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At least 163 records · Page 9
All sourcesNASA NTRSDOE OSTI
DOE OSTI2012

Pre-conceptual design study of ASTRID core

In the framework of the ASTRID project at CEA, core design studies are performed at CEA with the AREVA and EDF support. At the stage of the project, pre-conceptual design studies are conducted in accordance with GEN IV reactors criteria, in particularly for safety improvements. An improved safety for a sodium cooled reactor requires revisiting many aspects of the design and is a rather lengthy process in current design approach. Two types of cores are under evaluation, one classical derived from the SFR V2B and one more challenging called CFV (low void effect core) with a large gain on the sodium void effect. The SFR V2b core have the following specifications: a very low burn-up reactivity swing (due to a small cycle reactivity loss) and a reduced sodium void effect with regard to past designs such as the EFR (around 2$ minus). Its performances are an average burn-up of 100 GWd/t, and an internal conversion ratio equal to one given a very good behavior of this core during a control rod withdrawal transient). The CFV with its specific design offers a negative sodium void worth while maintaining core performances. In accordance of ASTRID needs for demonstration those cores are 1500 MWth power (600 MWe). This paper will focus on the CFV pre-conceptual design of the core and S/A, and the performances in terms of safety will be evaluated on different transient scenario like ULOF, in order to assess its intrinsic behavior compared to a more classical design like V2B core. The gap in term of margin to a severe accident due to a loss of flow initiator underlines the potential capability of this type of core to enhance prevention of severe accident in accordance to safety demonstration. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

The Phenix ultimate natural convection test

The French sodium cooled fast reactor Phenix was shut down in 2009 after 35 years of operation. Before decommissioning, a final set of tests were performed by the CEA during 9 months. Several topics were involved such as thermal hydraulics, core physics and fuel behaviour. Among these ultimate experiments, two thermal hydraulic tests were performed: an asymmetrical test consisting in a trip of one secondary pump and a natural convection test in the primary circuit. Recognizing the unique opportunity offered by these Phenix ultimate tests, IAEA decided in 2007 to launch a Coordinated Research Project (CRP) devoted to benchmarking analyses with system codes on the Phenix natural convection test. One objective of the natural convection test in Phenix reactor is the assessment of the CATHARE system code for safety studies on future and advanced sodium cooled fast reactors. The aim of this paper is to describe this test, which was performed on June 22-23, 2009, and the associated benchmark specifications for the CRP work. The paper reminds briefly the Phenix reactor with the main physical parameters and the instrumentation used during the natural convection test. After that, the test scenario is described: - initial state at a power of 120 MWth, - test beginning resulting from a manual dry out of the two steam generators, - manual scram, - manual trip on the three primary pumps without back-up by pony motors, - setting and development of natural convection in the primary circuit, in a first phase without significant heat sink in the secondary circuits and in a second phase with significant heat sink in the secondary circuits, by opening the casing of steam generators to create an efficient heat sink, by air natural circulation in the steam generators casing. The benchmark case ends after this second phase, which corresponds to the experimental test duration of nearly 7 hours. The paper presents also the benchmark specifications data supplied by the CEA to all participants of this CRP in order to perform calculations (core, primary circuit, primary pumps, IHX, shutdown system, operating parameters, test scenario and real test conditions). Finally, main test results and analyses are presented including the evolution of the core and of the heat exchangers inlet and outlet temperatures, and some local temperature measurements. The natural convection has been easily set up in the pool type reactor Phenix with different boundary conditions at the secondary side, with or without heat sink. The data obtained during this unique test represent some very useful and precious results for the development of SFR in a wide range of thematic such as numerical methods dedicated to thermal-hydraulics safety analyses (system codes, CFD codes and coupling system and CFD codes) and instrumentation. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2013

