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At least 127 records · Page 7
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Modeling of SRS Aluminum-clad Spent Nuclear Fuel in Standard DOE Sealed Canisters

At the Savannah River Site (SRS), approximately 7 MTHM of aluminum-clad spent nuclear fuel is currently in wet storage at L-basin. A significant portion of this fuel is slated to be dried and sealed in DOE standard canisters for long-term storage. This report will focus on a few fuel types, which should bound all other fuel stored at SRS. These consist of the reference fuel assembly, the MURR fuel and the HFIR fuel. These bound a hypothetical maximum scenario for MTR box fuel, a maximum for decay heat in an assembly, and a maximum for the aluminum surface area, respectively. The geometry for the packages were created from DOE specifications for storage of each of the fuels. As long as the minimum time from reactor discharge to sealed storage is 3 years, then the highest temperatures which occur within the sealed canisters do not exceed 100° C for all scenarios considered here. In the geometry considered, while the HFIR fuel has a large surface area, the packaging configuration leaves large void areas, such that the aluminum surface to free volume ratio is about 30% less than an ATR or MURR packaging configuration.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Colloid-Facilitated Actinide Transport in a Cementitious-Impacted SRS Groundwater

The objective of this report is to provide a literature review relevant to the question of whether colloid facilitated transport of plutonium is enhanced in a cementitious groundwater system on the Savannah River Site (SRS). Contaminant transport is commonly described as taking place in a system with a mobile aqueous phase and an immobile solid phase. There has been an increasing awareness of a third phase, a mobile solid phase, also referred to as a mobile colloidal phase. Mobile colloids consist of organic and/or inorganic submicron-particles that move with groundwater flow. When radionuclides are associated with the mobile colloids, the net effect is that radionuclides can move faster through the subsurface system than would be predicted by transport models that do not include mobile colloids. It is important to distinguish between subsurface colloids and subsurface mobile colloids. The subsurface environment includes an enormous reservoir of colloids, but only a tiny fraction, if any, are mobile. Mobile colloid formation is commonly described as involving a three-step process: genesis, stabilization, and transport. It was concluded that there is a strong likelihood that submicron colloids of plutonium exist near the source term, (i.e., the genesis step is likely completed). While such particles have not been directly detected, it is highly likely that plutonium could either attach to submicron particles in the sediment or that submicron plutonium fragments exist at the source. However, the tendency for these plutonium colloids to move in the SRS subsurface is extremely low, especially in engineered cementitious environments where the ionic strength of the solution and the elevated concentrations of divalent cations greatly curtails colloid suspension stability (i.e., the stabilization step is likely not completed). Together these data strongly indicate that plutonium-bearing colloids would likely exist in the near field of a tank closure facility, but the colloids would be unlikely to be mobile, thereby providing a vector for enhanced transport. Under cementitious leachate impacted groundwater conditions, it is reasonable to assume plutonium transport occurs primarily as a two-phase system, a mobile aqueous phase and an immobile solid phase.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Applications of Alternate Actinide Calibration on Inductively Coupled Plasma Mass Spectrometry (ICPMS) Analysis of H-Canyon Highly Enriched Uranium Legacy Material at the Savannah River Site (SRS)

The Savannah River National Laboratory (SRNL) uses Inductively Coupled Plasma Mass Spectrometry (ICPMS) to support numerous programs at the Savannah River Site (SRS). As part of an objective to prepare high assay low enriched uranium (HALEU), the SRS H-Canyon needed analysis of legacy uranyl nitrate that is stored at its Outside Facilities (HOF). With the application of a standard practice for alternate actinide calibration on ICP-MS, SRNL has successfully demonstrated its capability to characterize the uranium/actinide composition of the HOF tanks that are intended for consolidation prior to preparation of the HALEU. A comparison of the uranium analytical results between recently upgraded SRNL ICP-MS and Inductively Coupled Plasma Optical Emission Spectrometry (ICP-OES) instrumentation shows the alternate actinide calibration is a valid complementary method when used in conjunction with other analytical techniques (Davies-Gray Potentiometric Titration). The methods, instrumentation, and ongoing developments on ICP-MS at SRNL will be discussed.

