methods under a wide array of forensic scenarios. Sol–gel synthesis techniques can provide tunability of size, shape and composition for producing a wide variety of solid nuclear forensics benchmarking materials. Further, the sol–gel process consists of forming a metal oxide material, often silica, through polymerization of a metal-alkoxy precursor. In this work, we characterize the ability to load sol–gel particles with secondary elemental components such as iron, aluminum, and calcium toward producing benchmarking materials approximating the elemental composition of historic nuclear debris from the Nevada National Security Site. We also demonstrate quantitative radionuclide encapsulation toward producing benchmarking materials with controllable radionuclide content. Finally, we employ these techniques to produce nuclear debris benchmarking materials with controllable elemental matrix composition and radionuclide content and compare these samples with the composition of a historic fallout sample previously reported from the Nevada National Security Site.

36 MATERIALS SCIENCE↗

DOE OSTI2021

Penetration of Fission Products Ions into Complex Solids and the Effect of Ionic Wash Methods

During washing of radiologically impacted building surfaces, penetration of radionuclide ions into complex solids associated with these surfaces may occur. This study investigates the penetration of Cs-137, Sr-85, and Eu-152 solutions into numerous common building materials and radionuclide behavior when these materials were exposed to a static bath or low-pressure flow of tap water, 0.1 M potassium chloride (KCl), and 0.5 M KCl. The decontamination efficacy and the depth profile for residual contamination were measured to determine the conditions under which applying a wash solution has benefit compared to physically removing the surface material. On asphalt, 70-80% of the radionuclides were found to be within 0.02 mm of the surface. Concrete is more porous than asphalt, and 80% of the radionuclides were within 0.2 mm of the surface for Cs-137 and Eu-152 and 50-80% for Sr-85. Water effectively removed all contaminants from hard nonporous surfaces. Finally, this paper illustrates that a wash penalty factor concept-defined as ratio of the depth at which 50% of the radioactivity is found in the washed sample divided by the depth at which 50% of radioactivity is found in the control-can serve as a way to quantify whether the wash method increases the depth at which contamination penetrates into the material and thus the material becomes more difficult to decontaminate.

building materials↗

DOE OSTI2025

Investigating the Role of Accident Tolerant Cladding on Source Term Reduction for High-Burnup PWRs Using MELCOR

The use of accident tolerant fuel (ATF) cladding can increase coping times during and beyond design basis accidents. While such gains may be incremental, they provide a margin that can potentially be recovered to enable high-burnup (HBU) operation. Realizing such a margin requires demonstrating that the combination of HBU and ATF has not led to an overall increase in source term. This study investigates the influence of cladding technology (Zr-based, Cr-coated Zr, and FeCrAl) and fuel cycle length (18 and 24 months) on radiological dose at the boundary of the exclusion zone for a four-loop pressurized water reactor to investigate whether ATF claddings can provide such benefits. We analyze a recovered large break loss-of-coolant accident scenario to investigate the impact of transient timescale on the benefits of such coping time increases. The simulations have been performed using the MELCOR and MELCOR Accident Consequence Code System codes. For the cases analyzed, increased fuel cycle length did not necessarily increase radionuclide release and hydrogen generation, as these were found to be sensitive to the core power distribution. Similarly, off-site dose consequence is dominated by short-lived radionuclides that tend to saturate earlier in the burnup, so higher burnup operation did not necessarily increase the source term for the phenomena and transients analyzed here. Delays in recovery of the lowpressure safety injection system increase hydrogen production and radionuclide release, especially between 780 s and 1620 s, due to the nonlinear oxidation and core degradation behavior. Results show that Cr-coated Zr enhances safety by delaying heatup and gap release. Here, when uncertainty propagation on oxidation properties is considered, FeCrAl exhibits the lowest overall radionuclide release and off-site dose throughout the spectrum. However, while the considered “base model” performance is superior under delayed injection scenarios, upper-bound cases display hydrogen generation risk comparable to the Zr-based cladding.

Accident Tolerant Fuel↗

DOE OSTI2026

Statistical Uncertainty of Inhalation Dose Coefficients in Consequence Management: Propagated Dose Uncertainty in ICRP 66 Human Respiratory Tract Model

