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DOE OSTI2015

Setting an optimal procedure for transferring intermediate radioactive effluents: contribution of laboratory experiments - 15234

The INB35 nuclear facility at the CEA Saclay center is in charge of collecting and treating radioactive effluents generated from site's research activities. The first step of the process consists of an evaporation. The concentrate is an IL radioactive suspension with high salt concentration and suspended matter content. Until the 1990's, the resulting effluents were bitumen coated and since 2012 they have been cemented. During the 1990's, concentrates were stored in 7 tanks, named MA50X with X from 1 to 7. With today's changes to safety regulations, the effluents must now be transferred before tank dismantling. According to the site records, the compositions of the liquid wastes MA501, MA502, MA503, MA504, MA506 and MA507 are similar. From 2013 to 2014, effluents contained in MA502, MA503, MA504 tanks have been transferred. Several difficulties have been encountered: - The tanks were historically different, especially manhole tank access - Since the 1990's, effluent composition has changed due to water evaporation. The salt concentration has increased and insoluble elements have formed a deposit in the bottom of the tanks resulting in high composition heterogeneity - In some tanks, a wall breeze-blade exists dividing the tank into two zones, which complicates the interventions. Major challenges have been met and will be described in the presentation. - To obtain an effluent which is as homogeneous as possible, the duration of mixing has been optimized. - To pump the liquid, it has been necessary to treat the effluent with nitric acid. This step has been optimized thanks to experiments in laboratory and hydrodynamic modelling. - A mobile workshop has been designed in order to perform the effluent transfer. It is set up directly above the tanks and can be moved from one tank to another on rails. Laboratory experiments have been carried out to characterize the suspension and sludge in order to understand its chemical behavior. - The treatment of the effluent has been studied in the laboratory and is now mastered. - The analytical strategy has been optimized. The close collaboration between the analytical laboratory and the dismantling project was a real advantage. Thanks to this, the duration of the last transfer on MA 504 was reduced by half, compared to the first operations. Transfer operations will be described and the differences and similarities between laboratory and in-situ observations will be pointed out. (authors)

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DOE OSTI2015

Main results of the french R and D on minor actinides partitioning, a significant improvement towards nuclear waste reduction - 15618

In the framework of the successive 1991 and 2006 Waste Management Act, French government supported a very significant R and D program on partitioning and transmutation of minor actinides (MA). This program aims to study potential solutions for still minimizing the quantity and the hazardousness of final waste, by MA recycling. Indeed, MA recycling can reduce the heat load and the half-life of most of the waste to be buried to a couple of hundred years, overcoming the concerns of the public related to the long-life of the waste. Within this framework, this paper aims to present the most recent progress obtained in CEA on the development of innovative actinide partitioning hydrometallurgical processes in support of their recycling, either in an homogeneous mode (MA are recycled at low concentration in all the standard reactor fuel) or in an heterogeneous mode (MA are recycled at higher concentration in specific targets, at the periphery of the reactor core). Recovery performances obtained on recent tests in high active conditions of the so-called GANEX and DIAMEX-SANEX process will be presented and discussed in light of the potential P and T scenarios. Finally, recent developments regarding the recycling of the sole Am will be presented as well as the results obtained on highly active solutions for this so-called EXAM process. This set of results gives to the French government a portfolio of potential recycling processes which could be separately and progressively implemented if decided. (authors)

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DOE OSTI2017

ATALANTE research facility-implementation of a rule of fractions for the management of reflecting materials in mass-limited units