Jones, Mark A.↗
DOE OSTI2024

Utilizing HYSPLIT for Emergency Response Modeling at SRS

The threat of a disaster happening at the Savannah River Site (SRS) is low, but not zero. When an emergency arises, emergency responders utilize a variety of tools to detect, track, and mitigate the disaster. HYSPLIT is a community atmospheric dispersion model that is highly configurable and has been used operationally to simulate particle, chemical, and radionuclide releases. We have developed a preliminary user interface (UI) that configures HYSPLIT to run for a variety of Emergency Action Level (EAL) scenarios at SRS. Using a hypothetical EAL scenario of an accidental release of Pu-238 in H-area and K-area onsite, an analysis was done to show how a preliminary UI framework allows HYSPLIT to be run on a terminal without the need form an active network connection. The study was successful in multiple aspects. For the preliminary UI, users were able to input certain characteristics of the hypothetical EAL such as hours of emission (how long was Pu-238 being released), curies of release, release start time (when does the model run start), and type of radionuclide directly into the terminal. For HYSPLIT, the model was successful in being able to take the UI input, calculate the results from the inputs and output multiple figures of both total volumetric airborne concentrations and surface deposition. We used the Pu-238 Derived Intervention Level (DIL) deposition thresholds for produce, dairy, and beef for visualization of the surface deposition, showing how a release would impact ingestion pathways.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2024

Utilizing HYSPLIT for Emergency Response Modeling at SRS

At SRS, emergency responders use a variety of tools to detect, track, and mitigate hazardous material releases into the atmosphere. Two models currently used at SRS are Puff-Plume and the Lagrangian Particle Dispersion Model (LPDM), a Gaussian and Lagrangian model, respectively. A decision has been made to replace LPDM with the more widely-supported Hybrid Single-Particle Lagrangian Integrated Trajectory (HYSPLIT) model for evaluating inhalation and ingestion doses following a release. HYSPLIT is designed to compute complex dispersion and deposition simulation To achieve the implementation of HYSPLIT, we have developed a preliminary UI framework that will allow HYSPLIT to be run on ATG computers without the need for active network connections, thus avoiding the loss of capabilities in the event of a network outage during an emergency.

Earley, Ian↗
DOE OSTI2025

Literature Review Investigating Historical Plutonium Solubility in SRS Tank Waste

The Savannah River Site (SRS) has designed the Accelerated Basin De-inventory (ABD) program to accelerate the de-inventory of L-Basin and accelerate the Spent Nuclear Fuel (SNF) disposition mission. Similarly, the H-Canyon facility at SRS is reestablishing the 6.3D electrolytic dissolver for dissolving unirradiated stainless-steel (SS) clad Fast Critical Assembly (FCA) fuel. In both discard types (ABD and FCA), plutonium is present and its complex solubility when composited to Concentration, Storage, and Transfer Facility (CSTF) sludge is being investigated as it may have downstream impacts to the liquid waste (LW) organization. This literature review aims to highlight and compile the existing literature on plutonium solubility in waste streams relevant to ABD and FCA discards, as well as discuss some considerations in analyzing solubility data of plutonium. This review serves to help define the analysis methods for future experiments involving plutonium (and other actinides) and in designing appropriate testing conditions surrounding these studies. This review is broken up into five parts and will discuss: (i) The possible effects of testing hold time and temperature on plutonium solubility, (ii) the influence of neutralization rate and particle size of freshly precipitated discards, (iii) the coprecipitation of plutonium with iron and uranium, (iv) predictive solubility modeling and the influence of supernate anions on solubility, and (v) the speciation of plutonium in solutionas a result of supernate anions.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2011

PERFORMANCE TESTING OF THE NEXT-GENERATION CSSX SOLVENT WITH ACTUAL SRS TANK WASTE

Efforts are underway to qualify the Next-Generation Solvent for the Caustic Side Solvent Extraction (CSSX) process. Researchers at multiple national laboratories have been involved in this effort. As part of the effort to qualify the solvent extraction system at the Savannah River Site (SRS), SRNL performed a number of tests at various scales. First, SRNL completed a series of batch equilibrium, or Extraction-Scrub-Strip (ESS), tests. These tests used {approx}30 mL of Next-Generation Solvent and either actual SRS tank waste, or waste simulant solutions. The results from these cesium mass transfer tests were used to predict solvent behavior under a number of conditions. At a larger scale, SRNL assembled 12 stages of 2-cm (diameter) centrifugal contactors. This rack of contactors is structurally similar to one tested in 2001 during the demonstration of the baseline CSSX process. Assembly and mechanical testing found no issues. SRNL performed a nonradiological test using 35 L of cesium-spiked caustic waste simulant and 39 L of actual tank waste. Test results are discussed; particularly those related to the effectiveness of extraction.

