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Hanford Site Composite Analysis: Radionuclide Selection for Groundwater Pathway Evaluation

The updated Hanford Site Composite Analysis (CA) (DOE/RL-2019-52, Composite Analysis for Low-Level Waste Disposal in the Hanford Site Central Plateau (FY 2020)) provides an all-pathways dose projection to a hypothetical future member of the public from all planned low-level radioactive waste disposal facilities and potential contributions from all other projected end-state sources of radioactive material left at the Hanford Site following site closure. Its primary purpose is to support the decision-making process of the U.S. Department of Energy (DOE) under DOE O 435.1-1, Radioactive Waste Management, related to managing low-level waste disposal facilities at the Hanford Site. A key aspect of conducting a CA is selecting the radionuclides to be analyzed. This document describes the selection process for radionuclides to be included in the quantitative analysis.

07 ISOTOPE AND RADIATION SOURCES↗

Laboratory report on iodine ( 129 I and 127 I) speciation, transformation and mobility in Handford groundwater, suspended particles and sediments

The Hanford Site in eastern Washington produced plutonium for several decades and in the process generated billions of gallons of radioactive waste. Included in this complex mixture of waste was 50 Ci of iodine-129 ( 129 I). Iodine-129’s high abundance, due to its high fission yield, and extreme toxicity result in iodine-129 becoming a key risk driver at many Department of Energy (DOE) sites. The mobility of radioiodine in arid environments, such as the Hanford Site, depends largely on its chemical speciation and is also greatly affected by many other environmental factors, especially natural sediment organic matter (SOM). Groundwater radioiodine speciation has not been measured in arid regions with major plumes or large disposed 129 I inventories, including the Hanford Site, Idaho National Laboratory, and Nevada Test Site. In this study, stable iodine-127 and radioiodine-129 speciation, pH, and dissolved organic carbon (DOC) of groundwater samples collected from seven wells located in the 200-West Area of the Hanford site were investigated. The most striking finding was that iodate (IO 3 - ) was the most abundant species. Unexpectedly, iodide (I - ), which was likely the form of iodine in the source materials and the expected dominant groundwater species based on thermodynamic considerations, only accounted for 1-2% of the total iodine concentration. It is likely that the relatively high pH and the low abundance of sedimentary organic matter (SOM) that is present at the site slowed down or even inhibited the reduction of iodate, as SOM abiotically reduce iodate into iodide. Moreover, a study on the kinetics of iodide and iodate uptake and aqueous speciation transformation by three representative subsurface Hanford sediments was performed over a period of about one month. This study was carried out by using iodide-125 or iodate-125 at the ambient iodine-127concentration found at the site. Iodate K d values were on average 89% greater than iodide K d values, and the K d values for both species tended to increase with the amount of organic carbon (OC) present in the sediment. It is especially noteworthy that this trend existed at the very low OC concentrations that naturally exist in the Hanford sediments. Iodine and OC can form essentially irreversible covalent bonds, thereby providing a yet unstudied 129 I retardation reaction at the Hanford Site. In addition to the transformation of iodine species, the sediment collected from the vadose zone also released stable iodide into the aqueous phase. It was found that the three sediments all took up the ambient iodate from the groundwater and slowly transformed it into iodide under the laboratory conditions, likely dependent on the abundance of reducing agents such as organic matter and Fe 2+ . Therefore two competitive iodine processes were identified, the tendency for the sediment to reduce iodate to iodide, and the groundwater chemistry to maintain the iodine as iodate, presumably it is largely the result of natural pH and dissolved O 2 /Eh levels. Suspended carbonate (and silica) particles collected from Hanford groundwater contained elevated amounts of iodine (142 ± 8 μg/g iodine), consisting mainly of iodate (>99%). Iodate was likely incorporated into the carbonate structure during calcite precipitation upon degasing of CO 2 as the groundwater samples were removed from the subsurface. This concentration of groundwater iodate in precipitated carbonate has implication to long-term fate and transport of 129 I and on active in-situ 129 I groundwater remediation. This study provides some of the first groundwater radioiodine speciation studies conducted in arid environments and provides much needed mechanistic descriptions to permit making informed decisions about low-cost/high intellectual input remediation options, such as monitored natural attenuation, or long-term stewardship of nuclear waste disposal sites.

