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At least 19 records

Tritium Fires: Simulation and Safety Assessment

This is the Sandia report from a joint NSRD project between Sandia National Labs and Savannah River National Labs. The project involved development of simulation tools and data intended to be useful for tritium operations safety assessment. Tritium is a synthetic isotope of hydrogen that has a limited lifetime, and it is found at many tritium facilities in the form of elemental gas (T 2 ). The most serious risk of reasonable probability in an accident scenario is when the tritium is released and reacts with oxygen to form a water molecule, which is subsequently absorbed into the human body. This tritium oxide is more readily absorbed by the body and therefore represents a limiting factor for safety analysis. The abnormal condition of a fire may result in conversion of the safer T 2 inventory to the more hazardous oxidized form. It is this risk that tends to govern the safety protocols. Tritium fire datasets do not exist, so prescriptive safety guidance is largely conservative and reliant on means other than testing to formulate guidelines. This can have a consequence in terms of expensive and/or unnecessary mitigation design, handling protocols, and operational activities. This issue can be addressed through added studies on the behavior of tritium under representative conditions. Due to the hazards associated with the tests, this is being approached mainly from a modeling and simulation standpoint and surrogate testing. This study largely establishes the capability to generate simulation predictions with sufficiently credible characteristics to be accepted for safety guidelines as a surrogate for actual data through a variety of testing and modeling activities.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

SRS Small Modular Reactors

The small modular reactor program at the Savannah River Site and the Savannah River National Laboratory.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Disposition of Chicago Pile 5 (CP-5) Converter Tubes in the 10-160B Cask - 16632

This paper will focus on the unique characterization, packaging, and transportation issues associated with the disposition of the two CP-5 Converter Tube assemblies from Argonne National Laboratory. The converter tubes were constructed of combinations of HEU and alloys of zirconium, and were part of the original research facilities attached to the CP-5 reactor during operating evolutions. These assemblies were heavily irradiated during their operational lifetime, and were segregated from the balance of irradiated test specimens when the reactor was deactivated and slated for Decontamination and Demolition (D and D). In addition, the substantial contribution of fissile material to the assemblies' inventory made the potential disposition pathways extremely challenging. As a result, these items became part of Argonne's legacy 'nuclear footprint', and were added to the Nuclear Footprint Reduction Project scope for disposition. The Project was responsible for the size reduction and characterization of these items, as well as the ultimate disposition. After negotiating a disposal pathway for these tubes, there were significant transportation issues that required a small team to overcome, in order to successfully ship these items to the Nevada National Security Site (NNSS). The Project team at Argonne, technical support from transportation specialists, licensing support from the 10-160B license owner, the Savannah River National Lab (SRNL) Packaging Certification Team (PCT, and the DOE EM-33 staff contributed to license and safety analysis report amendments that eventually authorized the shipment of the material. The paper will identify the organizations, and the specific actions, required to successfully make three 'one of a kind' shipments of irradiated test specimen material. This will include the unique packaging configurations, contents modification for the cask license (via the Amendment process), criticality evaluations, and associated review and approval processes. (authors)

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Characterization Results for the October 2024 Tank Farm 3H Evaporator Overhead Sample

On an annual basis, Savannah River Mission Completion (SRMC) provides 2H and 3H evaporator overhead samples to Savannah River National Lab (SRNL) to be analyzed per Section 5.2 of the Effluent Treatment Project (ETP) Waste Compliance Plan (WCP) and the Waste Acceptance Criteria (WAC). This report presents characterization results for the October 2024 3H evaporator overhead sample. The sample was clear and colorless with no visible solids. The results provide measurements for cesium-137 (137Cs), strontium-90 (90Sr), and iodine-129 (129I) with the radionuclide concentration limits specified by the WAC. These analyses were performed in duplicate, and a summary of the analytical results for this 3H evaporator overhead sample includes the following: The measured cesium-137 activity in the 3H evaporator overhead sample averaged 6.72E+01 dpm/mL, (3.90E+00 %RSD), which is below the ETP WAC limit of 1.30E+03 dpm/mL. The strontium-90 activity in the 3H evaporator overhead sample averaged <2.72E+00 dpm/mL, which is below the ETP WAC limit of 1.76E+02 dpm/mL. The iodine-129 activity in the 3H evaporator overhead sample averaged <1.63E-01 dpm/mL, which is below the ETP WAC limit of 1.00E+00 dpm/mL.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Characterization Results for the August 2025 Tank Farm 2H Evaporator Overhead Sample