Industrial Program of Waste Management - Cigeo Project - 13033

The French Planning Act of 28 June 2006 prescribed that a reversible repository in a deep geological formation be chosen as the reference solution for the long-term management of high-level and intermediate-level long-lived radioactive waste. It also entrusted the responsibility of further studies and design of the repository (named Cigeo) upon the French Radioactive Waste Management Agency (Andra), in order for the review of the creation-license application to start in 2015 and, subject to its approval, the commissioning of the repository to take place in 2025. Andra is responsible for siting, designing, implementing, operating the future geological repository, including operational and long term safety and waste acceptance. Nuclear operators (Electricite de France (EDF), AREVA NC, and the French Commission in charge of Atomic Energy and Alternative Energies (CEA) are technically and financially responsible for the waste they generate, with no limit in time. They provide Andra, on one hand, with waste packages related input data, and on the other hand with their long term industrial experiences of high and intermediate-level long-lived radwaste management and nuclear operation. Andra, EDF, AREVA and CEA established a cooperation agreement for strengthening their collaborations in these fields. Within this agreement Andra and the nuclear operators have defined an industrial program for waste management. This program includes the waste inventory to be taken into account for the design of the Cigeo project and the structural hypothesis underlying its phased development. It schedules the delivery of the different categories of waste and defines associated flows. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

The Creation of a French Basic Nuclear Installation - Description of the Regulatory Process - 13293

CEA is a French government-funded technological research organization. It has to build a medium-level waste interim storage facility because the geological repository will not be available until 2025. This interim storage facility, called DIADEM, has to be available in 2017. These wastes are coming from the research facilities for spent fuel reprocessing and the dismantling of the most radioactive parts of nuclear facilities. The CEA handles the waste management by inventorying the needs and updating them regularly. The conception of the facility is mainly based on this inventory. It provides quantity and characteristics of wastes and it gives the production schedule until 2035. Beyond mass and volume, main characteristics of these radioactive wastes are chemical nature, radioisotopes, radioactivity, radiation dose, the heat emitted, corrosive or explosive gas production, etc. These characteristics provide information to study the repository safety. DIADEM mainly consists of a concrete cell, isolated from the outside, wherein stainless steel welded containers are stored, stacked in a vertical position in the racks. DIADEM is scheduled to store three types of 8 mm-thick, stainless steel cylindrical containers with an outside diameter 498 mm and height from 620 to 2120 mm. DIADEM will be a basic nuclear installation (INB in French) because of overall activity of radioactive substances stored. The creation of a French basic nuclear installation is subject to authorization according to the French law No. 2006-686 of 13 June 2006 on Transparency and Security in the Nuclear Field. The authorization takes into account the technical and financial capacities of the licensee which must allow him to conduct his project in compliance with these interests, especially to cover the costs of decommissioning the installation and conduct remediation work, and to monitor and maintain its location site or, for radioactive waste disposal installations, to cover the definitive shut-down, maintenance and surveillance expenditure. The authorization is issued by a decree adopted upon advice of the French Nuclear Safety Authority and after a public enquiry. In accordance with Decree No. 2007-1557 of November 2, 2007, the application is filed with the ministries responsible for nuclear safety and the Nuclear Safety Authority. It consists of twelve files and four records information. The favorable opinion of the Nuclear Safety Authority on the folder is required to start the public inquiry. Once the public inquiry is completed, the building permit is issued by the prefect. (authors)

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

Advanced instrumentation and analysis methods for in-pile thermal and nuclear measurements: from out-of-pile studies to irradiation campaigns

