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The project goals are to understand the current types of contamination and their impact on conversion yields, to develop decontamination strategies, and to understand the TEA/LCA tradeoffs for these proposed strategies.
Municipal solid waste (MSW) offers a significant opportunity for energy conversion pathways. Paper and plastic waste can be converted and upgraded to fuels through low- and high-temperature conversion, respectively. The objective of this study is to develop MSW paper and plastic decontamination methods to increase conversion efficiencies.
The UP1 plant in Marcoule reprocessed nearly 20,000 tons of used natural uranium gas cooled reactor fuel coming from the first generation of civil nuclear reactors in France. During more than 40 years, the decladding operations produced thousands of tons of processed waste, mainly magnesium and graphite fragments. In the absence of a French repository for the graphite waste, the graphite sludge content of the storage pits had to be retrieved and transferred into a newer and safer pit. After an extensive R and D program, the equipment and process necessary for retrieval operations were designed, built and tested. The innovative process is mainly based on the use of two pumps (one to capture and the other one to transfer the sludge) working one after the other and a robotic arm mounted on a telescopic mast. A dedicated process was also set up for the removal of the biggest fragments. The retrieval of the most irradiating fragments was a challenge. Today, the first pit is totally empty and its stainless steel walls have been decontaminated using gels. In the second pit, the sludge retrieval and transfer operations have been almost completed. Most of the non-pumpable graphite fragments has been removed and transferred to a new storage pit. After more than 6 years of operations in sludge retrieval, a lot of experience was acquired from which important 'lessons learned' could be shared. (authors)
Peculiar binding of Cesium to the soil clay minerals remained the major obstacle for the immediate Cs-decontamination of soil and materials containing clay minerals like sludge. Experiments for the removal of Cesium from soil and ash samples from different materials were performed in the lab scale. For soil and sludge ash formed by the incineration of municipal sewage sludge, acid treatment at high temperature is effective while washing with water removed Cesium from ashes of plants or burnable garbage. Though total removal seems a difficult task, water-washing of wood-ash or garbage-ash at 40 deg. C removes >90% radiocesium, while >60% activity can be removed from soil and sludge-ash by acid washing at 95 deg. C. (authors)
The Korea Atomic Energy Research Institute (KAERI) developed a Hydrazine base reductive metal ion decontamination (HyBRID) and related pre-oxidation techniques for use in decommissioning a nuclear facility. Inorganic chemicals were used to increase the integrity of the radioactive waste in a repository. Decomposition and precipitation of residual reagents with inorganic chemicals were introduced to minimize the final waste volume. The waste volume was calculated theoretically and demonstrated in a small-scale waste treatment system and bench-scale candle filtration system. The total waste volume of the HyBRID process was less than 1/10 of that of commercial processes, and spent ion-exchange resins were significantly reduced to 5% of that of commercial processes. (authors)
Canberra Industries, Inc. [CI], Now Mirion Technologies (Canberra) Inc. [MTC], has developed and validated a new Flexible Container Bag or Feed Roll Assay System 'FRAS' for Interim Storage Facilities [ISF] etc. There is an estimated 22 million cubic meters of soil and vegetation from the decontamination efforts around Fukushima. Removed soil and vegetation waste was put into flexible containers called Super Sacks [SS]. These SSs will be transported to ISF by trucks. These SSs will be divided into three categories according to total Cs radioactivity level (RL). This level is 'higher than 100,000 Bq/kg', '100,000>RL>8,000 Bq/kg' and '8,000 > RL >3,000 Bq/kg', respectively. The main uncertainty for SS measuring is internal inhomogeneity and heterogeneous source distribution. Japan Atomic Energy Agency [JAEA] has developed the Scattering Gamma Equivalent Model [SGE model] for decreasing uncertainty derived from these factors. This method was validated by Mirion Technologies (Canberra) KK [MCKK] and JAEA. According to the validation test, the SGE model can reduce the uncertainty of SS measurement from ± 25% to ± less than 15%. (authors)
The examples of application of imaging and remote sensing technologies during decontamination and decommissioning works in nuclear environment at Kurchatov Institute site are presented. The systems which realized these technologies are: radiometric system installed on mechanical arm of Brokk remote controlled vehicle, collimated scanning spectrometric system and portable gamma cameras of different types. Objects of application: facilities of research reactors at characterization and under dismantling, temporal radwaste storages during discharge and remediation. (authors)
