Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “Tanks”

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.

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

At least 523 records · Page 29
All sourcesNASA NTRSDOE OSTI
DOE OSTI2012

Tank 40 Final SB7b Chemical Characterization Results

A sample of Sludge Batch 7b (SB7b) was taken from Tank 40 in order to obtain radionuclide inventory analyses necessary for compliance with the Waste Acceptance Product Specifications (WAPS). The SB7b WAPS sample was also analyzed for chemical composition including noble metals and fissile constituents, and these results are reported here. These analyses along with the WAPS radionuclide analyses will help define the composition of the sludge in Tank 40 that is currently being fed to the Defense Waste Processing Facility (DWPF) as SB7b. At the Savannah River National Laboratory (SRNL) the 3-L Tank 40 SB7b sample was transferred from the shipping container into a 4-L high density polyethylene bottle and solids were allowed to settle over the weekend. Supernate was then siphoned off and circulated through the shipping container to complete the transfer of the sample. Following thorough mixing of the 3-L sample, a 558 g sub-sample was removed. This sub-sample was then utilized for all subsequent analytical samples. Eight separate aliquots of the slurry were digested, four with HNO 3 /HCl (aqua regia) in sealed Teflon® vessels and four with NaOH/Na 2 O 2 (alkali or peroxide fusion) using Zr crucibles. Two Analytical Reference Glass - 1 (ARG-1) standards were digested along with a blank for each preparation. Each aqua regia digestion and blank was diluted to 1:100 mL with deionized water and submitted to Analytical Development (AD) for inductively coupled plasma - atomic emission spectroscopy (ICP-AES) analysis, inductively coupled plasma - mass spectrometry (ICP-MS) analysis, atomic absorption spectroscopy (AA) for As and Se, and cold vapor atomic absorption spectroscopy (CV-AA) for Hg. Equivalent dilutions of the alkali fusion digestions and blank were submitted to AD for ICP-AES analysis. Tank 40 SB7b supernate was collected from a mixed slurry sample in the SRNL Shielded Cells and submitted to AD for ICP-AES, ion chromatography (IC), total base/free OH - /other base, total inorganic carbon/total organic carbon (TIC/TOC) analyses, and Cs-137 gamma scan. Weighted dilutions of slurry were submitted for IC, TIC/TOC, and total base/free OH - /other base analyses. Activities for U-233, U-235, and Pu-239 were determined from the ICP-MS data for the aqua regia digestions of the Tank 40 WAPS slurry using the specific activity of each isotope. The Pu-241 value was determined from a Pu-238/-241 method developed by SRNL AD and previously described. The following conclusions were drawn from the analytical results reported here: (1) The ratios of the major elements for the SB7b WAPS sample are different from those measured for the SB7a WAPS sample. There is less Al and Mn relative to Fe than the previous sludge batch. (2) The elemental composition of this sample and the analyses conducted here are reasonable and consistent with DWPF batch data measurements in light of DWPF pre-sample concentration and SRAT product heel contributions to the DWPF SRAT receipt analyses. The element ratios for Al/Fe, Ca/Fe, Mn/Fe, and U/Fe agree within 10% between this work and the DWPF Sludge Receipt and Adjustment Tank (SRAT) receipt analyses. (3) Sulfur in the SB7b WAPS sample is 82% soluble, slightly less than results reported for SB3, SB4, and SB6 samples but unlike the 50% insoluble sulfur observed in the SB5 WAPS sample. In addition, 23% of the soluble sulfur is not present as sulfate in SB7b. (4) The average activities of the fissile isotopes of interest in the SB7b WAPS sample are (in μCi/g of total dried solids): 4.22E-02 U-233, 6.12E-04 U-235, 1.08E+01 Pu-239, and 5.09E+01 Pu-241. The full radionuclide composition will be reported in a future document. (5) The fission product noble metal and Ag concentrations appear to have largely peaked in previous DWPF sludge batches, with the exception of Ru, which still shows a slight increase in SB7b.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