Reference inhalation dose models rely on deterministic biokinetics and reference computational phantoms, limiting their applicability to the variability present in population-specific exposures encountered in emergency response scenarios. Here, this study introduces REDCAL, a Python-based computational framework developed to propagate uncertainty in inhalation dose coefficients using the International Commission on Radiological Protection (ICRP) Publication 66 Human Respiratory Tract Model. REDCAL integrates ICRP deposition and clearance models, systemic biokinetics, and governing physics principles, and leverages Sandia National Laboratories’ Dakota toolkit for uncertainty quantification via Latin Hypercube Sampling. REDCAL was validated against DCAL, with biokinetic retention results differing by less than 1% and effective dose coefficients by less than 2% across all tested radionuclides. Stochastic sampling introduced variability in dose coefficients, with geometric standard deviations (GSD) in committed effective dose coefficients (CEDC) ranging from 1.0 to 1.5, based on lognormal distribution fits. Analysis demonstrated that variations in the activity median aerodynamic diameter (AMAD) notably influenced the computed CEDC values. Smaller particles (<1 µm) increased doses by 20–30% due to deeper lung deposition and prolonged retention for alpha emitting radionuclides, such as 241 Am and 239 Pu. Radionuclides with fast clearance, such as 133 I, demonstrated a dose reduction exceeding 50%, as AMAD increased beyond 5 µm due to upper airway deposition and rapid mucociliary clearance. The greatest GSD among the radionuclides reported in this study was for 241 Am. In most cases, the largest GSDs in the CEDC were associated with larger particle sizes, an expected outcome, as ICRP Publication 66 defines GSD in particle size as a function of AMAD, resulting in an extended tail of the lognormal distribution. The findings support improved inhalation dose assessments and enhance consequence management strategies for the U.S. Federal Radiological Monitoring and Assessment Center by quantifying uncertainty in dose coefficients and strengthening decision-making for emergency response scenarios.

Biokinetic Modeling↗

DOE OSTI2020

NA-22 Quarterly Report: April-June, 2020

At LLNL we contribute to two projects in NNSA-IAEC Science and Technology Working Group Area V: Basic Science of Waste Management & Subsurface Science. The motivation behind Topic Area V is to evaluate the scientific and safety case for an intermediate borehole to house Israel’s nuclear waste. WM1 (SNL lead) is Thermomechanical Damage (initially called Damage Accumulation and Wellbore Stability). This work focuses on evaluating the damage to the host rock induced by excavation and heat released from a possible radioactive waste repository/borehole. WM2 (LLNL lead) is Radionuclide Facilitated Transport in Carbonate Rock (initially called Colloidal Transport of Radionuclides). This work focuses on better understanding the transport of radionuclides and colloids from a nuclear waste repository/borehole in a vadose zone environment through experimentation and modeling. In the second quarter (this report) our work has been delayed in part due to COVID-19. LLNL went into shelter in place (SIP), minimum safe operations mid-March. Operations at LLNL have been slowly ramping up with 50% of the workforce back onsite in limited capacity and telecommuting widely used. The good news is that our labs opened up the last week in June and experiments are expected to start up again for WM2 Radionuclide Facilitated Transport in July. LLNL’s contribution to WM1 Thermochemical Damage modeling efforts also started back up in late June. We expect to be able to catch up on much of our work by the end of next quarter. The three national laboratories communication on a bi-monthly basis to make sure Area V projects remain on track and we have a monthly conference call with our Israeli counterparts to make sure we are communicating priorities and coordinating project details.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

DOE OSTI2020

NA-22 Quarterly Report: April-June, 2020

54 ENVIRONMENTAL SCIENCES↗

DOE OSTI2020

Vadose Zone Model for A Trenches Area for Composite Analysis

The objectives of the vadose modeling for the updated Hanford Site composite analysis (CA) are to simulate the flow and transport of water and radionuclide releases from the surface to the water table and to provide radionuclide transfer rates for the plateau to river (P2R) model, version 8.3. Water additions include natural recharge and water discharged to the ground as a result of industrial processes associated with Hanford Site operations. Contaminant sources include radionuclides in water discharged to the ground during operations and radionuclides disposed “dry” in solid waste burial grounds or other means. The parallel version of the Subsurface Transport Over Multiple Phases (STOMP) simulator, officially named the exascale Subsurface Transport Over Multiple Phases (eSTOMP), is used to simulate flow and transport for the vadose models. The scope of this ECF is to document the development and results of the A Trenches Area vadose zone model. The A Trenches Area model covers the region between the 200 East Area and southwest of the 216-B-3 Pond. The primary waste sites in this area are cribs associated with PUREX Plant (202-A).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2020

Vadose Zone Model for B Complex for Composite Analysis

The objectives of the vadose modeling for the updated Hanford Site composite analysis (CA) are to simulate the flow and transport of water and radionuclide releases from the surface to the water table and to provide radionuclide transfer rates for the plateau to river (P2R) model, version 8.3. Water additions include natural recharge and water discharged to the ground as a result of industrial processes associated with Hanford Site operations. Contaminant sources include radionuclides in water discharged to the ground during operations and radionuclides disposed “dry” in solid waste burial grounds or other means. The simulation time starts in 1943 and ends at 12070, which is 10,000 years after assumed Hanford Site closure in 2070. The scope of this ECF is to document the development and results of the B Complex vadose zone model. The B Complex model encompasses the region around single-shell tank farms 241-B, 241-BX, and 241-BY, and adjacent cribs, trenches, and burial grounds in the northwestern part of the 200 East Area. Sources of waste to this area were primarily from B Plant (221-B) and U Plant (221-U).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2020