On an annual basis, Savannah River Mission Completion (SRMC) provides 2H and 3H evaporator overhead samples to Savannah River National Lab (SRNL) to be analyzed per Section 5.2 of the Effluent Treatment Project (ETP) Waste Compliance Plan (WCP) and the Waste Acceptance Criteria (WAC). This report presents the average characterization results for the August 2025 2H evaporator overhead sample. The sample was clear and colorless with no visible solids. The results provide measurements for cesium-137 ( 137 Cs), strontium-90 ( 90 Sr), and iodine-129 ( 129 I) with the radionuclide concentration limits specified by the WAC. These analyses were performed in duplicate, and all three measured radionuclide concentrations were within ETP WAC limits. A summary of the analytical results for this 2H evaporator overhead sample includes the following: The measured cesium-137 activity in the 2H evaporator overhead sample averaged 3.02E+02 dpm/mL, which is below the ETP WAC limit of 1.30E+03 dpm/mL. The strontium-90 activity in the 2H evaporator overhead sample averaged 3.33E+00 dpm/mL, which is below the ETP WAC limit of 1.76E+02 dpm/mL. The iodine-129 activity in the 2H evaporator overhead sample averaged 1.20E-01 dpm/mL, which is below the ETP WAC limit of 1.00E+00 dpm/mL. Although the measurements for Cs-137 and Sr-90 are lower compared to previous years, the results remain within a similar range to past measurements and are well below the WAC limits. Iodine-129 indicates detectable results compared to previous years, which can be attributed to recent salt batch processing trending close to the WAC limit for saltstone production. However, these results are still below the ETP WAC limit.

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Chemical and Radiological Compatibility Testing of 3D Printed Materials

To dramatically increase the adaptability, performance, and safety of processes in support of the Defense Waste Processing Facility (DWPF), Savannah River National Lab (SRNL) plans to perform chemical and radiological compatibility testing on a wide variety of 3D printed materials of interest. The 3D printing process provides numerous strategic operational benefits such as rapid prototyping of complex designs and geometry specific to the needs of the nuclear waste disposition process, as well as on-demand rapid prototyping and iteration with materials that aren’t as accessible through traditional manufacturing methods. Reaction chemistry in simulated waste batches can be matched closely to its radioactive counterpart, but glass reactor vessels have limitations. Vessel geometry can play a big factor in mixing transport limitations, process chemistry, and degradation reaction kinetics. In addition, additive manufacturing allows for much more detailed vessel design than traditional alternatives. Waste processing techniques in DWPF also encounter extreme chemical environments including high pH, strong acids, abrasive slurries, and significant irradiation. To meet these challenges, a matrix of various polymer, ceramic, and metal additive manufacturing materials have been exposed to a suite of chemical environments of interest as well as radioactive dose (such as gamma radiation from 60 Co) to properly test their durability under these conditions. Mass change has been monitored over a period of up to a week in these conditions, as well as added characterization for surface modification through Scanning Electron Microscopy/Electron Dispersive X-ray analysis (SEM/EDX). Further chemical characterization has been monitored through Fourier-Transform InfraRed Spectroscopy (FTIR), with planned investigation via thermal and tensile strength degradation. While the direct product of this research is identification of material(s) that can withstand specific hazardous environments encountered by the mercury water wash tank in DWPF process simulation experiments, the reference base of materials will be used for many other nuclear processes in the pursuit of rapidly developed, cost-efficient, and highly specific devices for environmental remediation and much more.