Analyses of HTF-48-12-20/24 (February, 2012) and Archived HTF-E-05-021 Tank 48H Slurry Samples

Personnel characterized a Savannah River National Laboratory (SRNL) archived sample of Tank 48H slurry (HTF-E-05-021) in addition to the composite of samples HTF-48-12-20 and HTF-48-12-24, which were both retrieved in February 2012. The combined February 2012 sample is referred to as HTF-48-12-20/24 in this report. The results from these analyses are compared with Tank 48H samples analyzed in 2003, 2004, and 2005. This work supports the effort to demonstrate copper-catalyzed peroxide oxidation (CCPO) of organic content in this material. The principal findings with respect to the chemical and physical characteristics of the most recent sample are: (1) The measured potassium tetraphenylborate (KTPB) solid concentration is 1.76 wt %; (2) Titanium was in line with 2004 and 2005 slurry measurements at 897 mg/L, it represents 0.1535 ± 0.0012 wt % monosodium titanate (MST); (3) The measured insoluble solids content was 1.467 wt %; (4) The free hydroxide concentration in the Tank 48H filtrate sample (1.02 ± 0.02 M) is close to the Tank 48H limit (1.0 M); (5) Carbonate reported by total inorganic carbon (TIC, 1.39 ± 0.03 M) is more than double the concentrations measured in past (2003-2005) samples; (6) The soluble potassium content (measured at 286 ± 23 mg/L) in the filtrate is in line with all past measurements; and (7) The measured 137 Cs concentration is 7.81E + 08 ± 3.9E + 07 dpm/mL of slurry (1.33 ± 5% Ci/gallon or 3.18E + 05 ± 5% curies of 137 Cs in the tank) in the slurry which is in agreement with the 2005 report of 3.14E + 05 ± 1.5% curies of 137 Cs in the tank. The filtrate 137 Cs concentration is 2.57E + 07 ± 2.6E + 05 dpm/mL. This result is consistent with previous results. Significant analytical data are summarized in Table 1.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Determining the Release of Radionuclides from Tank Waste Residual Solids: FY2015 Report

Methodology development for pore water leaching studies has been continued to support Savannah River Site High Level Waste tank closure efforts. For FY2015, the primary goal of this testing was the achievement of target pH and Eh values for pore water solutions representative of local groundwater in the presence of grout or grout-representative (CaCO 3 or FeS) solids as well as waste surrogate solids representative of residual solids expected to be present in a closed tank. For oxidizing conditions representative of a closed tank after aging, a focus was placed on using solid phases believed to be controlling pH and E h at equilibrium conditions. For three pore water conditions (shown below), the target pH values were achieved to within 0.5 pH units. Tank 18 residual surrogate solids leaching studies were conducted over an E h range of approximately 630 mV. Significantly higher Eh values were achieved for the oxidizing conditions (ORII and ORIII) than were previously observed. For the ORII condition, the target Eh value was nearly achieved (within 50 mV). However, E h values observed for the ORIII condition were approximately 160 mV less positive than the target. E h values observed for the RRII condition were approximately 370 mV less negative than the target. Achievement of more positive and more negative E h values is believed to require the addition of non-representative oxidants and reductants, respectively. Plutonium and uranium concentrations measured during Tank 18 residual surrogate solids leaching studies under these conditions (shown below) followed the general trends predicted for plutonium and uranium oxide phases, assuming equilibrium with dissolved oxygen. The highest plutonium and uranium concentrations were observed for the ORIII condition and the lowest concentrations were observed for the RRII condition. Based on these results, it is recommended that these test methodologies be used to conduct leaching studies with actual Tank 18 residual solids material. Actual waste testing will include leaching evaluations of technetium and neptunium, as well as plutonium and uranium.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