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Technology for Treatment of Liquid Radioactive Waste Generated during Uranium and Plutonium Chemical and Metallurgical Manufacturing in FSUE PO Mayak - 13616

Created technological scheme for treatment of liquid radioactive waste generated while uranium and plutonium chemical and metallurgical manufacturing consists of: - Liquid radioactive waste (LRW) purification from radionuclides and its transfer into category of manufacturing waste; - Concentration of suspensions containing alpha-nuclides and their further conversion to safe dry state (calcinate) and moving to long controlled storage. The following technologies are implemented in LRW treatment complex: - Settling and filtering technology for treatment of liquid intermediate-level waste (ILW) with volume about 1500m{sup 3}/year and alpha-activity from 10{sup 6} to 10{sup 8} Bq/dm{sup 3} - Membrane and sorption technology for processing of low-level waste (LLW) of radioactive drain waters with volume about 150 000 m{sup 3}/year and alpha-activity from 10{sup 3} to 10{sup 4} Bq/dm{sup 3}. Settling and filtering technology includes two stages of ILW immobilization accompanied with primary settling of radionuclides on transition metal hydroxides with the following flushing and drying of the pulp generated; secondary deep after settling of radionuclides on transition metal hydroxides with the following solid phase concentration by the method of tangential flow ultrafiltration. Besides, the installation capacity on permeate is not less than 3 m{sup 3}/h. Concentrates generated are sent to calcination on microwave drying (MW drying) unit. Membrane and sorption technology includes processing of averaged sewage flux by the method of tangential flow ultrafiltration with total capacity of installations on permeate not less than 18 m{sup 3}/h and sorption extraction of uranium from permeate on anionite. According to radionuclide contamination level purified solution refers to general industrial waste. Concentrates generated during suspension filtering are evaporated in rotary film evaporator (RFE) in order to remove excess water, thereafter they are dried on infrared heating facility. Solid concentrate produced is sent for long controlled storage. Complex of the procedures carried out makes it possible to solve problems on treatment of LRW generated while uranium and plutonium chemical and metallurgical manufacturing in Federal State Unitary Enterprise (FSUE) Mayak and cease its discharge into open water reservoirs. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Simplifying the Tracking of Waste Disposal Inventories: Process Improvement for Rapid Data Interpretation and Reduced Errors - 16143

Tracking the inventory of low-level radioactive waste at an operating disposal facility offers unique challenges. At the Savannah River Site's Saltstone Disposal Facility (SDF) the inventory used in Performance Assessment modeling is determined based on radionuclide concentrations from the aqueous waste in a feed tank and the volume of the waste that is transferred from the feed tank to the Saltstone Production Facility (SPF). These inputs are reported quarterly. As such, the disposal inventories are also updated quarterly; however, due to decay and ingrowth, the inventory values for some radionuclides can be misrepresented if the data is not tracked appropriately. Additionally, it is not feasible to analyze the feed tank samples for every possible radionuclide; therefore, a series of adjustments are necessary to estimate the concentrations of radionuclides that are not explicitly measured but still pose a potential dose risk. Prior to 2015, the SDF disposal inventories were updated using a series of spreadsheets that required users to perform a significant number of steps (including manually transferring data back and forth between a separate decay and ingrowth calculation software). Despite this labor intensive approach, the process only provided single point-in-time results. With multiple steps, minimal automation, and limited result data, this approach to inventory tracking was not ideal and minor mistakes were often identified during the equally cumbersome review process. As a process improvement, the simplified SDF Waste Inventory Disposed Estimator (SDF-WIDE) Model was developed using GoldSim software. The software allows users to identify specific dates to define the duration of the simulation, from the beginning of operations until the time of future facility closure, and provides graphical interfaces for displaying data over time. Further, GoldSim's Contaminant Transport Module is used to perform the decay and ingrowth calculations, making the model completely self-contained (i.e., there is no need to transfer intermediate data back and forth between different programs). GoldSim also includes model 'versioning' features that significantly reduce the time needed to perform the appropriate quality assurance reviews as required for each quarterly inventory update. By transitioning from the complex spreadsheets to the simplified SDFWIDE model, the time needed to perform updates to the SDF inventory has been reduced from days to hours, the risk of making human errors has been greatly mitigated, and users are able to quickly observe trends in the data over time. (authors)