Research and development on nuclear fuel behavior under irradiations and accelerated ageing of structure materials is a key issue for sustainable nuclear energy in order to meet specific needs by keeping the best level of safety. A new Material Testing Reactor (MTR), the Jules Horowitz Reactor (JHR) currently under construction in the South of France in the CEA Cadarache research centre will offer a real opportunity to perform R and D programs and hence will crucially contribute to the selection, optimization and qualification of innovative materials and fuels. To perform such programs advanced accurate and innovative experiments, irradiation devices that contain material and fuel samples are required to be set up inside or beside the reactor core. These experiments needs beforehand in situ and on line sophisticated measurements to accurately reach specific and determining parameters such as thermal and fast neutron fluxes, nuclear heating and temperature conditions to precisely monitor and control the conducted assays. Consequently, since 2009 CEA and Aix-Marseille University collaborate in order to design and develop a new multi-sensor device which will be dedicated to measuring profiles of such conditions inside the experimental channels of the JHR. These works are performed in the framework of two complementary joint research programs called MAHRI-BETHY and INCORE. These programs couple experimental studies carried out both out-of nuclear fluxes (in laboratory) and under irradiation conditions (in OSIRIS MTR reactor in France and MARIA MTR reactor in Poland) with numerical works realized by thermal simulations (CAST3M code) and Monte Carlo simulations (MCNP code). These programs deal with three main aims. The first one corresponds to the design and/or the test of new in-pile instrumentation. The second one concerns the development of advanced calibration procedures in particular in the case of one specific sensor: a differential calorimeter used to quantify nuclear heating. The last one consists in the development of accurate measurement and analysis methods. The paper will be dedicated to a complete review of the experimental and numerical works performed since 2009 thanks to two parts. The first part will detail a new thermal approach implemented to improve nuclear heating measurements by radiometric calorimeters. New experimental tools (calorimeter prototypes and set-ups such BETHY Bench) developed to perform preliminary out-of-pile studies under suitable conditions will be presented (temperature and velocity of the external cooling fluid, heat source localization and intensity inside the calorimetric cells). Then the response of two kinds of sensors, their calibrations curves and their thermal behaviors will be compared for various parameters. Finally validated numerical thermal and Monte Carlo works will be discussed to propose new improvements. The second parts of the paper will focus on works realized in order to design, develop and test the first prototype of the multi-sensor device called CARMEN [7-9]. The two mock-ups dedicated respectively to neutron measurements and photon measurements will be detailed. The results obtained during two irradiation campaigns inside the periphery of OSIRIS reactor will be shown. The new analysis method will be discussed. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Calculation of neutron and gamma fluxes in support to the interpretation of measuring devices irradiated in the core periphery of the OSIRIS Material Testing Reactor

Technological irradiations carried out in material testing reactors (MTRs) are used to study the behavior of materials under irradiation conditions required by different types of nuclear power plants (NPPs). For MTRs, specific instrumentation is required for the experiment monitoring and for the characterization of irradiation conditions, in particular the flux of neutrons and photons. To measure neutron and photon flux in experimental locations, different sensors can be used, such as SPNDs (self-powered neutron detectors), SPGDs (self-powered gamma detectors) and ionization chambers. These sensors involve interactions producing ultimately a measurable electric current. Various sensors have been recently tested in the core periphery of the OSIRIS reactor (located at the CEA-Saclay center) in order to qualify their responses to the neutron and the photon flux. One of the key input data for this qualification is to have a relevant evaluation of neutron and gamma fluxes at the irradiation location. The objective of this work is to evaluate the neutron and the gamma flux in the core periphery of the OSIRIS reactor. With this intention, specific neutron-photonic three-dimensional calculations have been performed and are mainly based on the TRIPOLI-4{sup R} three-dimensional continuous-energy Monte Carlo code, developed by CEA (Saclay Center) and extensively validated against reactor dosimetry benchmarks. In the case of the OSIRIS reactor, TRIPOLI-4{sup R} code has been validated against experimental results based on neutron flux and nuclear heating measurements performed in ex-core and in-core experiments. In this work, simultaneous contribution of neutrons and gamma photons in the core periphery is considered using neutron-photon coupled transport calculations. Contributions of prompt and decay photons have been taken into account for the gamma flux calculation. Specific depletion codes are used upstream to provide the decay-gamma sources required by TRIPOLI-4 calculations. In this paper, the neutron-photonic calculation scheme used for the neutron-gamma flux evaluation in the OSIRIS core periphery will be first detailed. Then, a comparison of calculated and measured values of neutron and gamma fluxes will be introduced. (author)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

A complete dosimetry experimental program in support to the core characterization and to the power calibration of the CABRI reactor. A complete dosimetry experimental program in support of the core characterization and of the power calibration of the CABRI reactor