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

Optimization of Saltcake Removal Flowsheet at SRS through Incorporation of Testing and In-Tank Waste Experience

Saltcake removal at SRS may be performed for several reasons: to provide space for evaporator operation (i.e., to precipitate more salt in the drop tank), to provide feed for salt processing (i.e. immobilize the waste), or to remove the salt for tank closure. Many different salt dissolution techniques have been employed in the 40 years that SRS has been performing salt removal, from a basic “Add, Sit, Remove” method (water is added on top of the saltcake and time is allowed for diffusion), to performing interstitial liquid removal, or using mixing devices to promote contact with the liquid. Lessons learned from previous saltcake removal campaigns, in addition to testing and modeling, have led to opportunities for improvements to the overall saltcake removal process. This includes better understanding of salt properties and behavior during dissolution; the primary concerns for salt dissolution are the release of radiolytic hydrogen and criticality prevention (post-dissolution). Recent developments in salt dissolution include the reuse of dilute supernate and a semi-continuous dissolution (SCD) process, where low volume mixing eductors are used to deliver water near the surface of the saltcake at the same rate as the salt solution is removed and transferred to a receipt tank.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

SRS 2022 Annual Site Environmental Report (ASER)

ASER stands for the Annual Site Environmental Report and is the primary document that the U.S. Department of Energy uses to inform the public of environmental performance and conditions at SRS. • ASERs are required by U.S. Department of Energy to provide the public and stakeholders information on: – Environmental program performance – Site wide environmental monitoring and surveillance effectiveness – Compliance status with environmental standards and requirements • SRS published its first ASER in 1959

DAVIS, COLLEEN↗
DOE OSTI2013

Groundwater Monitoring Optimization of Post Closure Waste Sites at SRS - 13184

Groundwater monitoring at the Savannah River Site (SRS) is required at dozens of waste sites and includes sampling at over 1,000 monitoring wells. The expected longevity of groundwater contamination and associated groundwater monitoring and reporting constitutes a significant long-term cost that represents an increasing proportion of the environmental management budget as surface waste units are closed. Therefore, a comprehensive evaluation of the monitoring program for eighteen regulated waste units was conducted to identify areas where monitoring could be optimized. The units evaluated varied considerably in the scope of monitoring; ranging from two wells to hundreds of wells. In order to systematically evaluate such disparate monitoring networks, SRS developed a decision-logic analysis using flow sheets to address potential areas of optimization. Five areas were identified for evaluation, including: (1) Comparison of current monitoring to regulatory requirements, (2) Spatial distribution, (3) Temporal sampling, (4) Analyte requirements, and (5) Reporting frequency and content. Optimization recommendations were made for fifteen of the eighteen groundwater units. The spatial evaluation resulted in recommendations to suspend sampling in 79 wells and add sampling at 16 wells. The temporal evaluation resulted in recommendations to reduce the number of well visits per year by 504. Analyte reductions were recommended at three groundwater units, with increases at three other units. Reporting frequency reductions were recommended for five units. Approximately $700,000 (direct dollars) of potential annualized cost savings were identified for these groundwater units, provided all recommendations are approved. The largest area of savings was associated with reducing the reporting frequency. The optimization approach has been presented to the EPA and South Carolina Department of Environmental Control (SCHDEC), with unit-specific recommendations approved for all five units presented. This approach can be expected to be highly successful for sites with rich historical data sets and where the requirements in regulatory monitoring plans can be negotiated. (authors)

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

Optimization of sludge waste removal flowsheet at SRS through incorporation of testing and in-tank waste experience - 15262

Sludge waste removal at SRS serves two purposes: to remove the sludge for tank closure, and to provide feed for sludge processing and immobilization via vitrification. Lessons learned from previous sludge removal campaigns, in addition to testing and modeling, have led to opportunities for improvements to the overall sludge removal process, as well as marked improvements in sludge removal equipment technology and operation. Testing, modeling, and experience have led to better understanding of sludge properties and behavior during agitation and settling; the primary consideration for sludge agitation is the retention and release of radiolytic hydrogen, which must be managed in order to prevent flammability concerns in the waste tanks. Mixing devices have improved considerably, from early sluicing devices to the long-shafted slurry pumps and submersible mixer pumps currently in use at SRS, as well as ongoing development with new commercial submersible mixer pumps. (authors)

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

Optimization of salt-cake removal flowsheet at SRS through incorporation of testing and in-tank waste experience - 15263