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Recent Developments in Radioactive Waste Management in the Federal Republic of Germany - 17450

In 2013 the Federal Parliament and the Federal Council passed the Act Concerning the Search and Selection of a Site for a Repository for Heat-Generating Radioactive Waste and the Amendment of Other Acts (Site Selection Act). It was a restart for the search for a repository for especially heat-generating waste in Germany. As a first step in May 2014 the Commission 'Storage of High-Level Radioactive Waste' started their work. Their aim was to debate and clarify the principle questions concerning the management of heat-generating radioactive waste. As well as to evaluate and determine the fundamentals for the Site Selection Process e.g. safety requirements, exclusion criteria, minimum requirements and consideration criteria for the site selection. They also evaluated the organizational and procedural arrangements in the scope of waste management. As a first act to implement the suggestions of this Commission the organizational structures in waste management were greatly altered. This reorganization not only had a significant influence on the waste management of high-active, but also on the management of low- and intermediate-active waste management in the Federal Republic of Germany. In 2015 a second Commission started its work focusing on the funding of waste management in Germany. The results of this Commission were presented in April 2016. At the moment the suggestions of this Commission are being implemented, changing the financial responsibilities for waste management in the Federal Republic of Germany. The presentation will focus on changes and the developments of radioactive waste management in Germany and talk about the ongoing process in implementing the suggestions of the two Commissions. (authors)

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Initiating the Validation of CCIM Processability for Multi-phase all Ceramic (SYNROC) HLW Form: Plan for Test BFY14CCIM-C

This plan covers test BFY14CCIM-C which will be a first–of–its-kind demonstration for the complete non-radioactive surrogate production of multi-phase ceramic (SYNROC) High Level Waste Forms (HLW) using Cold Crucible Induction Melting (CCIM) Technology. The test will occur in the Idaho National Laboratory’s (INL) CCIM Pilot Plant and is tentatively scheduled for the week of September 15, 2014. The purpose of the test is to begin collecting qualitative data for validating the ceramic HLW form processability advantages using CCIM technology- as opposed to existing ceramic–lined Joule Heated Melters (JHM) currently producing BSG HLW forms. The major objectives of BFY14CCIM-C are to complete crystalline melt initiation with a new joule-heated resistive starter ring, sustain inductive melting at temperatures between 1600 to 1700°C for two different relatively high conductive materials representative of the SYNROC ceramic formation inclusive of a HLW surrogate, complete melter tapping and pouring of molten ceramic material in to a preheated 4 inch graphite canister and a similar canister at room temperature. Other goals include assessing the performance of a new crucible specially designed to accommodate the tapping and pouring of pure crystalline forms in contrast to less recalcitrant amorphous glass, assessing the overall operational effectiveness of melt initiation using a resistive starter ring with a dedicated power source, and observing the tapped molten flow and subsequent relatively quick crystallization behavior in pans with areas identical to standard HLW disposal canisters. Surrogate waste compositions with ceramic SYNROC forming additives and their measured properties for inductive melting, testing parameters, pre-test conditions and modifications, data collection requirements, and sampling/post-demonstration analysis requirements for the produced forms are provided and defined.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

NOCHAR Polymers: An Aqueous and Organic Liquid Solidification Process for Cadarache LOR (Liquides Organiques Radioactifs) - 13195