ATALANTE, a research unit located in Marcoule, is one of the main Nuclear Facilities of the French CEA. As concerns criticality risk prevention, this facility is divided into work units mostly managed through a mass control mode. The authorized limit is 350 g of fissile material with {sup 239}Pu-H{sub 2}O as reference fissile medium. This mass limit was determined considering a reflection by 20 cm of water. Under these conditions, some neutronic reflectors which are more efficient than water can only be authorized in limited quantities. In ATALANTE the reflecting materials identified as requiring a specific form of management are lead, uranium ({sup 235}U/U{sub total} = 1%), graphite, heavy water and beryllium. Initially a maximum permissible mass was determined for each of these materials taken separately. However, this method requires that when different reflectors are present simultaneously, the sum of the masses of all these reflectors must be less than the limit specified for the most penalizing of them. This rule has proved to be too restrictive for the operators. A new rule has therefore been implemented: the rule of fractions. Criticality calculations were performed to demonstrate that this rule is acceptable. The geometric configurations studied involved a sphere of 350 g of {sup 239}Pu moderated by water and reflected by various masses of 2 to 6 reflectors (successions of concentric shells), followed by 20 cm of water. It was concluded that compliance with the new rule makes it possible to ensure criticality safety in the simultaneous presence of different reflectors. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

CRISTAL V2: New package for criticality calculations

A new release of the CRISTAL French criticality calculation package, named CRISTAL V2, has been developed and intensively validated by IRSN and CEA, in collaboration with AREVA. This package was designed to cover the needs of criticality practitioners in the fields of nuclear fuel cycle facilities and transportation of fissile materials. Based on an optimized modular architecture in an updated software development environment, the design of CRISTAL package V2 integrates new generation of cross-section libraries, new calculation schemes consistent with new applications, and up-to-date calculation codes capabilities. Four calculation routes are currently available allowing multi-group and continuous energy calculations. The validation database is composed of more than 3000 critical experiments, in order to cover a wide range of configurations. The CRISTAL V2 package also includes customer services covering a large scope of activities like documentations, hotline, maintenance and training courses. A significant promotion and an increased recognition referred both nationally and internationally are foreseen, including the provision of the CRISTAL V2 package to the OECD / NEA Data Bank. (authors)

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DOE OSTI2018

Flammability Limits of H2 Based Mixtures in Conditions Relevant to Waste Management in Containment - 18032

The aim of the present work is to assess the risk of explosion in closed containments used for the transportation of nuclear materials or nuclear waste. Indeed, it is very well-known that hydrogen can be produced due to (i) the radiolysis of different materials within the containment, (ii) the thermal decomposition of mainly the organic part in the containment. Since hydrogen has a very low ignition energy and a very wide flammability domain, it is important to determine the risk of ignition of the subsequent mixture produced by the aforementioned mechanisms. The quantity of the hydrogen that can be produced can vary depending on the containment type and on the state of the material/waste being transported. It is then mandatory to have a very good knowledge not only of the flammability domain of hydrogen in air diluted with water vapor. To do so, an experimental work on the flammability domain of multiple ternary mixtures containing H{sub 2}/air/H{sub 2}O{sub vap} is being conducted at the CNRS-ICARE laboratory within collaboration with the CEA. The present work will focus on the experimental determination of the lower flammability limit of H{sub 2}/air/H{sub 2}O{sub vap} using the spherical method. These experiments have been carried out for an initial temperature of 200 deg. C and at three different initial pressures: 0.3, 1 and 2 bar. The monitoring of the flammability conditions has been conducted using fast flame imaging and pressure measurements. (authors)

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DOE OSTI2018

Development of a Dosage System for the Incineration of Radioactive and Particle-laden Liquid - 18080

The alpha incinerator has been operating since 1999 at the CEA Valduc center in France for the treatment of alpha-bearing technological waste. The fact that the incinerator has been working smoothly for 18 years and the experience gained during this period indicate that it can also be used to treat other types of waste, such as Ion-Exchange Resin (IER) suspensions. The aims of an R and D project initiated in 2014 were to develop a dosage system allowing these suspensions to be injected into the heart of the treatment system and to evaluate the performance of the incinerator when these materials are added to technological waste. This article describes the work carried out to develop a dosage system for heavily particle-laden fluids (with more than 50 vol% solid particles), with the aim of delivering waste continuously into the treatment furnace. The proposed device consists of a feeding tank equipped with a stirrer, a piston pump, and a pneumatic transport system. Reliability has been improved by replacing the initial sealing system (a piston with seals), by a very tight (∼4 μm gap), therefore seal-less, sleeve/piston unit. Adjusting the carrying gas flow has optimized the system as a whole, making it safer and more robust. The system has been qualified through more than 1000 hours of operation and the results obtained for the incineration of IER suspensions suggest that industrial transport is feasible and will thus begin soon. (authors)