Wilson, Nathan W. [Savannah River National Laborat↗

Revised Guidelines for using Cellulose Degradation Product-Impacted Kd Values for Performance Assessments and Composite Analyses

Cellulosic materials include wood, paper, rags, and cardboard products. These materials are co-disposed with radiological waste at the Savannah River Site's (SRS) E-Area Low-Level Waste Facility (ELLWF). Cellulosic materials readily degrade in the environment to form cellulose degradation products (CDP) that will partition to the sediment or remain mobile in the groundwater. Savannah River National Lab (SRNL) has conducted studies to estimate the impact of CDP on radionuclide sorption to SRS sediments (Kd values). It was found that CDP impact on radionuclide sorption varies with radionuclide and CDP concentration. Furthermore, it was found that the amount of carbon (C) in the system could increase or decrease Kd values with respect to the base case of when no CDP was added. Throughout the expected pH range of the ELLWF, a low concentration of CDP in the system would increase Kd values (because C would sorb to the sediment and provide more exchange sites for radionuclides to sorb), whereas greater concentrations of CDP (≥20 mg/L C) would decrease Kd values (because C would remain in solution and complex the radionuclide and not permit the radionuclide to sorb to the sediment). A review of >230 dissolved organic carbon (DOC) groundwater concentrations in the Old Radioactive Waste Burial Ground (ORWBG) at the SRS indicated that the average DOC concentration, a gross measure of CDP, was 5 mg/L C. At approximately this DOC concentration, the laboratory studies demonstrated that no anions (Tc, I, or Se) or cations (Ni, Sr, Ce, Eu, Zr, or Th) have decreased sorption in the presence of carbon (an analogue for CDP).

09 BIOMASS FUELS↗

Rapid determination of 226Ra in environmental samples

A new rapid method for the determination of {sup 228}Ra in natural water samples has been developed at the SRNL/EBL (Savannah River National Lab/ Environmental Bioassay Laboratory) that can be used for emergency response or routine samples. While gamma spectrometry can be employed with sufficient detection limits to determine {sup 228}Ra in solid samples (via {sup 228}Ac) , radiochemical methods that employ gas flow proportional counting techniques typically provide lower MDA (Minimal Detectable Activity) levels for the determination of {sup 228}Ra in water samples. Most radiochemical methods for {sup 228}Ra collect and purify {sup 228}Ra and allow for {sup 228}Ac daughter ingrowth for ~36 hours. In this new SRNL/EBL approach, {sup 228}Ac is collected and purified from the water sample without waiting to eliminate this delay. The sample preparation requires only about 4 hours so that {sup 228}Ra assay results on water samples can be achieved in < 6 hours. The method uses a rapid calcium carbonate precipitation enhanced with a small amount of phosphate added to enhance chemical yields (typically >90%), followed by rapid cation exchange removal of calcium. Lead, bismuth, uranium, thorium and protactinium isotopes are also removed by the cation exchange separation. {sup 228}Ac is eluted from the cation resin directly onto a DGA Resin cartridge attached to the bottom of the cation column to purify {sup 228}Ac. DGA Resin also removes lead and bismuth isotopes, along with Sr isotopes and {sup 90}Y. La is used to determine {sup 228}Ac chemical yield via ICP-MS, but {sup 133}Ba can also be used instead if ICP-MS assay is not available. Unlike some older methods, no lead or strontium holdback carriers or continual readjustment of sample pH is required.