Annual Site Environmental Report (2019)

Los Alamos National Laboratory’s (the Laboratory’s) annual site environmental reports are prepared by the Laboratory’s environmental organizations, as required by U.S. Department of Energy Order 231.1B, Administrative Change 1, Environment, Safety, and Health Reporting, and Order 458.1, Administrative Change 3, Radiation Protection of the Public and the Environment. The following chapters in this report discuss our success in complying with environmental laws, regulations, and orders (Chapter 2, Compliance Summary); how we manage the Laboratory’s environmental performance (Chapter 3, Environmental Programs); how we monitor for air emissions of radioactive materials and climate conditions (Chapter 4, Air Quality); how we monitor for effects of Laboratory operations on groundwater quality (Chapter 5, Groundwater Protection); how we monitor the movement of chemicals and radionuclides by storm water runoff and the levels of chemicals and radionuclides in deposited sediment (Chapter 6, Watershed Quality); how we monitor for the presence, levels, and effects of chemicals and radionuclides in plants, animals, and soil (Chapter 7, Ecosystem Health); and finally, what radionuclide dose or risk from chemical exposure members of the public may experience as a result of Laboratory operations (Chapter 8, Public Dose and Risk Assessment).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2020

Vadose Zone Model for B Plant Area for Composite Analysis

The objectives of the vadose modeling for the updated Hanford Site composite analysis (CA) are to simulate the flow and transport of water and radionuclide releases from the surface to the water table and to provide radionuclide transfer rates for the plateau to river (P2R) model, version 8.3. Water additions include natural recharge and water discharged to the ground as a result of industrial processes associated with Hanford Site operations. Contaminant sources include radionuclides in water discharged to the ground during operations and radionuclides disposed “dry” in solid waste burial grounds or other means. The simulation time starts in 1943 and ends at 12070, which is 10,000 years after assumed Hanford Site closure in 2070. The scope of this ECF is to document the development and results of the B Plant Area vadose zone model. The B Plant Area model encompasses the region around B Plant (221-B) in the western part of 200 East Area. The plant was used to separate plutonium from irradiated fuel using the bismuth-phosphate process from 1945 to 1956, and for strontium and cesium recovery from 1968 to 1985.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2020

Vadose Zone Model for B-3 Pond for Composite Analysis

The objectives of the vadose modeling for the updated Hanford Site composite analysis (CA) are to simulate the flow and transport of water and radionuclide releases from the surface to the water table and to provide radionuclide transfer rates for the plateau to river (P2R) model, version 8.3. Water additions include natural recharge and water discharged to the ground as a result of industrial processes associated with Hanford Site operations. Contaminant sources include radionuclides in water discharged to the ground during operations and radionuclides disposed “dry” in solid waste burial grounds or other means. The simulation time starts in 1943 and ends at 12070, which is 10,000 years after assumed Hanford Site closure in 2070. The scope of this ECF is to document the development and results of the B-3 Pond vadose zone model. The B-3 Pond model covers the main lobe of the 216-B-3 Pond System located in a topographic low east of the 200 East Area. The pond system operated from 1945 to 1997 and received effluents from the Plutonium Uranium Extraction Plant, B Plant, 284-E Powerhouse, and other 200 East Area facilities.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2020

Vadose Zone Model for B-3C Pond for Composite Analysis

The objectives of the vadose modeling for the updated Hanford Site composite analysis (CA) are to simulate the flow and transport of water and radionuclide releases from the surface to the water table and to provide radionuclide transfer rates for the plateau to river (P2R) model, version 8.3. Water additions include natural recharge and water discharged to the ground as a result of industrial processes associated with Hanford Site operations. Contaminant sources include radionuclides in water discharged to the ground during operations and radionuclides disposed “dry” in solid waste burial grounds or other means. The simulation time starts in 1943 and ends at 12070, which is 10,000 years after assumed Hanford Site closure in 2070. The scope of this ECF is to document the development and results of the B-3C Pond vadose zone model. The B-3C Pond model covers the 216-B-3C Expansion Pond of the 216-B-3 Pond System located in a topographic low east of the 200 East Area. The pond system operated from 1945 to 1997 and received effluents from the Plutonium Uranium Extraction Plant, B Plant, 284-E Powerhouse, and other 200 East Area facilities.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