CABRI is an experimental pulse reactor operated by CEA at the Cadarache research center. Since 1978 the experimental programs have aimed at studying the fuel behavior under Reactivity Initiated Accident (RIA) conditions. Since 2003, it has been refurbished in order to be able to provide RIA and LOCA (Loss Of Coolant Accident) experiments in prototypical PWR conditions (155 bar, 300 deg. C). This project is part of a broader scope including an overall facility refurbishment and a safety review. The global modification is conducted by the CEA project team. It is funded by IRSN, which is conducting the CIP experimental program, in the framework of the OECD/NEA project CIP. It is financed in the framework of an international collaboration. During the reactor restart, commissioning tests are realized for all equipment, systems and circuits of the reactor. In particular neutronics and power commissioning tests will be performed respectively in 2015 and 2016. This paper focuses on the design of a complete and original dosimetry program that was built in support to the CABRI core characterization and to the power calibration. Each one of the above experimental goals will be fully described, as well as the target uncertainties and the forecasted experimental techniques and data treatment. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Increasing of MERARG experimental performances: on-line fission gas release measurement by mass spectrometry

The MERARG device - implemented at the LECASTAR Hot Laboratory, at the CEA Cadarache - allows characterizing nuclear fuels with respect to the behaviour of fission gases during thermal transients representative of normal or off normal operating nuclear power plant conditions. The fuel is heated in order to extract a part or the total gas inventory it contains. Fission Gas Release (FGR) is actually recorded by mean of both on-line gamma spectrometry station and micro gas chromatography. These two devices monitor the quantity and kinetics of fission gas release rate. They only address {sup 85}Kr radioactive isotope and the elemental quantification of Kr, Xe and He (with a relatively low detection limit in the latter case, typically 5-10 ppm). In order to better estimate the basic mechanisms that promote fission gas release from irradiated nuclear fuels, the CEA fuel study department decided to improve its experimental facility by modifying MERARG to extend the studies of gamma emitter fission gases to all gases (including Helium) with a complete isotopic distribution capability. To match these specifications, a Residual Gas Analyser (RGA) has been chosen as mass spectrometer. This paper presents a review of the main aspects of the qualification/calibration phase of the RGA type analyser. In particular, results recorded over three mass ranges 1-10 u, 80-90 u and 120-140 u in the two classical modes of MERARG, i.e. on-line and off-line measurements are discussed. Results obtained from a standard gas bottle show that the quantitative analysis at a few ppm levels can be achieved for all isotopes of Kr and Xe, as well as masses 2 and 4 u. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Nuclear fuel behavior at high temperature: new insights from post-test examinations on the VERDON-1 sample

In order to reduce the uncertainties concerning Source Term assessment under Light Water Reactor (LWR) severe accident conditions, 4 VERDON tests were recently performed. They have been defined in the framework of the International Source Term Program and carried out at CEA Cadarache, in the LECA-STAR facility. Among these 4 VERDON tests, VERDON-1 test was devoted to high burn-up UO{sub 2} fuel behavior and FP release under H{sub 2} reducing conditions at very high temperature. This test was performed with the Release Circuit, which is devoted to the precise characterization of the FP release, in terms of FP release kinetics and fission products (FPs) released fraction. The main components of the circuits are a furnace heating the fuel sample, a filter trapping the emitted aerosols and gamma spectrometry detectors for online and post-test measurements. VERDON-1 test was conducted up to about 2600 Celsius degrees without reaching global fuel relocation, allowing thereby post-test micro-analyses. After a brief description of the fuel sample nature and the experimental set-up, we present the progress of the experimental sequence and the associated gamma spectrometry measurements. We then focus on the major results of the VERDON-1 tests regarding the fuel global behavior. These results are obtained by two complementary techniques. First, the fuel sight gamma station can be used either on-line to detect fuel relocation or just after the test, as a longitudinal bench, to get preliminary information on the global shape of the fuel sample inside the furnace. Secondly, these information are completed by post-test examinations performed on the fuel sample after extraction from the furnace. Complete qualitative and quantitative characterizations, including EPMA (Electron Probe Micro-Analysis) and SEM (Scanning Electron Microscopy) analysis, were performed for the first time at the CEA Cadarache on a sample that has been submitted to such severe accident conditions. These results show an advanced state of sample degradation, with locally previously molten areas, meaning that fuel global relocation was imminent by the end of thermal-hydraulic sequence. The relocation temperature is compared to that of previous VERCORS tests and discussed regarding the atmosphere of the test and possible UO{sub 2}-ZrO{sub 2} interaction. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Starting DIADEM medium-level waste interim storage's construction - 15034