54 ENVIRONMENTAL SCIENCES↗

A reduced order sulfuric acid decomposition model for a nuclear-powered hybrid sulfur cycle

A reduced order model of sulfuric acid decomposition within a bayonet chemical reactor was developed to support the U. S. Department of Energy Integrated Energy System program, and address the lack of knowledge in scaling and integration for joint chemical and nuclear processes. Sulfuric acid decomposition within a bayonet reactor was modeled to provide chemical and thermodynamic data relevant to advanced nuclear reactor-driven integrated energy systems based on desired operational scale and operational conditions. The temperature range required for high-temperature advanced nuclear reactor integrated energy systems, 750-850 °C, was shown to produce reasonable agreement (within a few percent relative error) with past models and experimental data, and yielded good efficiency results for bayonet reactor operations. The results of the reduced order model agreed with previous work from Savannah River National Labs within a maximum of 3.4% error on the decomposition of sulfur trioxide, and on previous Hybrid Sulfur flowsheets from Gorensek and Summers that showed operational temperature, pressure, and composition ranges for efficiency which made the Hybrid Sulfur cycle competitive with water electrolysis. Here, the agreement with previous high-fidelity models provided a framework for future Integrated Energy System grid evaluations with an advanced nuclear reactor and large-scale hydrogen production using a mathematical model to represent chemical operations.

08 HYDROGEN↗

Solvent Hold Tank Sample Results for MCU-15-661-662-663: April 2015 Monthly Sample

The Savannah River National Lab (SRNL) received one set of Solvent Hold Tank (SHT) samples (MCU-15-661, MCU-15-662, and MCU-15-663 pulled on April 2, 2015) for analysis. The samples were combined and analyzed for composition. Analysis of the composite sample MCU-15-661-662-663 indicated a low concentration (~ 63% of nominal) of the suppressor (TiDG) and a slightly below the nominal concentration (~ 10% below nominal) of the extractant (MaxCalix). The modifier (CS-7SB) level was also 10% below its nominal value while the Isopar™ L level was slightly above its nominal value. This analysis confirms the addition of Isopar™L to the solvent on March 6, 2015. Despite that the values are below target component levels, the current levels of TiDG, CS-7SB and MaxCalix are sufficient for continuing operation without adding a trim at this time until the next monthly sample. No impurities above the 1000 ppm level were found in this solvent. However, the sample was found to contain approximately 18.4 ug/g solvent mercury. The gamma level increased to 8 E5 dpm/mL solvent and it represents an order of magnitude increase relative to previous solvent samples. The increase means less cesium is being stripped from the solvent. Further analysis is needed to determine if the recent spike in the gamma measurement is due to external factors such as algae or other material that may impede stripping. The laboratory will continue to monitor the quality of the solvent in particular for any new impurity or degradation of the solvent components.

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Solvent Hold Tank Sample Results for MCU-16-270-271-272: February 2016 Monthly Sample

Savannah River National Lab (SRNL) received one set of Solvent Hold Tank (SHT) samples (MCU-16-270-271-272), pulled on 02/21/2016 for analysis. The samples were combined and analyzed for composition. Analysis of the composite sample MCU-16-270-271-272 indicated the Isopar TM L concentration is above its nominal level (101%). The modifier (CS-7SB), the extractant (MaxCalix), and the TiDG concentrations are 7%, 6%, and 40% below their nominal concentrations. This analysis confirms the addition of TiDG, MaxCalix, and modifier to the solvent in November 2015. Based on the current monthly sample, the levels of TiDG, Isopar TM L, MaxCalix, and modifier are sufficient for continuing operation but are expected to decrease with time. Periodic characterization and trimming additions to the solvent are recommended.

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Gamma imaging and mapping advancements

The Savannah River National Lab (SRNL) in partnership with Los Alamos National Lab (LANL) developed an in-situ GrayQb™ radiation mapping device eliminating the need to remove the radiation sensitive phosphor storage plate (PSP) from the device for scanning between exposures. By incorporating a customized version of the LANL patented MiniMax camera-based PSP reader system into the GrayQb™ housing, the exposed PSP can now be read in-situ and digitally transmitted to a remote computer for viewing. The ability to generate radiation contour maps showing source locations and relative radiological levels present in an area has high utility for targeting clean-up efforts, maintaining worker ALARA, determining leak locations and radiological situational awareness. In-situ GrayQb™ reduces the number of required interactions with the device providing deployment cost savings, increasing personnel safety and removing human handling errors during the manual scan process. Additionally, by negating the need for an external scanner, the device becomes more mobile. Other upgrades to the device included identification of a more sensitive imaging medium and incorporation of a dual pinhole rendering the device physically capable of obtaining 3D images. Laboratory tests with Am-241 sources using a mockup of the in-situ GrayQ™, successfully demonstrated the ability to digitally capture radiation source information using the dual pinhole and camera-based PSP reader.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