CST Batch Contact Equilibrium Testing of TCCR Tank 10H Batch 1 and 1A Waste Supernate Samples

The Tank Closure Cesium Removal (TCCR) system uses ion exchange columns filled with crystalline silicotitanate media to process radioactive waste solutions for the removal of Cs-137. TCCR currently focuses on dissolving Savannah River Site (SRS) Tank 10H waste (primarily sodium saltcake solids) within the tank followed by at-tank ion exchange column treatment. Equilibrium batch contact tests conducted in the SRNL Shielded Cells laboratory and associated analyses indicated that both the Tank 10H Batch 1 (pre-production) and 1A (production) waste supernates are unique solutions for which CST cesium removal performance is lower than predicted by ZAM modeling. Cesium distribution coefficients (K d ), percent removal, and CST loading data for the TCCR production supernate batch are provided in Table ES-1. ZAM-based cesium loading predictions are compared to the observed loading for the production batch in Figure ES-1, where significantly lower performance was observed than predicted by the model. Even lower cesium loading results (versus ZAM) were observed previously for the in-tank “teabag” batch contact samples.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Hydrogen Generation Rates for Saltstone Related Samples with Tank 50 Waste Grout and Glycoate Using Sealed Reactors

This report documents the results of sealed reactor tests to measure the hydrogen generation rate (HGR) from Tank 50 waste, which may be on the order of 3 x 10 -9 (ft 3 /h)/gal. Because the estimated HGR limit of detection (LOD) for an open, continuous flow system is approximately 6 x 10 -8 (ft 3 /h)/gal a more sensitive method was needed. By accumulating generated gases in a sealed system without a continuous purge gas, the resulting concentrations can increase the sensitivity of measurement, which is based on duration of heating. Prior sealed-reactor testing demonstrated measurements down to 7 x 10 -10 (ft 3 /h)/gal. In 2017, Savannah River Remediation, LLC (SRR) declared a Potential Inadequacy in the Safety Analysis (PISA) for the Saltstone Production Facility (SPF), among other facilities. The research for this report concerns waste from Tank 50 which is processed through SPF to produce saltstone. Previous studies examined the thermolysis of Tank 50 waste mixed with small individual amounts of saltstone grout premix additives, as well as glycolate. In contrast, the current study evaluated the impact of Tank 50 waste containing glycolate and stabilized within saltstone, on the generation of flammable gases, specifically hydrogen. The task included holding at target temperatures Tank 50 samples of salt solution in grout and spiked with 350 mg/L of glycolate to determine if the glycolate decomposes into H 2 . The measured hydrogen generation rates for radioactive saltstone monoliths over 28 days are: 2.0 x 10 -7 (ft 3 /h)/gal at 64-68°C, 1.4 x 10 -7 (ft 3 /h)/gal at 80-82°C, and 4.5 x 10 -7 (ft 3 /h)/gal at 112°C. Testing also included a non-radioactive salt simulant of Tank 50 waste, which was spiked with a concentration of glycolate expected in Savannah River Site (SRS) low-level waste, after the waste is solidified in grout. Included in the simulant tests was using grout made with two different surface areas, a monolith form with only the upper surface exposed and a pellet, or cube, form to increase the exposed surface area, to demonstrate the importance of exposed area in H 2 generation.

08 HYDROGEN↗
DOE OSTI2020

Impacts of the Addition of Sodium Reactor Experiment (SRE) and DR-3 Fuel from H-Canyon to Tank 40 on Acid Stoichiometry and Hydrogen Production