Irradiated graphite is one of the most significant, large volume waste streams in the UK. After shut down of gas cooled reactors there will be ∼96,000 tonnes of nuclear graphite arising from pressure vessels, sealed or unsealed stacks, temporary surface storage and in silos, which may account for up to 30% by ILW volume of any future UK geological disposal facility. High temperature molten salt treatment (HMST) could be considered as a partitioning process of the activation and fission products from irradiated graphite. The main objective of the research is to optimize a specific graphite treatment technology compatible with older, current and future reactors to provide a safe and effective process to decontaminate graphite and reduce the waste inventory. In order to reach that purpose principal radionuclides contained in irradiated samples from Magnox reactors were investigated by germanium (Ge) gamma spectrometry. The next key tasks were to employ treatment at 450 deg. C in LiCl-KCl eutectic which included the following procedures: electrochemical cleaning of the salt, initial cyclic voltammetry (CV) of graphite followed by several steps of chrono-potentiometry (CP) with different current applied and CV was taken after each step. Once the treatment was established, experiments investigating the electro-refinement of the resultant salt mixtures were conducted as well as the comparison of gamma spectroscopy results of graphite before and after the treatment providing the result of total activity reduction of around 60%. (authors)
The Arvia Process is being adapted for use in the nuclear industry for the decontamination of oils containing radionuclides. This waste treatment process uses adsorption followed by electrochemical oxidation as an alternative technique to conventional incineration. A study has been carried out on components of the Arvia Process with the aim of understanding the changes in chemistry due to ionizing radiation. Samples of adsorbent in water have been exposed to gamma-rays from a Co-60 irradiator at absorbed doses of 3, 19, and 304 kGy. Preliminary results show that the concentration of inorganic ions and carbon in the water increased as a function of dose. This increase suggests that degradation of the adsorbent may take place at high doses. Further characterization of the microstructure and adsorption studies post-irradiation are planned. (authors)
On February 14, 2014, americium and plutonium contamination was released in the Waste Isolation Pilot Plant (WIPP) salt caverns. Several practical, easily deployable methods of decontaminating WIPP salt, using a surrogate contaminant and americium ({sup 241}Am), were developed and tested. The effectiveness of the methods is evaluated qualitatively, and to the extent practical, quantitatively. Of the methods tested (dry brushing, vacuum cleaning, water washing, strippable coatings, and mechanical grinding), the most practical seems to be water washing. Effectiveness is very high, and water washing is easy and rapid to deploy. The amount of wastewater produced (2 L/m{sup 2}) would be substantial and may not be easy to manage, but the method is the clear winner from a usability perspective. Removable surface contamination levels (smear results) from water washed coupons found no residual removable contamination. Thus, whatever contamination is left is likely adhered to (or trapped within) the salt. The other option that shows promise is the use of a fixative barrier. Bartlett Nuclear, Inc.'s Polymeric Barrier System proved the most durable of the coatings tested. The coatings were not tested for contaminant entrapment, only for coating integrity and durability. (authors)
Deactivation and decommissioning (D and D) is the process of taking an active/excess/abandoned facility to a final disposition end state. D and D presents unique hazards that must be addressed from a safety, programmatic, environmental, and technological standpoint. This definition is especially applicable to Savannah River Site's 235-F facility. The facility was constructed in the 1950's as part of the weapons complex materials production and fabrication. The most significant quantities of residual contamination remaining in 235-F building is in the form of Plutonium-238. A Decision Model was created using MatLab to work hand-in-hand with a product list containing commercially available D and D products. This model was created to further enhance and accelerate the decontamination process. It is envisioned that D and D engineers will be able to use this Decision Model in the field via a mobile device. It allows the user to select four inputs to match their situation. The model searches the database and returns the products that fit the criteria selected by the user. This Decision Model will be made available in a web-based application and as a mobile application to ensure end users can access it when needed. The MatLab code recognizes what position in the text box is selected and then this input is converted to the corresponding string. The string is then compared with the information in the product list in the specified column in order to reduce processing time. Then the products that contain the string inputted are displayed on the table and are separated by category. The web-based version of the model will receive three specified criteria and sub-criteria from the user and update the table with products that fit each criteria. A web-based version of this Decision Model is under way and will be set up on the D and D KM-IT platform. There will be two versions of this mobile application: for devices with internet connection and for those that do not have internet connection. This will ensure end users have access to the Decision application at any time or place during the D and D process. A mobile version of the model will also be developed in the near future.