Salt-cake removal at SRS may be performed for several reasons: to provide space for evaporator operation (i.e., to precipitate more salt in the drop tank), to provide feed for salt processing (i.e. immobilize the waste), or to remove the salt for tank closure. Many different salt dissolution techniques have been employed in the 40 years that SRS has been performing salt removal, from a basic 'Add, Sit, Remove' method (water is added on top of the salt-cake and time is allowed for diffusion), to performing interstitial liquid removal, or using mixing devices to promote contact with the liquid. Lessons learned from previous salt-cake removal campaigns, in addition to testing and modeling, have led to opportunities for improvements to the overall salt-cake removal process. This includes better understanding of salt properties and behavior during dissolution; the primary concerns for salt dissolution are the release of radiolytic hydrogen and criticality prevention (post-dissolution). Recent developments in salt dissolution include the reuse of dilute supernate and a semi-continuous dissolution (SCD) process, where low volume mixing eductors are used to deliver water near the surface of the salt-cake at the same rate as the salt solution is removed and transferred to a receipt tank. (authors)

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

The influence of humic acid and colloidal silica on the sorption of U(VI) onto SRS Sediments Collected from the F/H Area - 15499

Between 1955 and 1988, the F/H Area Seepage Basins located in the center of SRS received approximately 6,814,000 m{sup 3} of acidic waste solutions containing radionuclides and dissolved metals. Uranium (VI) is a key contaminant of concern in groundwater associated with the basins. The pump-and-treat water treatment unit designed and built in 1997 to remove metals and radionuclides eventually became less effective, prompting research for new remedial alternatives. The pump-and-treat system was discontinued in 2004, replaced with a funnel-and-gate system in which base solutions are injected at the gates to raise pH. To take advantage of the potential formation of uranium silicate minerals, Florida International University - Applied Research Center (FIU-ARC), in collaboration with Savannah River National Laboratory (SRNL), is conducting an investigation into a base containing sodium silicate to evaluate whether these solutions have sufficient alkalinity to correct the acidic nature of the aquifer sediments. The research was extended to investigate if U(VI) is bound to colloidal silica and if any synergy exists between humic acid (HA) and colloidal Si that would influence the removal of uranium. Humic substances (HS) are a major component of soil organic matter and are known for their abilities to influence the migration behavior and fate of heavy metals. Studies showed that HA function as an important ion exchange and metal-complexing ligand, carrying a large number of functional groups with high complexing capacity that can greatly affect the mobility behavior of actinides in natural systems. The study investigates if there are any synergistic interactions between U(VI) ions, humic acid and colloidal silica under oxidized conditions and studies the influence of HA and Si on the sorption of U(VI) onto sediments collected from the F/H Area. The experiments also evaluated the effects of different environmental variables such as pH, presence of U(VI) and varying concentrations of HA on the sorption behavior of U(VI) in multi-component batch systems in the pH range between 3 and 8. Several background sediment samples were collected at SRS from archived cores obtained during installation of the well FSB 91C and used for preparation of triplicate samples. The experimental matrix was set up as follows: - Batch 1: Si (3.5 mM) + U (VI) (0.5 ppm) (no sediments and no HA); - Batch 2: Si (3.5 mM) + U (VI) (0.5 ppm) + HA (10 ppm) (no sediments); - Batch 3: U (VI) (0.5 ppm) + HA (10 ppm) (no Si and no sediments); - Batch 4: Sediments + Si (3.5 mM) + U (VI) (0.5 ppm) (no HA); - Batch 5: Sediments + Si (3.5 mM) + U (VI) (0.5 ppm) + HA; - Batch 6: Sediments + U (VI) (0.5 ppm) + HA, (no Si); - Batch 7: Sediments + U (VI) (0.5 ppm) (no Si and no HA). All samples were prepared in triplicate 50-mL conical tubes. All control and experimental tubes were vortexed and kept on a shaker for 24 h at 100 rpm and room temperature. Results showed that the sediment-bearing batches typically exceeded the performance of non-sediment batches. Results showed that the sorption of U(VI) onto sediments was influenced by the pH and the presence of HA. At low pH 3, the addition of HA slightly enhances U(VI) sorption compared to the HA-free samples. This result is consistent with other studies; in the acidic pH range, the addition of HA slightly enhances U(VI) sorption relative to the HA-free system. Conversely, in the pH range of 4 to 7, the sorption of U(VI) on Si was reduced in the presence of HA in comparison to the system without HA. In subsequent experiments the concentration of HA will be increased to 50 ppm. (authors)