H-Canyon plans to discard Sodium Reactor Experiment (SRE) material that is currently stored in Tanks 16.3 and 16.4. Savannah River Remediation (SRR) determined that a portion of the SRE material can be added to Sludge Batch 9 (SB9) in Tank 40 and the remainder will be added to Sludge Batch 10 (SB10) that is currently being prepared in Tank 51. A processing flowsheet for both the nitric-formic acid flowsheet and the nitric-glycolic acid flowsheet were previously developed for SB9 sludge-only and coupled operation with the Actinide Removal Process – Modular Caustic Side Solvent Extraction Unit (ARP-MCU) and the Salt Waste Processing Facility (SWPF). Due to the compositional change in Tank 40 after the addition of SRE material, it is necessary to determine the influence on the acid stoichiometry operating window and hydrogen production for SB9 processing. An assessment was completed to address the influence of an addition of SRE and DR-3 Fuel from H-Canyon on the chemistry of Tank 40, the defined acid stoichiometry window (with and without SWPF streams) for the nitric-formic and nitric-glycolic acid flowsheets, and the resulting influence of the hydrogen production. Hydrogen production during the SRAT/SME cycles consists of thermolytic, radiolytic, and catalytic hydrogen. For the nitric-formic acid flowsheet, catalytic hydrogen dominates the hydrogen production during DWPF operation mode for the SRAT and SME vessels. This report was written to summarize that assessment. No SRNL testing is needed prior to adding up to 7,100 additional gallons of SRE and DR-3 Fuel from H-Canyon to Tank 40. Also, the addition of SRE will not exceed the Shielded Cells reported values for catalytic hydrogen for either the nitric-formic acid flowsheet (Technical Safety Requirement (TSR) limit of 0.15 lb/h) or the nitric-glycolic acid flowsheet (TSR limit 0.024 lb/h). The addition of the ARP stream from SWPF and the use of the nitric-glycolic acid flowsheet are both expected to decrease hydrogen generation, leading to a larger safety margin in the CPC.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Glycolate Analysis in Tank 22: Developing and Testing Analytical Methods for the Savannah River Site Liquid Waste System

Researchers developed and tested a range of techniques to support a low mg/L Limit of Quantitation (LOQ) of glycolate in radioactive Tank 22 waste solution by Ion Chromatography (IC) and by proton NMR. For the IC method, Dionex OnGuard II cartridges were tested as a means of removing alkali earth and transition metals that can interfere during ion chromatography applications especially when analyzing for low-concentration, chelating analytes. Additionally, concentrations of nitrate in the raw Tank 22 sample (5,000 to 10,000 mg/L) were managed by using reasonable levels of sample dilution. The resulting IC performance quality was enhanced by improving the baseline, increasing sensitivity [Limit of Quantitation (LOQ) of 12 mg/L and Limit of Detection (LOD) of 4.0 mg/L], and resolving analytes into well-defined Gaussian peaks when using the Dionex OnGuard II H⁺ cartridges to remove matrix interferences. High concentrations of nitrate limited the performance of the IC method. To achieve an acceptable baseline, samples with high nitrate content require more dilution, resulting in higher detection limits. Tests on samples with higher concentrations of nitrate (Tank 30 and 32 supernate with approximately 150000 mg/L nitrate) suggested that IC detection limits for glycolate in these samples would be > 500 mg/L. Thus, an LOQ of 12 mg/L is not feasible by this IC method on evaporator feed samples. An alternative method of glycolate analysis, using proton nuclear magnetic resonance (H NMR), was developed by the research team. In initial tests, the H NMR technique provided reasonable quantitation of glycolate in Tank 22 conditions by direct observation of the liquid. The H NMR method may provide improved detection limits for solutions with higher nitrate concentrations. We recommend further development of this analysis for high nitrate LWS samples such as evaporator feed and evaporator drop tank content. A method for pretreatment of samples using crystalline silicotitanate (CST) was developed and tested. The objective of the pretreatment was to facilitate analysis of samples with higher levels of radioactivity. The pretreatment did not influence the glycolate concentration in solution, and the pretreatment is projected to reduce Cs 137 activity in a sample by a factor of 16,200. The double strike CST pretreatment was demonstrated and would allow milliliters of higher activity samples to be transferred from the Shielded Cells and handled in a containment unit for glycolate analysis. The various studies validated IC and H NMR methods for glycolate analysis, defined the range of applicability, and demonstrated key supporting analytical protocols. Based on the results, high quality glycolate analysis of Tank 22 is feasible down to approximately 12 mg/L, with the potential for broader applicability of the methods to other conditions in the Savannah River Site Liquid Waste System (LWS).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