MULTI-PRED: A Software Module for Predictive Modeling of Coupled Multi-Physics Systems: User's Manual

This User’s Manual describes the code module MULTI-PRED, written in FORTRAN which implements the methodology for “predictive modeling of coupled multi-physics systems (PM-CMPS)” formulated by Cacuci (2014). This methodology fully takes into account the coupling terms between the systems but requires only the computational resources that would be needed to perform predictive modeling on each system separately. The PM-CMPS methodology uses the maximum entropy principle to construct an optimal approximation of the unknown a priori distribution based on a priori known mean values and uncertainties characterizing the experimental and computational parameters and results of interest responses, called for the multi-physics models under consideration. This “maximum entropy” a priori distribution is combined, using Bayes’ theorem, with the “likelihood” provided by the multi-physics simulation models to obtain a formal posterior distribution. Subsequently, the posterior distribution thus obtained is evaluated using the saddle-point method to obtain analytical expressions for the optimally predicted values for the multi-physics models parameters and responses along with corresponding reduced uncertainties. Noteworthy, the predictive modeling methodology for the coupled systems is constructed such that the systems can be considered sequentially rather than simultaneously, while preserving exactly the same results as if the systems were treated simultaneously. Consequently, very large coupled systems, which could perhaps exceed available computational resources if treated simultaneously, can be treated with the PM-CMPS methodology presented in this work sequentially and without any loss of generality or information, requiring just the resources that would be needed if the systems were treated sequentially. Three illustrative demonstration problems are also provided. The first problem presents the application of the PM-CMPS methodology to a simple particle diffusion problem which admits a closed-form analytical solution which facilitates a rapid understanding of this methodology and its predicted results. The second demonstration problem presents the application of the PM-CMPS methodology to the problem of inverse prediction, from detector responses in the presence of counting uncertainties, of the thickness of a homogeneous slab of material containing uniformly distributed gamma-emitting sources, for optically thin and thick slabs. This problem highlights the essential role played by the relative uncertainties (or, conversely, accuracies) of measured and computed responses. The third demonstration problem presents the application of the PM-CMPS methodology to the F-area cooling towers at the Savannah River National Lab. This problem demonstrates that the PM-CMPS methodology reduces the predicted response uncertainties not only at locations where measurements are available, but also at locations where measurements are not available.

97 MATHEMATICS AND COMPUTING↗

Phase separation and crystallization of complex borosilicate melts for glass-ceramic waste forms

Waste loadings of reprocessed spent nuclear fuel vitrified into borosilicate glass can be increased by precipitating environmentally stable phases concentrated with waste components in a chemically stable glass matrix. The principal objective of this study was to characterize the development of crystalline powellite (CaMoO 4 and related phases) and oxyapatite (Ca 2 LN 8 Si 6 O 26 ) in borosilicate glass-ceramics and to determine how the formation of those phases affected its chemical durability. Borosilicate glasses provided by PNNL were re-melted and quenched at rates from over 300°C/s to ~0.05°C/s. Isothermal heat treatment experiments were conducted by quenching melts in a molten tin bath at various temperatures, holding for periods of time, and then quenching in a water bath. Analytical electron microscopy and x-ray diffraction provided information about the kinetics of the phase separation and crystallization processes responsible for microstructural development. Powellite and oxyapatite crystals formed during slower quench rates and longer isothermal times, and time-temperature-transformation (TTT) diagrams were developed from the latter experiments. A hot thermocouple test (HTT) system was built to rapidly quench melts and hold them at set times to provide supplementary TTT information, but that system was not as sensitive to the phase transformations as other tests. Corrosion tests were performed to understand how the individual phases in the glass-ceramic affect its overall chemical durability. Product consistency tests, performed at Missouri S&T and at Savannah River National Lab, provided release rates of major elements from samples as a function of cooling rate, and atomic force microscopy and profilometry measurements of surface topology determined the relative corrosion rates of the residual glass and oxyapatite phases. Faster dissolution rates were measured from samples cooled more slowly and these were explained by the greater fractions of B 2 O 3 in the residual glass phase after the formation of oxyapatite and powellite. Oxyapatite was found to be more durable than the residual glass.