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

Modeling and Quantitative Assessment of Green and Sustainable Remediation Options for the M1 Air Stripper System at DOE SRS - 15660

The effectiveness of remediation systems needs to be reviewed periodically to assess their performance over the years relative to efficiency and effectiveness. New technologies and Best Management Practices have emerged that can improve existing remediation systems. From a Green and Sustainable Remediation analysis, the higher energy and/or water usage of older systems might not be justified from an environmental protection rationale. For example, remediation systems can run intermittently and with slower flow rates, use less power, and still provide the required level of environmental remediation and protection. This research by the Applied Research Center at Florida International University (FIU) is the preliminary phase of a Green and Sustainable Remediation (GSR) analysis of the operations of the M1 air stripper and the A/M Area groundwater remediation system at the Savannah River Site (SRS). An initial baseline analysis has helped identify opportunities for implementing Green and Sustainable Remediation (GSR) practices while maintaining the hydraulic containment and effectiveness of the remediation system. This study included the collection of contaminant recovery data from SRS reports into a database of trichloroethene (TCE) and tetrachloroethene (PCE) removal from contaminant recovery wells RWM-1 through RWM-12. Presented below are data and preliminary findings from these recovery wells. Findings show that 7 of 12 recovery wells have transitioned to more PCE removal than TCE removal; the nearby steam injection remediation of 'Dynamic Underground Stripping' mobilized DNAPL and resulted in increased recovery in nearby wells. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Evaluation and Impacts of Mercury in the SRS Liquid Waste System - 16121

The Savannah River Site (SRS) Liquid Waste System (LWS) stores radioactive waste in 43 underground tanks. The radionuclides in the waste are removed through a series of separation processes and the low-level fraction is immobilized in a grout waste form while the high level fraction is disposed in a glass waste form. Mercury originated from decades of canyon processing (used as a catalyst for dissolving the aluminum cladding of reactor fuel) and is present throughout the LWS (∼60 metric tons). Mercury has long been a consideration in the LWS, from both hazard and processing perspectives. Mercury is removed from the LWS in many ways including: 1) the Tank Farm evaporator condensates; 2) at the Defense Waste Processing Facility (DWPF) where it is steam-stripped and removed during the feed preparation processes; 3) it is removed at the Effluent Treatment Plant (ETP); and it is immobilized in the low-level, grout waste form. An integrated, system-wide evaluation and assessment of mercury behavior in the LWS is being performed. The recently completed Phase I evaluation focused on mercury inventory and speciation in the LWS; mercury holdup and chemical processing behavior; mercury impact Identification, including worker safety and equipment degradation; and mercury removal and disposal options. The mercury removal system at the DWPF has been unable to remove mercury successfully due to equipment and chemistry challenges since the 2008 timeframe. Phase I evaluations indicate that approximately 43 percent of the mercury in waste sent to the DWPF is now returning to the Tank Farm in the recycle stream and is concentrating to higher than expected levels (∼500 mg/l) around the 2H evaporator system. The chemical behavior of mercury is further complicated by the formation of higher than expected quantities of organomercury species, specifically methyl mercury, now present in the DWPF recycle evaporator system (2H Tank farm evaporator), which is used to volume-reduce the recycle stream. Phase II activities are building on the Phase I mercury activities, including additional sampling and characterization activities, a re-assessment of overall system knowledge, ranking and prioritizing critical gaps/information in mercury behavior across the flowsheet, assessing impacts of mercury removal and disposal options, and documenting an action plan for overall mercury management in the LWS. Phase II activities include two System Engineering Evaluations to 1) 'Reestablish Mercury removal Capability within DWPF', and 2) 'Determine the Best Alternative Mercury Removal Location with the LWS'. This paper provides the results of the Phase I and II mercury evaluations for the SRS LWS. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Hydraulic and mixing evaluations for SRS miscible liquid solutions

The hydraulic results demonstrate that pump head pressure of 20 psi recirculates about 5.6 liters/min flowrate through the existing 0.131-inch orifice when a valve connected to NT-41 is closed. In case of the valve open to NT-41, the solution flowrates to Savannah River Site (SRS) tanks, NT-21 and NT-41, are found to be about 0.5 lpm and 5.2 lpm, respectively. The modeling calculations for the mixing operations of miscible fluids contained in the SRS HB-Line tank NT-21 were performed by taking a three-dimensional Computational Fluid Dynamics (CFD) approach. The CFD modeling results were benchmarked against the literature results and the previous SRNL test results to validate the model. Final performance calculations were performed for the nominal case by using the validated model to quantify the mixing time for the HB-Line tank. The results demonstrate that when a pump recirculates a solution volume of 5.7 liters every minute out of the 72-liter tank contents containing two acid solutions of 2.7 M and 0 M concentrations (i.e., water), a minimum mixing time of 1.5 hours is adequate for the tank contents to get the tank contents adequately mixed. In addition, the sensitivity results for the tank contents of 8 M existing solution and 1.5 M incoming species show that the mixing time takes about 2 hours to get the solutions mixed. (authors)