IDP Analysis of Tank 9H Mined Well Samples

The Savannah River Site (SRS) Tank Farm Transfer Control Program Description Document (PDD) requires Inhalation Dose Potential (IDP) testing to determine whether a transfer qualifies as a “HIGH-REM” (> 2.08E+08 rem/gal) or a “LOW-REM” (≤ 2.08E+08 rem/gal) waste transfer. Due to the unexpected presence of solids in a variable depth sample taken from Tank 9H, transfers from salt tanks potentially containing insoluble solids are currently being assessed as sludge slurry transfers. The insoluble solids concentration of such samples were requested to be adjusted to 16.7 wt% for the results to be used to bound future transfers from that tank. Two samples, HTF-09-21-100 and -101, were received from Tank 9H for analysis to determine IDP. The initial percent insoluble solids concentrations of HTF-09-21-100 and -101 were measured at 4.04 wt% and 6.23 wt%, respectively. Sample HTF-09-21-101 was adjusted to a target of at least 16.7 wt% insoluble solids by removing supernate. Analysis measured the adjusted insoluble solids concentration at 19.0 wt%. A representative sample was obtained for analysis. Then, based on calculations, supernate was added back to HTF-09-21-101 to achieve a percent insoluble solids concentration of 17.1 wt%, and a representative sample was collected for analysis. Based on duplicate analyses at 17.1 wt% and 19.0 wt% insoluble solids, the samples contain less than 0.0012 Ci/L of gross alpha activity. Therefore, the Tank 9H material can be transferred as a “LOW REM” transfer.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Sludge Batch 11 Assembly: Tank 15

Savannah River Mission Completion Nuclear Safety and Engineering Integration (SRMC-E) has requested that Savannah River National Laboratory (SRNL) perform Tank 15 characterization analyses in support of Sludge Batch 11 (SB11) assembly. This report provides important characterization of the slurry in Tank 15 prior to transfer from Tank 15 to Tank 13 (prior to sending to Tank 51) in order to demonstrate that the transfer is "Low Rem" and ensure the sludge concurs with the estimated transfer mass for the SB11 recipe. A total of 3 sets of Tank 15 samples were delivered to SRNL from December 2022 to August 2023. The individual sets of composite samples were analyzed for density, weight percent solids, chemical composition, radionuclides, and/or supernate corrosion control tests.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Hanford Double Shell Waste Tank Corrosion Studies (Final Report FY2023)

For fiscal year (FY) 2023, the Savannah River National Laboratory (SRNL) focused on two experimental tasks related to Hanford Double Shell Tank (DST) chemistry and integrity. The first task focused on understanding risk of corrosion due to formation of either continuous layers or discrete patches of solids on the tanks’ inner sidewalls and bottoms. Differences in the conductivity between various layers of the tank (e.g., solids, liquid, etc.) could result in differences in the electrochemical potential of the tank metal at various locations. The electrochemical potential difference may result in a corrosion current between the coupled surfaces. In FY23, SRNL investigated test configurations and protocols that could evaluate the presence of a galvanic couple between the tank bottom and the tank wall.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2026

Conformable, Composite Tank for Liquid Hydrogen Storage in Heavy-duty Ground Transportation