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DOE OSTI2018

French Innovative Processes in the Field of Thermal Treatment For Decommissioning and Legacy Waste - 18085

In order to optimize decommissioning waste management and better anticipate related issues, French nuclear waste producers CEA and AREVA, and the French Agency in charge of radioactive waste disposal (ANDRA) have decided to collectively develop technologies mastering costs, schedules and dose uptake as well as optimizing waste volumes and repository safety. In this context, several innovative projects have emerged over the past few years to develop specific thermal treatment processes. This paper focus on four new processes, presenting their goals, technical descriptions and first experimental results. (authors)

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DOE OSTI2018

The D and D of Nuclear Facilities: a Systemic Approach for Project Organization and Monitoring - 18099

The Decontamination and Decommissioning (D and D) of Licensed Nuclear facilities takes place under circumstances which make it a very complex field. On the one hand, more and more facilities of widely varying types are concerned, and many D and D projects are currently running or planned. A nuclear facility's life cycle is often very long and the building may be modified or impacted by incidents during its operation. These changes need to be known and taken into account for the D and D plans, in order to ensure safety at all times. In addition, the numerous stakeholders involved in a project express needs which impose many requirements. These encompass various points of view, and data from different detail levels must be managed to satisfactorily meet them. On the other hand, legacy data (including records, plans etc.) is a recurring issue for D and D projects. The quality and availability of relevant data regarding the different phases of the nuclear facility lifetime may be insufficient, for instance due to the lack of traceability practices in the past. The uncertainty level is not always specified and explained, old documents can be damaged and are not necessarily digitized, etc. It is therefore particularly important to define the relevant type of raw data to be collected and traced for D and D projects as early as possible. In this context, CEA (Commissariat a l'energie Atomique et aux energies Alternatives) has been analyzing past D and D projects for many years. The goal is to facilitate reproducibility by generalizing best practices and promoting efficient organizational rules and technical solutions. A method is currently being studied to help project managers to prepare specifications for future D and D operations and improve both the efficiency and the relevance of the decommissioning strategy throughout each project, depending on the expected quality of its end-state. This method first consists in formalizing and specifying the entire set of requirements to be taken into consideration, e.g. technical, operational, logistical, regulatory, financial, organizational, safety and security, human, ethical, or, naturally, environmental aspects. Second, based on these requirements, it will enable the project team to structure, check, and then to demonstrate the coherence and feasibility of the project from both the technological and organizational points of view. Lastly, the method will enable the evaluation of the various possible solutions for the management of dismantling products, for instance, regarding the waste outlets: direct disposal or interim storage. To conclude, the method should permit a constant re-evaluation of the D and D strategy depending on the possible evolution of the stakeholders' requirements, and thus it should optimize the monitoring of D and D projects. (authors)

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DOE OSTI2018

LIBS Measurement in Vitrification of HLW Process: Towards a Development of Online Chemical Analyses - 18290

Research programs are currently underway in the Joint CEA-AREVA Vitrification Laboratory (LCV) to develop and optimize Laser Induced Breakdown Spectroscopy (LIBS) for nuclear waste streams analyses. The laboratory has full scale prototypes with the advantage of working on simulated inactive solutions of fission products. LIBS is an elemental chemical analysis technology that uses a short laser pulse to create a micro-plasma on the sample surface. This analytical technique offers many advantages compared to other techniques. It includes a sample preparation-free measurement and an extremely fast measurement time, from a few seconds to a few minutes. This paper presents the LIBS technique available in the laboratory. The first LIBS analyzes results of liquid samples from the off-gas treatment unit integrated on the full scale vitrification process are also presented and discussed. These complex solutions contain suspended particles and up to 30 chemical elements. A vision of an online LIBS analysis cell is also proposed with the development and associated research steps. (authors)