On May 2012, Best Medical Belgium S.A. (BMB), a manufacturer of pharmaceutical products located on the nuclear site of Fleurus, was declared bankrupt. According to the Belgian law, ONDRAF/NIRAS (Belgian Radioactive Waste Management Agency) was entrusted with the clean-up and decommissioning operations of the former BMB facilities, nowadays known as 'ONDRAF Site Fleurus' (ONSF). In June 2015, ONSF launched a Decontamination and Decommissioning (D and D) program, for which Tractebel: - Supports the definition of the clean-up and decommissioning strategies; - Drafts the Final Decommissioning Plan (FDP) and Safety Analysis Report (SAR) that will be submitted to FANC/AFCN (Belgian Federal Agency for Nuclear Control) in order to get the License for decommissioning; - Drafts the technical specifications of the public tendering process related to the D and D program; - Ensures the day to-day follow-up of the D and D activities on site. Four nuclear buildings are part of the D and D activities. One of them, building 14 (B14), represents the most challenging and extensive task with the clean-up of highly contaminated areas and its further decommissioning up to the final state. This building features two cyclotrons (CGR and IBA cyclotrons) successively used to produce radioisotopes, as well as hot cells and glove boxes used for their purification and the conditioning of the final pharmaceutical product. The present paper aims at presenting the most recent development steps of the clean-up phase regarding the adaptations of existing facilities and the upcoming activities in preparation of the decommissioning activities. To allow the sorting and the treatment of the radioactive material identified in the physical and radiological inventory, several workshops have been implemented within the B14 building in order to deal with radioactive materials arising from the clean-up and D and D activities. Following the conceptual design phase, Tractebel has actively monitored the on-site activities for the construction of those workshops such as: - The modification of the civil works; - The revamping of the ventilation system; - The installation of dedicated lifting equipment and workstations equipped with cutting devices; - The assembly of a docking station for filling the radioactive packages. The removal of the activated concrete surrounding the IBA cyclotron requires the erection of a containment structure and an extension structure. The containment structure, composed of light metallic beams, is located above the cyclotron and its irradiation zones. For operational security reasons and working efficiency, the cutting process of the concrete ceilings and walls will be performed from the top inside the containment structure by means of an eight tons overhead bridge crane. Concrete blocks resulting from the cutting process will then be lifted, without any loss of containment, towards the extension zone of the nuclear zone of the IBA cyclotron for further resizing, sorting and evacuation towards the identified waste routes. To prepare this task which will be performed during the decommissioning phase, Tractebel sets up the necessary technical specifications. (authors)
A number of commercially available products, designed for use as strippable coating and fixatives for decontamination purposes, were tested to determine their tolerability to high radiation fields and how they potentially degrade. Each product was cured and exposed to γ-radiation from a Co-60 source, up to a total dose of 500 kGy. Obvious physical colour changes and deformities were detected by visual inspection, and images captured by scanning electron microscope (SEM) showed major differences post-irradiation in some coatings. Other physical changes to the coatings were measured using Fourier-transform infrared spectroscopy (FT-IR), Raman spectroscopy, and energy-dispersive X-ray spectroscopy (EDS); both pre- and post-irradiation. FT-IR suggested that some chemical changes were taking place in the coatings, in particular the C-H bonds and the fingerprint region appeared to be drastically altered in some coatings. EDS data indicated an increase in surface oxygen content as a result of irradiation in the majority of the samples. Additionally, coating samples were subjected to analysis by gas chromatography mass spectrometry (GCMS) in order to determine the evolved