To handle the Very Low Level Waste (VLLW) and the Low Level Waste (LLW) in France, two options can be considered: the incineration at CENTRACO facility and the disposal facility on ANDRA sites. The waste acceptance in these radwaste routes is dependent upon the adequacy between the waste characteristics (physical chemistry and radiological) and the radwaste route specifications. If the waste characteristics are incompatible with the radwaste route specifications (presence of significant quantities of chlorine, fluorine, organic component etc or/and high activity limits), it is necessary to find an alternative solution that consists of a waste pre-treatment process. In the context of the problematic Cadarache LOR (Liquides Organiques Radioactifs) waste streams, two radioactive scintillation cocktails have to be treated. The first one is composed of organic liquids at 13.1 % (diphenyloxazol, mesitylene, TBP, xylene) and water at 86.9 %. The second one is composed of TBP at 8.6 % and water at 91.4 %. They contain chlorine, fluorine and sulphate and have got alpha/beta/gamma spectra with mass activities equal to some kBq.g{sup -1}. Therefore, tritium is present and creates the second problematic waste stream. As a consequence, in order for disposal acceptance at the ANDRA site, it is necessary to pre-treat the waste. The NOCHAR polymers as an aqueous and organic liquid solidification process seem to be an adequate solution. Indeed, these polymers constitute an important variety of products applied to the treatment of radioactive aqueous and organic liquids (solvent, oil, solvent/oil mixing etc) and sludge through a mechanical and chemical solidification process. For Cadarache LOR, N910 and N960 respectively dedicated to the organic and aqueous liquids solidification are considered. With the N910, the organic waste solidification occurs in two steps. As the organic liquid travels moves through the polymer strands, the strands swell and immobilise the liquid. Then as the polymer-organic cure, over time, the polymer continues to collapse on the organic to create a permanent bond. The N960 has the ability to absorb aqueous waste up to 100 times its own weight. It creates a strong mechanical bond which permanently traps the contamination imbedded in the aqueous liquids. As a consequence, these two polymers seem to be able to constitute a suitable solidification matrix for a final acceptance in storage on ANDRA sites. In order to validate the polymers as an acceptable aqueous and organic solidification process for Cadarache LOR, some solidification tests realized with N910 and N960, have been carried out for different Waste/Polymer ratios. The determination of the best Waste/Polymer ratio and of the optimal experimental parameters has been made through an exudation test. Indeed, the process prevents leaching and it results in the absence of residual free organic or aqueous liquid which is forbidden in storage by ANDRA specifications. With these test results, we generated scientific data which are fundamental to obtain an ANDRA agreement. As a conclusion, the aim of this study is to demonstrate that the solidification by polymers can constitute a pre-treatment solution for Cadarache LOR and more generally, for various organic and mixed organic/aqueous waste which can not be directly acceptable at CENTRACO facility or at ANDRA storage sites. This work is, therefore, a solid background to demonstrate the feasibility of the waste pre-treatment by solidification with polymers and to encourage the development of this process. (authors)

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Progress of the High Level Waste Program at the Defense Waste Processing Facility - 13178

The Defense Waste Processing Facility at the Savannah River Site treats and immobilizes High Level Waste into a durable borosilicate glass for safe, permanent storage. The High Level Waste program significantly reduces environmental risks associated with the storage of radioactive waste from legacy efforts to separate fissionable nuclear material from irradiated targets and fuels. In an effort to support the disposition of radioactive waste and accelerate tank closure at the Savannah River Site, the Defense Waste Processing Facility recently implemented facility and flowsheet modifications to improve production by 25%. These improvements, while low in cost, translated to record facility production in fiscal years 2011 and 2012. In addition, significant progress has been accomplished on longer term projects aimed at simplifying and expanding the flexibility of the existing flowsheet in order to accommodate future processing needs and goals. (authors)

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Update on the Status of Deep Borehole Disposal of High-Level Radioactive Waste in Germany - 20285