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DOE OSTI2018

Optimization of LIBS Acquisition Parameters for Chemical Analysis on HLW Liquid Solution - 18291

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DOE OSTI2018

Advanced Waste Management Concepts for Uranium and Plutonium Multi-Recycling - 18533

The multiple recycling of plutonium and uranium will play a major role in gradually achieving a fully closed nuclear fuel cycle in the future, once the first Generation IV fast reactors (FR) are deployed in the NPP fleet. To reach this objective, optimizations of current processes and technologies and, in some cases, the development of innovative concepts are required to meet the scientific, technical and economic challenges imposed by the treatment of spent fuels with an increasingly higher Pu content. Based on the significant past experience acquired by France in LWR and FR MOX fuel treatment, the CEA is running R and D studies to develop new efficient processes for Pu recycling and for FR MOX fuel fabrication. More compact and economically competitive processes, ensuring the safety/security of the facilities (particularly as regards criticality and proliferation resistance) have to be developed taking into account that the amounts of plutonium involved will be one order of magnitude higher than in UOX fuels. New concepts of head-end treatment, dissolution, and separation have been proposed. Consequently, effluent and waste management must be adapted to these processes and to the characteristics of fast reactor MOX fuels. New waste treatment processes are suggested: 1) Cladding waste and undissolved metallic particles could be treated simultaneously in a melting process. The molten metal could be decontaminated by a specific slag and the oxides recovered, in particular PuO{sub 2}, would be recycled at the frond-end of the process, 2) Fission Products and Actinides (FPA) containment glass will also be optimized to accommodate significant amounts of iron, nickel and chromium coming from hull corrosion, and / or decontamination slag coming from the metal melting process. 3) New peraluminous-type glass formulations are under development to accept increased amounts of FPA, as well as iron, nickel, and chromium. The first trials to evaluate the melting process have indicated that metallic compounds from the dissolution residues could be easily incorporated into a molten metal at levels of 1, 2 and 3 w%. (authors)

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DOE OSTI2018

Proven Safe and Effective Process for the Treatment of Bulk Sodium at Shutdown Reactors and Research Facilities - 18601

Elemental sodium is an alkaline metal that reacts when exposed to air or water. This reaction is exothermic in nature and presents the risk of attendant fire as well as hydrogen generation. Within the United States sodium was used in liquid metal fast breeder reactors (LMFBR) during the 1960's and 70's as a heat transfer medium for commercial, experimental and prototype reactors as well as for nuclear energy related research and development experiments due to the element's good thermal conductivity. Shutdown of these reactors and the end of these experiments present a challenge to decommissioning teams as the wastes associated with each project have different characteristics affecting treatment based on the sodium's origin as well as different volumes requiring treatment. French CEA (Commissariat a l'Energie Atomique) and FRAMATOME worked together in the 1980's to develop a treatment process called the NOAH (NaOH - sodium hydroxide) process for the disposal of large volumes of bulk sodium. This paper provides an overview of the treatment process, details past uses of the technology and discusses the process' applications within the United States Department of Energy complex. (authors)

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DOE OSTI2019

NM in HLW Borosilicate Glass: Focus on the High Performances of the IHMM Operated at La Hague Plant - 19019

Ruthenium, palladium and rhodium, which belong to Platinum-Group Metals (PGM), are part of fission products resulting from nuclear fuel reprocessing. All High Level Waste (HLW) vitrification facilities have experienced operational issues that can be related to the presence of Noble Metals (NM) in the melter. These metals have low solubility and modify glass properties such as viscosity or electrical and thermal conductivity. These dense metals also tend to settle to the bottom of the melter. Whatever the vitrification process operated, the specific properties and behavior of NM have a propensity to impact the production capacity by decreasing the glass throughput and increasing downtimes. Orano and the CEA have studied the behavior of NM in borosilicate glass for many years. It has led to several technological improvements and process operation optimizations on the Induction Heated Metallic Melter (IHMM) currently operated at La Hague plant. These enhancements, implemented since active commissioning, have headed to a high efficiency in terms glass production and NM incorporation (more than 21000 canisters produced with up to 3 % of Noble Metals in the glass). This paper sets out the continuous improvement program implemented on the La Hague vitrification lines that has led to outstanding records of operation in terms of production capacity and NM incorporation. The benefits of the close collaboration between R and D, engineering, and operation teams is particularly emphasized. (authors)