gases. The GC-MS results show that short chain hydrocarbons are produced to some degree in all the coatings during irradiation. Finally, three of the best performing coatings were selected and were irradiated using an ion beam, mimicking Pu alpha particles (with energy of around 5.25 MeV). Using the Dalton Cumbrian Facility's 5 MV tandem accelerator, coating samples were irradiated by ∼5.25 MeV He{sup 2+} particles to total dose of approximately 500 kGy. Samples were analysed using SEM, FT-IR, Raman and EDS, as was done for the gamma irradiated samples, allowing a comparison to be made between α-particle and γ-ray damage. Results from the alpha and gamma irradiations showed many similarities for the comparable assessments. A number of spectral changes observed in the FT-IR analysis were consistent across both γ-ray and α-particle irradiated samples. While there were some differences, these may be attributed to the conditions under which the irradiations were carried out. The observed similarities may allow for the inference of alpha degradation by Co-60 gamma irradiation as a more easily accessible method for determining the extent of radiolytic damage. Furthermore, gamma irradiation allows for additional analysis that is currently otherwise not possible to achieve with ion-beam exposure. (authors)
Lawrence Livermore National Security, LLC (LLNS) is the prime contractor responsible for managing and operating Lawrence Livermore National Laboratory (LLNL). The mission of the LLNL is to strengthen security of the United States through development and application of world-class science and technology to enhance the nation's defense, reduce the global threat from terrorism and weapons of mass destruction, and respond with vision, quality, integrity, and technical excellence to scientific issues of national importance. To accomplish its mission, LLNL must plan and manage its campus space and facilities to optimize use of its relatively small one-square-mile footprint. LLNL currently has numerous excess facilities that require decontamination and demolition (D and D). The LLNL Legacy Facility Program is responsible for stewardship and risk reduction programs to effectively manage these excess assets at LLNL. The Legacy Facility Program identified 10 excess facilities at LLNL that were candidates for demolition based on risks from further degradation of the facilities, potential exposure to hazardous materials, or need for removal to make way for new program facilities. The buildings range from a structure built in 1943 to serve as a US Navy drill hall during World War II, to a facility that provided neutrons to study material properties for the fusion energy program. The National Nuclear Security Administration (NNSA) appropriated FY19 funding for the development of Class 3 estimates for these 10 excess facilities that would be used to develop a Program Management Plan for funding consideration. Class 3 estimates are not conceptual, nor overly detailed, but adequate to be used for budget and appropriation purposes. Using existing strategic sourcing agreements, LLNL was able to organize a team of senior subject matter experts (SMEs) with diverse technical backgrounds across the Department of Energy (DOE) Complex to complete extensive process knowledge (PK) reviews and walk-downs of each excess facility and to deliver accurate and complete estimates under an expedited schedule (i.e., within eight weeks). To facilitate the expedited schedule, LLNL was fully prepared for the arrival of the JGMS team, which consisted of the following companies: J.G. Management Systems, Inc. (JGMS); Strata-G, LLC; and Michael Baker International. LLNL provided background and PK information on the excess facilities, pictures, supporting historical documentation, and radiological survey history as well as escorted access to the facilities. The JGMS team has previous experience supporting Program Management Plans for excess facilities at the Y-12 National Security Complex. The team mobilized to LLNL and performed the facility walk-downs from January 22-25, 2019. Using a methodical evaluation process, expertise, and experience, the team was able to deliver walk-down draft reports and estimates to LLNL before their February 8, 2019 deadline. LLNL then in turn reviewed, completed the estimates by adding waste management costs and other LLNL