The phase-out of nuclear energy in Germany will take place in 2022. A site for final disposal of high level radioactive waste (HLRW) has not yet been found, but a site selection process was restarted by Act on the Search for and Selection of a Site for a Disposal Facility for High-Level Radioactive Waste (Site Selection Act - StandAG 2017). This act was based on a recommendation by a commission which also advised to follow up the development of deep borehole disposal (DBD) as an alternative option for final disposal of HLRW. This paper summarizes briefly the status of DBD in Germany and if this option should be pursued in Germany. Although there are some merits of DBD, it can only be a real option if research and development is supported. The technical equipment for boreholes of the required size will only be developed if there is funding and a feasibility test. Furthermore, any concept of DBD and technology must be detailed further, and some requirements of the act must be reconsidered. The paper concludes that despite the possible merits of DBD, the political and financial support for R and D will only be provided for the time being if DBD is pushed by interested parties. Alternatively, if the site selection procedure is not progressing well, then the alternative option of DBD may find greater interest in the future. However, its availability will be severely limited due to the missing development of detailed concepts and its demonstration of feasibility. (authors)

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Peridynamic modeling of cementitious materials for nuclear waste management

Radioactive and hazardous waste generated from fuel processing plants, nuclear reactors, and hospitals, requires effective management strategies. Cementitious materials are widely applied for these needs, serving as structural materials, reactive barriers, or waste forms. In these applications cracking poses a significant risk to performance, driven by inconsistent shrinkage of the components and the varying strength and permeability of their interfaces. Traditional modeling approaches face challenges in representing the complex fracture behavior of cementitious materials due to the reliance on spatial derivatives and difficulties with mesh generation. Here, to overcome these limitations, we employ peridynamics, a novel continuum mechanics formulation that uses integrals to describe mechanical equilibrium, avoiding discontinuities associated with traditional methods. Through incorporation of a bi-linear softening model and quasistatics, an experimentally validated model for Portland cement concrete samples was created. Mechanical parameters, including compressive strength and elastic modulus were validated and variation due to aggregate packing was evaluated. Additionally, sensitivity analysis of the peridynamic parameters to the bulk material properties was established. The results lay the groundwork for evaluating the impact of unique conditions of cementitious waste forms that can be assessed to improve the reliability of waste management strategies.

Aggregates↗

On the possibility of reprocessing spent nuclear fuel and radioactive waste by plasma methods

The concept of plasma separation of spent nuclear fuel and radioactive waste is presented. An approach that is based on using an accelerating potential to overcome the energy and angular spread of plasma ions at the separation region inlet and utilizing a potential well to separate spatially the ions of different masses is proposed. It is demonstrated that such separation may be performed at distances of about 1 m with electrical potentials of about 1 kV and a magnetic field of about 1 kG. The estimates of energy consumption and performance of the plasma separation method are presented. These estimates illustrate its potential for technological application. The results of development and construction of an experimental setup for testing the method of plasma separation are presented.

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Prediction of Peak Hydrogen Concentrations for Deep Sludge Retrieval in Tanks AN-101 and AN-106 from Historical Data of Spontaneous Gas Release Events

Radioactive and chemical wastes from nuclear fuel processing are stored in large underground storage tanks at the Hanford Site. The Tank Operations Contractor is continuing a program of moving solid wastes from single-shell tanks (SSTs) to double-shell tanks (DSTs) and preparing for waste feed delivery (WFD). A new mechanism for a large spontaneous gas release event (GRE) in deep sludge sediments has been postulated. The creation of this potential new GRE hazard, deep sludge gas release events (DSGREs), is the retrieval of sludge waste into a single DST that results in a sediment depth greater than operating experience has demonstrated is safe. The Tank Operations Contractor program of moving solid wastes from SSTs to DSTs and preparing for WFD is being negatively impacted by this sediment depth limit.