Raytheon Technologies Research Center (RTRC), with its partners, Argonne National Labs and University of Dayton Research Institute proposed to design and manufacture a conformable composite tank prototype to store liquid Hydrogen (LH2) for heavy-duty truck applications. A dual-wall tank concept was proposed where an outer shell covers and protects an inner pressure vessel that stores liquid hydrogen. The inner pressure vessel is liner-less and designed with a conformable concept, derived from prior RTRC ARPA-E program where multiple vessels are coalesced into one, maximizing space utilization and reducing surface area for heat leak. The gap between the dual wall is evacuated of air pressure, and filled with multi-layer insulation to minimize radiative heat transfer. Fiber reinforced composites were proposed for light-weighting and conformable shape manufacture. To minimize LH2 leakage, carbon nanotubes-infused composites were proposed, in addition to use of thin plies. The project was divided into two 18-month budget periods (BP), BP1 and BP2, separated by a Go/No-Go gate review. The high-level goal for BP1 was to demonstrate the feasibility of the overall approach, the tank design, material selection, and manufacturing methodology, build and test a prototype tank; and the goal for BP2 was to design and manufacture a conformal tank per DOE requirements (minimum 20 kg of LH2) and perform a successful test of the tank. The program was terminated about two-thirds of the way through BP1. The outcome of this program would have permitted long-haul trucks and construction vehicles to use liquid Hydrogen, reducing the emission of green-house gases. This technology is applicable to other transportation sectors as well.

08 HYDROGEN↗
NASA NTRS2013

Temperature Stratification in a Cryogenic Fuel Tank

A reduced dynamical model describing temperature stratification effects driven by natural convection in a liquid hydrogen cryogenic fuel tank has been developed. It accounts for cryogenic propellant loading, storage, and unloading in the conditions of normal, increased, and micro- gravity. The model involves multiple horizontal control volumes in both liquid and ullage spaces. Temperature and velocity boundary layers at the tank walls are taken into account by using correlation relations. Heat exchange involving the tank wall is considered by means of the lumped-parameter method. By employing basic conservation laws, the model takes into consideration the major multi-phase mass and energy exchange processes involved, such as condensation-evaporation of the hydrogen, as well as flows of hydrogen liquid and vapor in the presence of pressurizing helium gas. The model involves a liquid hydrogen feed line and a tank ullage vent valve for pressure control. The temperature stratification effects are investigated, including in the presence of vent valve oscillations. A simulation of temperature stratification effects in a generic cryogenic tank has been implemented in Matlab and results are presented for various tank conditions.

stratification↗
NASA NTRSDate not supplied

Parametric Analysis of the Charge-Hold-Vent Method for Cryogenic Propellant Tank Chill Down

In the absence of external heat exchangers, the on-orbit transfer of cryogenic propellants requires the receiver tank to first be quenched to a sufficiently low energy state to allow for a continuous no-vent fill to avoid unnecessary venting of liquid. One proposed method for tank chilldown that minimizes the potential for venting liquid is the charge hold vent (CHV) method. CHV follows a cyclic process that gradually removes thermal energy from the receiver tank by injecting liquid with the vent valve closed and allowing the fluid and wall to reach near-thermal equilibrium before venting the superheated vapor. However, the CHV method must be optimized to minimize complexity, mass, and time. This paper presents a modular CHV analytical model used to quantify the number of cycles and propellant mass consumed based on first principles. The model is used to examine the effect of eight parameters: receiver tank material, volume, mass, maximum expected operating pressure, and initial pressure, liquid injection pressure and temperature, and the target temperature. The model is validated against the only two available CHV datasets. Based on results, the tank mass-to-volume ratio is the most important factor in determining the number of CHV cycles and thus degree of difficulty in tank chilldown. The model can easily be used for early-stage design, sizing, and analysis of cryogenic propellant transfer systems.