costs, and delivered the estimates to NNSA by February 15, 2019. The completed final reports, including draft D and D estimates, were delivered to LLNL before their March 6, 2019 deadline. The work was successfully completed ahead of the expedited schedule required by LLNL, resulting in a commendation from the client. This paper describes how the partnership and collaboration as a fully integrated team comprised of the prime contractor, LLNS, and the JGMS team successfully met the needs of the US government, resulting in the performance of expedited site walk-downs, the development of summary evaluation reports, and the preparation of draft Class 3 estimates and preliminary schedules for the D and D of 10 excess facilities at LLNL. The efforts of the team will enable the D and D of the first excess facility in the 2020-2021 time frame. (authors)
99 Tc contamination at legacy nuclear sites is a serious and unsolved environmental issue. The selective remediation of 99 TcO 4 - in the presence of a large excess of NO 3 - and SO 4 2- from natural waste systems represents a significant scientific and technical challenge, since anions with a higher charge density are often preferentially sorbed by traditional anion exchange materials. We present a solution to this challenge based on a stable cationic metal-organic framework, SCU-102 (Ni 2 (tipm) 3 (NO 3 ) 4 ), which exhibits fast sorption kinetics, a large capacity (291 mg g 1 ), a high distribution coefficient, and, most importantly, a record-high TcO 4 - uptake selectivity. This material can almost quantitatively remove TcO 4 - in the presence of a large excess of NO 3 - and SO 4 2- . Finally, decontamination experiments confirm that SCU-102 represents the optimal Tc scavenger with the highest reported clean-up efficiency, while first-principle simulations reveal that the origin of the selectivity is the recognition of TcO 4 - by the hydrophobic pockets of the structure.
The use of surface microdischarge (SMD) for the bacterial decontamination of raw produce was evaluated. With 1 min of SMD treatment, >2 logarithmic reduction in Escherichia coli O157:H7 was consistently observed. The scanning electron microscopy of E. coli O157:H7 show that SMD damages the cell membrane, leads to cell expansion, and eventually lysis. The attenuated total reflectance-Fourier-transform infrared spectroscopy characterization of E. coli O157:H7 and lipopolysaccharides (LPS) shows that SMD causes (a) the oxidation of cellular components by forming COOH and COO - groups inside and on the cell wall, and (b) the modification of polysaccharides and phosphorus-containing groups found in phospholipids and DNA. Further characterization with X-ray photoelectron spectroscopy suggests SMD mainly modifies the O-chain and core-polysaccharide part of LPS.
Multiple tests of a co-decontamination (CoDCon) tributyl phosphate (TBP) based solvent extraction flowsheet are being performed to determine the uncertainties associated with maintaining a target U/Pu ratio in the Pu-containing product. The first two flowsheet tests used simple dissolved fuel simulants containing only U and Pu, but increasingly more complex simulants will but used in subsequent tests, with the ultimate goal of performing tests with irradiated fuel. The flowsheet is designed so as not to create a purified Pu stream, but rather a mixture of U and Pu with a U/Pu mass ratio of 7/3. Uranium and Pu are first co-extracted into the TBP solvent, and then U(IV) is used to strip the Pu from the loaded solvent. A fresh TBP solvent stream is introduced at the same stage the Pu-containing aqueous stream exits. The flowrate of the fresh TBP solvent stream is adjusted so that sufficient U bleeds over into the Pu product to produce the desired U/Pu ratio. The system is monitored in real time using optical spectroscopy, which allows for control of the U/Pu ratio through adjustment of the fresh TBP solvent flowrate. This paper will describe the results of the first two CoDCon flowsheet tests.
The Single Cell Data Decontamination Pipeline is a fully-automated software tool which classifies unscreened contigs from single cell datasets through a combination of homology and feature-based methodologies using the organism's nucleotide sequences and known NCBI taxonomony. The software is freely available to download and install, and can be run on any system.