Deep sludge retrieval↗

Chemical Composition Measurements of LAWA44 Glass Samples

DOE is building the Hanford Tank Waste Treatment and Immobilization Plant (WTP) at the Hanford Site in Washington to remediate 55 million gallons of radioactive waste that is temporarily stored in 177 underground tanks. Both low-activity and high-level wastes will then be vitrified into borosilicate glass using Joule-heated ceramic melters. Efforts are being made to increase the loading of Hanford tank wastes in the glass. One area of work is enhancing waste glass composition/property models and broadening the compositional regions over which those models are applicable. In this report, the Savannah River National Laboratory provides chemical analysis results for several samples of a simulated low-activity waste glass, LAWA44, provided by the Pacific Northwest National Laboratory as part of an ongoing development task. The measured chemical composition data are reported and compared with the targeted values for each component for each glass. A detailed review showed no indications of errors in the preparation or measurement of the study glasses. All of the measured sums of oxides for the study glasses fell within the interval of 97.9 to 102.6 wt %, indicating acceptable recovery of the glass components. Comparisons of the targeted and measured chemical compositions showed that the measured values for the glasses met the targeted concentrations within 10% for those components present at more than 5 wt %. It was noted that the measured B 2 O 3 concentrations are somewhat above the targeted values for the study glasses. No obvious trends were observed with regard to the multiple melting steps used to prepare the study glasses, indicating that any potential effects of volatility were below measurable thresholds.

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Performance Assessment of a Generic Nuclear Waste Repository in Shale - 18392

A performance assessment (PA) is an important component of a comprehensive safety analysis for a nuclear waste repository. In a PA, probabilistic simulations of the total repository system are performed, and results are evaluated against performance metrics. Uncertainty and sensitivity analyses help prioritize additional research and model development. The United States Department of Energy (DOE) has been developing a state of the art PA simulation software tool-kit, Geologic Disposal Safety Assessment (GDSA) Framework, that couples increasingly higher fidelity models of subsystem processes into total system PA simulations. Over the past several years, PAs of generic repositories in three geologic media (salt, shale, and crystalline rock) have demonstrated ongoing developments in capability. The current PA of a nuclear waste repository in a generic shale formation showcases GDSA Framework, including capabilities in domain discretization (Cubit), multi-physics simulations (PFLOTRAN), uncertainty and sensitivity analysis (Dakota), and visualization (Paraview). The generic shale reference case considers the disposal of 22,000 metric tons heavy metal of commercial spent nuclear fuel (SNF) in a generic shale formation. PA simulations account for the thermal load and radionuclide inventory of the waste form, components of the engineered barrier system including waste package and buffer, and components of the natural barrier system including the host rock shale and underlying and overlying stratigraphic units. Two repository layouts are considered, one for emplacement of waste packages containing 12 pressurized water reactor (PWR) assemblies, and one for 4-PWR waste package emplacement. Model domains are half-symmetry and contain between 7 and 22 million grid cells. Grid refinement captures the detail of individual waste packages, emplacement drifts, access drifts, and shafts. Simulations are run in a high-performance computing (HPC) environment on 512 or 2048 processes. The governing equations describing coupled heat and fluid flow and reactive transport are solved with PFLOTRAN, an open-source, massively parallel multiphase flow and reactive transport code. Additional simulated processes include waste package degradation; waste form dissolution; radioactive decay and ingrowth in aqueous, solid, and adsorbed phases; and a simple biosphere based on a pumping well. Simulations are run to 106 y; dose and radionuclide concentrations are observed within aquifers at a point approximately 5 km downgradient of the repository. Dakota is used to sample likely ranges of input parameters including waste form and waste package degradation rates and properties of engineered and natural materials to quantify uncertainty in predicted concentrations and sensitivity to input parameters. Given the assumptions of the reference case, results of 12-PWR and 4-PWR simulations are very similar. Dose at the pumping well 5-km downgradient of the repository does not exceed 10-9 Sv/y. I-129 is the largest contributor to dose. The dose due to Cl-36, the next largest contributor, is orders of magnitude smaller. At the pumping well, the concentration of I-129 5-km downgradient does not exceed 10-14 mol/L. Rank correlation coefficients indicate that the maximum concentration of I-129 within the aquifer is sensitive to shale (repository host rock) porosity and aquifer permeability. These results affirm that HPC-capable codes can be used to simulate important multi-physics couplings directly in a total system performance assessment of a geologic repository, and can be used in prioritization of future research and model development. (authors)

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On-Site Decontamination System for Liquid Low Level Radioactive Waste - 13010