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

Using Correlations to Cs-137 to Reduce Uncertainty in SRS Tank Farm Tc-99 and I-129 Inventories - 19266

Performance Assessment (PA) modeling of the Saltstone Disposal Facility (SDF) shows that iodine-129 (I-129) dominates peaks of potential future doses, followed by contributions from technetium-99 (Tc-99). While these radionuclides are important from a PA perspective, they are relatively unimportant from a safety analysis or occupational exposure perspective. As such, the storage of Tc-99 and I-129 within liquid waste at Savannah River Site (SRS) have not been well tracked, resulting in considerable uncertainty in the estimated inventories of Tc-99 and I-129 used for PA modeling. Alternatively, cesium-137 (Cs-137) is considered a high-risk radionuclide from an occupational exposure perspective and is well characterized. Within the liquid waste at SRS, analyses have shown that concentrations of Cs-137 are closely correlated to both Tc-99 and I-129. These correlations and the abundance of Cs-137 data were leveraged to significantly reduce inventory uncertainty for waste destined to be disposed at the SDF. Previous attempts to mathematically define these correlations had two flaws: (1) the defined correlations failed to address the impact of the relatively short half-life of Cs-137 compared to the long half-lives of Tc-99 and I-129, and (2) the analyses relied on a small number of samples. To address the issue of the decay rate from the short half-life, all Cs-137 concentrations were decayed to a single future date. To improve the number of samples used to develop the correlations, a literature review was performed resulting in a larger data set. The data underwent a validation process, then the correlations were updated. The resulting correlations were used to improve inventory estimates for Tc-99 and I-129 for an ongoing revision to the SDF PA. (author)

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

History of Ultrasonic Inspection of High-Level Waste Tanks at SRS - 20220

The Liquid Waste Disposition Projects (LWDP) Structural Authority and Inspection Engineering Support Group has teamed with the Materials Science and Technology Non-destructive Evaluation Group of Savannah River National Laboratory (SRNL) to implement a comprehensive In-Service Inspection (ISI) program for the High-Level Waste (HLW) tanks in the sites tank farms. The purpose of this program is to demonstrate the structural and leak integrity of the tanks to maintain the function of waste containment throughout the desired service life. Four types of waste tanks were constructed at Savannah River Site (SRS) from 1951 thru 1981 to store radioactive nuclear waste from the separations process. The Type I, II and III are carbon steel cylindrical tanks with varying degrees of access to the side walls and very limited access to the tank bottom. The tanks contain 750,000 to 1.3 million gallons of waste and are buried or backfilled with soil. The tanks are 75'-85' in diameter and 24'-33' from tank bottom to the tank roof. Inspection of the HLW tanks is made difficult by radiation, contamination, and small access openings. Through the years these challenges have been overcome by progressive improvements to programs and equipment resulting in the sophisticated wall crawler Ultrasonic Testing (UT) technology of today. A formal inspection program was developed and put in service in 1972. UT thickness measurements were first made in 1967 and 1969 of selected waste tanks using an analog-type instrument to measure the thickness of the carbon steel walls. Routine inspections for thinning were performed from 1972-1985 and then discontinued because no thinning had been observed and this technique could not detect pitting or cracking. UT inspections resumed in 1994 using a sophisticated robotic wall crawler with UT technology capable of detecting thinning, cracking and pitting. The current program focuses on the newer Type III tanks that are used primarily for storage. The program was upgraded in 2003 to include selected regions of the tank wall, secondary wall and annulus floor on a 7 to 10-year frequency. The facility is currently in its third cycle of tank inspections. Thus far, there have been no reportable indications of service induced degradation. This result provides assurance that the chemistry control program has been effective at mitigating corrosion and that the tanks remain structurally sound even after more than 40 years of service. This paper will present the UT technology first used at SRS and the evolution to the current systems and their capabilities. Additionally, plans for equipment upgrades that will improve the extent of inspection will be discussed. (authors)

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