Tank Chilldown↗
NASA NTRS2024

Helium Conservation by Diffusion Limited Purging of Liquid Hydrogen Tanks

Recently, the new 1.2-million-gallon liquid hydrogen (LH 2 ) storage tank at the Kennedy Space Center Launch Complex 39 was purged, replacing nitrogen with helium, using a piston purge process. This purging approach had been used on Space Shuttle External Tanks (ET) in the past, but not on one of the large LH 2 tanks. The result was a huge success, potentially saving more than 1 million cubic feet of helium and significant labor. To better understand this result, a diffusion-based purge model was developed. The model predictions accurately match the helium concentration data, providing understanding of the helium savings and suggesting how to better optimize the purging of this large tank for future operations. The model was then applied to the piston purging of the Space Shuttle ETs, showing that these purges were not optimal, but that the amount of wasted helium was not excessive and that additional helium might have been saved by small changes to the process. The Space Shuttle Program (SSP) has ended, but these results are applicable to the purging of future large cylindrical LH 2 tanks. Finally, data have been obtained on inerting the Space Launch System (SLS) Core Stage (CS) LH 2 tank, replacing hydrogen with helium after an aborted launch. This is a more complicated case, but the model predicts that substantial helium might be saved by modifying the current purging process.

Helium Conservation↗
DOE OSTI2011

AUTOMATED LEAK DETECTION OF BURIED TANKS USING GEOPHYSICAL METHODS AT THE HANFORD NUCLEAR SITE

At the Hanford Nuclear Site in Washington State, the Department of Energy oversees the containment, treatment, and retrieval of liquid high-level radioactive waste. Much of the waste is stored in single-shelled tanks (SSTs) built between 1943 and 1964. Currently, the waste is being retrieved from the SSTs and transferred into newer double-shelled tanks (DSTs) for temporary storage before final treatment. Monitoring the tanks during the retrieval process is critical to identifying leaks. An electrically-based geophysics monitoring program for leak detection and monitoring (LDM) has been successfully deployed on several SSTs at the Hanford site since 2004. The monitoring program takes advantage of changes in contact resistance that will occur when conductive tank liquid leaks into the soil. During monitoring, electrical current is transmitted on a number of different electrode types (e.g., steel cased wells and surface electrodes) while voltages are measured on all other electrodes, including the tanks. Data acquisition hardware and software allow for continuous real-time monitoring of the received voltages and the leak assessment is conducted through a time-series data analysis. The specific hardware and software combination creates a highly sensitive method of leak detection, complementing existing drywell logging as a means to detect and quantify leaks. Working in an industrial environment such as the Hanford site presents many challenges for electrical monitoring: cathodic protection, grounded electrical infrastructure, lightning strikes, diurnal and seasonal temperature trends, and precipitation, all of which create a complex environment for leak detection. In this discussion we present examples of challenges and solutions to working in the tank farms of the Hanford site.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

Limit Load and Buckling Analysis for Assessing Hanford Single-Shell Tank Dome Structural Integrity

The U.S. Department of Energy, Office of River Protection has commissioned a structural analysis of record (AOR) for the Hanford single shell tanks (SSTs) to assess their structural integrity. The analysis used finite element techniques to predict the tank response to the historical thermal and operating loads. The analysis also addressed the potential tank response to a postulated design basis earthquake. The combined response to static and seismic loads was then evaluated against the design requirements of American Concrete Institute (ACI) standard, ACI-349-06, for nuclear safety-related concrete structures. Further analysis was conducted to estimate the plastic limit load and the elastic-plastic buckling capacity of the tanks. The limit load and buckling analyses estimate the margin between the applied loads and the limiting load capacities of the tank structure. The potential for additional dome loads from waste retrieval equipment and the addition of large dome penetrations to accommodate retrieval equipment has generated additional interest in the limit load and buckling analyses. This paper summarizes the structural analysis methods that were used to evaluate the limit load and buckling of the single shell tanks.

nuclear waste tank, structural evaluation, limit l↗
DOE OSTI2013

Transuranic Waste Processing Center (TWPC) Legacy Tank RH-TRU Sludge Processing and Compliance Strategy - 13255