This study is based on an evaluation of purification methods for liquid low-level radioactive waste (LLLW) by using natural zeolite. Generally the volume of liquid low-level waste is relatively large and the specific activity is rather low when compared to other radioactive waste types. In this study, a pilot scale column was used with natural zeolite as an ion exchanger media. Decontamination and minimization of LLLW especially at the generation site decrease operational cost in waste management operations. Portable pilot scale column was constructed for decontamination of LLW on site. Effect of temperature on the radionuclide adsorption of the zeolite was determined to optimize the waste solution temperature for the plant scale operations. In addition, effect of pH on the radionuclide uptake of the zeolite column was determined to optimize the waste solution pH for the plant scale operations. The advantages of this method used for the processing of LLLW are discussed in this paper. (authors)

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Thermal Neutron Die-Way-Time Studies for P and DGNAA of Radioactive Waste Drums at the MEDINA Facility

In Germany, radioactive waste with negligible heat production has to pass through a process of quality checking in order to check its conformance with national regulations prior to its transport, intermediate storage and final disposal. Additionally to its radioactive components, the waste may contain non-radioactive chemically toxic substances that can adversely affect human health and pollute the environment, especially the ground water. After an adequate decay time, the waste radioactivity will become harmless but the non-radioactive substances will persist over time. In principle, these hazardous substances may be quantified from traceability and quality controls performed during the production of the waste packages. As a consequence, a research and development program was initiated in 2007 with the aim to develop a nondestructive analytical technique for radioactive waste packages based on prompt and delayed gamma neutron activation analysis (P and DGNAA) employing a DT-neutron generator in pulsed mode. In a preliminary study it was experimentally demonstrated that P and DGNAA is suitable to determine the chemical composition of large samples. In 2010 a facility called MEDINA (Multi Element Detection based on Instrumental Neutron Activation) was developed for the qualitative and quantitative determination of nonradioactive, toxic elements and substances in 200-l steel drums. The determination of hazardous substances and elements is generally achieved measuring the prompt gamma-rays induced by thermal neutrons. Additional information about the composition of the waste matrix could be derived measuring the delayed gamma-rays from short life activation products. However a sensitive detection of these delayed gamma-rays requires that thermal neutrons have almost vanished. Therefore, the thermal neutron die-away-time has to be known in order to achieve an optimal discrimination between prompt and delayed gamma-ray spectra acquisition. Measurements Thermal neutron die-away times have been determined for the following cases: a) the empty chamber, b ) an empty 200-l steel drum, for a 200-l steel drum filled c) with concrete d) with polyethylene and e) with a mixture of polyethylene and concrete by measuring the prompt-gamma ray count rate of relevant isotopes like of {sup 1}H, {sup 10}B, {sup 12}C, {sup 28}Si, {sup 35}Cl, {sup 40}Ca and {sup 56}Fe which are emitted from different parts of the facility and the sample. Additionally, the average die-away-time was determined from the total detector count rate. The neutron generator was operated with a neutron emission of 8x10{sup 7} n.s{sup -1}, a neutron pulse with a length of 250 μs and a repetition time of 5 ms. The spectra were acquired between the neutron pulses over t{sub c}=500 μs after a pre-defined waiting time t{sub D} (multiple of 500 μs). The thermal neutron die-away time was ranging between 0.9 ms and 5 ms according to the sample composition. As an example the measured thermal neutron die-away-time Λ [μs] of a drum filled with concrete is presented. Detailed results of this study will be presented and discussed. (authors)

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

The Performance of Underground Radioactive Waste Storage Tanks at the Savannah River Site: A 60-Year Historical Perspective

The Savannah River Site produced weapons-grade materials for nearly 35 years between 1953 and 1988. The legacy of this production is nearly 37 million gallons of radioactive waste. Since the 1950s, the liquid waste has been stored in large, underground carbon steel waste tanks. During the past 20 years, the site has begun to process the waste so that it may be stored in vitrified and grout forms, which are more suitable for long-term storage. Over the history of the site, some tanks have experienced leakage of the waste to the secondary containment. This article is a review of the instances of leakage and corrosion degradation that the tanks and associated equipment have experienced since the first tanks were built. Furthermore, the activities that the site has taken to mitigate the degradation and manage the service life of the tank for its anticipated lifetime are reviewed.

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