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DOE OSTI2020

DEM and MELT: A Modular and Compact In-Can Melting Technology Dedicated to D and D and Remediation Waste - 20033

The DEM and MELT In-Can vitrification process is designed to process intermediate and high level waste coming for D and D and remediation operations. It is developed to treat liquid and solid waste, to produce a small amount of secondary waste and to minimize investment and operating costs. This In-Can vitrification process is also developed with a compact and modular design and can be adapted regarding nuclear operators' needs and requirements. Furthermore, this robust and simple process is flexible enough to accommodate most uncertainties on waste composition. The DEM and MELT process is developed through the French consortium coordinated by CEA which gathers Orano, ECM technologies and Andra. In the frame of this project, Orano has designed a compact and modular layout of the process allowing an efficient in situ implementation, close to the waste to be treated. This paper presents the compact and modular design of the DEM and MELT process and the way it can be used for a short duration, as a decommissioning tool aimed to be dismantled just after the treatment operation. (authors)

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DOE OSTI2020

Feedback on Forty-year Long Clean-up Operations of a Contaminated Soil for Environmental Purpose - 20026

One of the CEA facilities in France is the place of miscellaneous mapping and clean-up operations since the 1980's. The final purpose of the conducted work deals with the environmental remediation of the building. In France, the absence of a regulatory framework for the management of sites polluted by radioactive substances, especially as the lack of release thresholds, has led to develop different approaches of soil radiological characterization. The geostatistical approach is part of them. Preliminary investigations are an unavoidable step prior to clearance and remediation. First, historical analysis enables to define site perimeters to investigate as well as to precisely locate expected contaminated parts. For radioactive pollutants, the media (water, soil, air) and transfer routes participate in the definition of the investigation perimeter and thus also need to be identified through a geological study. At the same time, functional analysis leads to choose the best measurement means which could provide useful information when combined with visual inspection. An equipment well adapted to easy-to-measure radionuclides as gamma emitters, such as gamma probe or on-site gamma spectrometry, is usually selected as a non-destructive assay. Coupled with spatial positions, processed on-site radiological data can be mapped. Data processing consists in mapping and kriging before a geostatistical analysis in order to identify the zones of interest to be targeted. Selected zones are not only linked to high level count rates but also to low level ones so that a pollution-free reference is known in the area. This data processing also enables to categorize waste types according to their origin and contamination levels. An expected volume of different waste then follows as well as the waste characterization equipments. In-depth investigations, e.g. core drillings and samplings, naturally ensue to aim at remediation optimization through the assessment of environmental impact. After clean-up operations, non-destructive assays as well as destructive samplings enable to check the effective clearance. Following this methodology in the present remediation work, results of the first non-destructive assay campaign highlighted a long-lived transuranic radionuclides contamination in the concrete flagstone. Successive non-destructive assays were then performed in the defined zones of interest. Count rates measured with a surface probe were first mapped. The best localization of destructive assays, i.e. core drillings at a 1 meter depth, came from this surface mapping followed by a geostatistical analysis. The concrete flagstone was totally removed. Then, the depth of soil to remove under the flagstone was optimized from the results of radiological activities measured in core drillings samples. A new surface mapping was then drawn after this partial soil excavation. The sand and demolition rubble samples, homogeneously constituted from the flagstone fragments, were analyzed by gamma spectrometry. Meanwhile, waste drums and other containers were also measured by gamma spectrometry and passive neutron measurement devices. All these results have brought miscellaneous pieces of information. In this context, major issues appear in terms of physical constraints such as the premises tininess as well as of soil sampling techniques and in terms of measurement performances assessment and the representativeness of homogeneous samples. The choice of radiological soil and waste characterization devices is described in the light of enhanced performances by keeping in mind radiation protection requirements and a willing to always optimize waste categories (very low level waste) and volume. The improvement of the characterization methods appears through this forty-year long work while sharing the feedback of encountered difficulties. (authors)