The U.S. Department of Energy (DOE) needs to safely and efficiently treat its 'legacy' transuranic (TRU) waste and mixed low-level waste (LLW) from past research and defense activities at the Oak Ridge National Laboratory (ORNL) so that the waste is prepared for safe and secure disposal. The TWPC operates an Environmental Management (EM) waste processing facility on the Oak Ridge Reservation (ORR). The TWPC is classified as a Hazard Category 2, non-reactor nuclear facility. This facility receives, treats, and packages low-level waste and TRU waste stored at various facilities on the ORR for eventual off-site disposal at various DOE sites and commercial facilities. The Remote Handled TRU Waste Sludge held in the Melton Valley Storage Tanks (MVSTs) was produced as a result of the collection, treatment, and storage of liquid radioactive waste originating from the ORNL radiochemical processing and radioisotope production programs. The MVSTs contain most of the associated waste from the Gunite and Associated Tanks (GAAT) in the ORNL's Tank Farms in Bethel Valley and the sludge (SL) and associated waste from the Old Hydro-fracture Facility tanks and other Federal Facility Agreement (FFA) tanks. The SL Processing Facility Build-outs (SL-PFB) Project is integral to the EM cleanup mission at ORNL and is being accelerated by DOE to meet updated regulatory commitments in the Site Treatment Plan. To meet these commitments a Baseline (BL) Change Proposal (BCP) is being submitted to provide continued spending authority as the project re-initiation extends across fiscal year 2012 (FY2012) into fiscal year 2013. Future waste from the ORNL Building 3019 U-233 Disposition project, in the form of U-233 dissolved in nitric acid and water, down-blended with depleted uranyl nitrate solution is also expected to be transferred to the 7856 MVST Annex Facility (formally the Capacity Increase Project (CIP) Tanks) for co-processing with the SL. The SL-PFB project will construct and install the necessary integrated systems to process the accumulated MVST Facilities SL inventory at the TWPC thus enabling safe and effective disposal of the waste. This BCP does not include work to support current MVST Facility Surveillance and Maintenance programs or the ORNL Building 3019 U-233 Disposition project, since they are not currently part of the TWPC prime contract. The purpose of the environmental compliance strategy is to identify the environmental permits and other required regulatory documents necessary for the construction and operation of the SL- PFB at the TWPC, Oak Ridge, TN. The permits and other regulatory documents identified are necessary to comply with the environmental laws and regulations of DOE Orders, and other requirements documented in the SL-PFB, Safety Design Strategy (SDS), SL-A-AD-002, R0 draft, and the Systems, Function and Requirements Document (SFRD), SL-X-AD-002, R1 draft. This compliance strategy is considered a 'living strategy' and it is anticipated that it will be revised as design progresses and more detail is known. The design basis on which this environmental permitting and compliance strategy is based is the Wastren Advantage, Inc., (WAI), TWPC, SL-PFB (WAI-BL-B.01.06) baseline. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Grout testing to support tank closure at the Savannah River Site - 15370

Final operational closure of the Savannah River Site (SRS) waste tanks is performed by filling the voids inside the tanks with a series of tailored grout formulas thereby stabilizing the remaining residuals. These tailored grout formulas were each developed to meet the demands of the closure process. The needs identified for the closure process include (1) a flowable, self-consolidating, reducing grout for tank fill, (2) a highly flowable grout capable of filling the 7-1/4 kilometers (4-1/2 miles) of 5-centimeter (2-inch) diameter cooling coils inside the tank, and (3) a highly flowable grout capable of filling equipment remaining in the tank. The specific characteristics of each of the grout formulations were tested to ensure they would function as required during implementation. Performance testing was also done on some of the grout formulations to confirm mix effectiveness and improve worker handling proficiency and conducted additional testing on the bulk fill grout to expand the placement capabilities of the grout. This paper focuses on the evolution of the closure process' identified needs, formulation development, and the subsequent testing of the tailored grout formulations utilized in the past tank closure processes at SRS. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