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Tris(isodecyl)guanidine Degradation in the MCU System

The current solvent blend consists of four components; an extractant, the modifier, a suppressor, and the diluent. Of the four components, only the suppressor – tris(isodecyl)guanidine (TiDG) has exhibited an appreciable depletion rate during facility operations. Using data derived from Modular Caustic-Side Solvent Extraction Unit (MCU) process samples, Savannah River National Lab (SRNL) derived a method to predict the TiDG depletion based upon time and volume of feed processed.

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Fork Union EPC Installation Observation Report

On June 22nd and 23rd, a representative from Savannah River National Lab (SRNL) traveled to the Fork Union substation in Fork Union, Virginia (owned and operated by Dominion Energy) to observe the installation, calibration, and commissioning of three Electric Phenomena Cluster (EPC) sensors. The installation in its entirety included the physical securing of the EPC sensors by Dominion Energy, fiber optic cable (furnished by SmartSenseComm) being run in conduits and cable trays from the EPC sensor to the Optical Processing Unit (OPU) located in the substation switch house. The location of the EPC sensors, OPU, and cable routing can be seen in the figures included.

47 OTHER INSTRUMENTATION↗

Characterization Results for the March 2022 Tank Farm 3H Evaporator Overhead Sample

On an annual basis, Savannah River Mission Completion (SRMC) provides 2H and 3H evaporator overhead samples to Savannah River National Lab (SRNL) to be analyzed per Section 5.2 of the Effluent Treatment Project (ETP) Waste Compliance Plan (WCP) and the Waste Acceptance Criteria (WAC). This report presents characterization results for the March 2022 3H evaporator overhead sample. The sample was clear and colorless with no visible solids. The results provide measurements for cesium-137 ( 137 Cs), strontium-90 ( 90 Sr), and iodine-129 ( 129 I) with the radionuclide concentration limits specified by the WAC. These analyses were performed on a single sample replicate. A summary of the analytical results for the 3H evaporator overhead sample includes the following. The measured cesium-137 activity in the 3H evaporator overhead sample is 4.59E+02 dpm/mL which is above the ETP WAC limit of 3.28E+02 dpm/mL. The strontium-90 activity in the 3H evaporator overhead sample is <2.95E+01 dpm/mL. This is below the ETP WAC limit of 1.76E+02 dpm/mL. The iodine-129 activity in the 3H evaporator overhead sample is <1.92E-02 dpm/mL. This is below the ETP WAC limit of 1.00E+00 dpm/mL.

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Characterization Results for the April 2022 Tank Farm 2H Evaporator Overhead Sample

On an annual basis, Savannah River Mission Completion (SRMC) provides 2H and 3H evaporator overhead samples to Savannah River National Lab (SRNL) to be analyzed per Section 5.2 of the Effluent Treatment Project (ETP) Waste Compliance Plan (WCP) and the Waste Acceptance Criteria (WAC). This report presents characterization results for the April 2022 2H evaporator overhead sample. The sample was clear and colorless with no visible solids. The results provide measurements for cesium-137 ( 137 Cs), strontium-90 ( 90 Sr), and iodine-129 ( 129 I) with the radionuclide concentration limits specified by the WAC.

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