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DOE OSTI2020

In-Can Incineration and Vitrification Process for the Treatment of Transuranic Wastes - 20035

Technological alpha contaminated wastes are produced by the nuclear fuel cycle management industry, essentially from MOX fuel production units. These Plutonium Contaminated Materials (PCM) contain a mixture of organic, metal and mineral materials, contaminated with uranium, plutonium and americium. One treatment scenario under study is to incinerate the organic materials, melt the metallic fraction and incorporate the mineral fraction (including ashes from organics incineration) into a glass matrix without sorting or pretreatment (grinding for instance). A scale 1 pilot of this PIVIC process (Process for Incineration and Vitrification In Can) has been commissioned in 2018 at the CEA R and D vitrification facility in Marcoule site. A R and D program aiming to demonstrate the feasibility of this process is currently on going. This paper starts with a description of PIVIC process with explanation of the technologies involved. Then, main results of the first year of operation with the technological demonstration goal are presented. (authors)

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DOE OSTI2020

Characterization Strategy of a Limited Number of Radioactive Waste Effluent Tanks in View of Decommissioning a Large Number of Tanks - 20286

In order to organize the D and D of the former effluent treatment facility at the CEA Cadarache Center (France), the key data to gather through the facility characterization are the following:: volumes and masses, liquid/solid distribution and specific activity. By studying the treatment processes and the history of the facility, families of tanks have been defined and a reference tank chosen for investigation in each family. The intrusive investigations (video, residue thickness, dose rate..), the gamma spectrometry measurements carried out on site on the residue samples and radiochemical results collected in laboratories, led us to conclude that there was a relative radiological homogeneity (for specific activity and Am-241/Cs-137 ratio) within each tank. The next step consisted in extrapolating the specific activity evaluated for the reference tanks, to the other tanks of each family by using data such as geometry, dose rate or the function of the tanks within the process. Thus, a cartography of the 130 tanks of the facility is available. Finally, the conclusions drawn from the radiological homogeneity of the tanks and their affiliation to homogeneous, well-identified tank clusters lead to a simple and efficient operational approach for the D and D of the 130 tanks of the facility. Indeed, because of the residue characteristics (homogeneous activity and strong differentiation of Cs/Am ratio), a sample collected when opening the tank, and measured on site by gamma spectrometry will allow us to validate the operational design and the emptying techniques defined during the D and D study phase thanks to the characterization of a limited number of reference tanks. (authors)

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DOE OSTI2020

Vitrification of Inactive Incineration Ash in a Full Industrial Scale In Can Melter - 20502

In the framework of the H2020 THERAMIN project funded by the European Commission, CEA performed a pilot-scale trial consisting of the vitrification of inactive ash (Al, Ca, Si, Zn, Bi, K, Mg, etc.) using the In- Can Melter technology. These ashes were coming from incineration of simulated technological wastes such as cotton, latex, neoprene, polyethylene. The borosilicate glass feasibility was first evaluated at a laboratory scale (10-100 g of glass). The volatile behavior of fly ash was controlled by pelletizing to avoid pipes clogging when performing the pilot scale test. A 10 wt.% inorganic binder was added to obtain cohesive pellets. The glass frit composition was selected and its ratio was fixed at 50 wt.%. After this step, a bench test (1 kg of glass) was done to verify the absence of container corrosion. Finally, it was followed by a pilot scale test (50 kg of glass) with 75,000 ash pellets used. The same glass microstructure and chemical composition were observed at each scale. Typical crystals of apatites (Ca, P), zincochromites (Zn, Cr, Al) and bismuth alloys (Bi, Sb) dispersed into a borosilicate glass matrix were observed. The chemical durability of the wasteform was measured during 28 days in a 90 deg. C water. Although there were durable crystalline phases, the hydrolysis rate of the vitreous part is relatively high because of the B and Na contents. However, leach rate profiles suggest that an alteration layer will form, leading to an alteration rate decrease. (authors)

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