As part of its decommissioning policy, the nuclear energy division of CEA has identified materials whose high enough radioactive features authorize neither sustainable storage in existing facilities nor evacuation to a French operational geological disposal. Pending availability of definitive solutions, it was decided to explore the possibility to store these materials in a new installation, the subject of DIADEM project [1]. The need was first identified in Marcoule's site, especially to accommodate future waste dismantling of Phenix reactor. Quickly the interest of a centralized storage for all CEA waste of this type appeared to bring together the best expertise to the design and operation of storage. Waste will be stored either to use benefits of radioactive decay or to wait for the commissioning of a new disposal. DIADEM is designed for an operating estimated lifetime of fifty years. This paper deals with the prior safety assessment completed in 2014, the project management and the first steps of the facility construction. (authors)

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

Robotic D and D: Smart robots - 15093

Project' managers continually seek an ever-greater optimization of time between remote handling operations and those carried out manually; therefore, new technological solutions must be deployed. Robotics offers a great opportunity in this new field of technology to carry out, for example, samplings or remediation in hostile, cluttered surroundings. Teams in charge of dismantling at the CEA have therefore first defined robotizable functions. These functions have been assembled from existing technological blocks to arrive at robots which are operating today [RICAIII, patent: FR 2925702]. Lessons learned, particularly from experience with the RICA robot, have enabled the operating technical specifications to be fine-tuned. A new study phase has been launched applying the same principle of adapting existing, proven means. The growing role of robotics today is unquestioned. Led by research and the academic world; robots such as those equipped with wheels, tracks, feet or even helicopter rotors, are today accessible to the general public, particularly via broadening of the 'open source' concept. Added to these we need tools able to manage large component deconstruction systems, like MAESTRO. Industrialization of such high-potential technological solutions has been aided by: - Easy use, - Increasing reliability, - Flexibility of 'open source' solutions, - Widening skill networks, and therefore greater technical support - Lower costs. Decontamination and dismantling (D and D) projects must be able to meet a number of special demands, increasing the number of unit designs, their costs and delivery times. The complexity of dismantling works sites mean that each is a special case to be dealt with almost independently. Such a way of approaching these projects is not on the same wavelength as industry, with tool and method standardization. The answer to the challenge of operations in difficult environments is an eco-system of functions, performed by a set of inter-connected robots. The first step towards the construction of such robot teams is devoted to functions where strength is not necessary: investigating and clean-up in hostile environments. With this in mind, the CEA Marcoule teams have been given the objective of merging the strengthening commercial robotic world with the needs of D and D, and thus to improve the transversal use of the systems. (authors)

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

Contribution of the numerical simulations in the development of the French HLLW vitrification Processes - 15126

For many years, the CEA (Commissariat a l'energie Atomique et aux energies Alternatives) Marcoule France has developed various processes dedicated to radioactive waste confinement, especially vitrification processes for HLLW. For 10 years now, the numerical simulation has become an important tool for research and development projects held in the CEA-AREVA Joint Vitrification Laboratory (LCV). Induction heating, fluid mechanics and thermal simulations take part of all new R and D projects. The contributions of such simulations are, first, the enhancement of the working knowledge of existing process. Those data are very useful to define optimisation choices, for example upgrades made on the hot metallic melter used since the 90's at La Hague facility. Second, the simulations are, of course, also used at the conception stage of new processes as a tool allowing wide ranges parametric tests. This has been extensively used in the design of the cold crucible inductive melter (CCIM) commissioned in 2010 at La Hague plant. Finally, it is a powerful and relatively cheap tool for prospective studies for processes of the future. Whatever the purpose, the potential benefits are gains on the reliability, the output capacity and the life time. (authors)

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

Defining WAC for interim storage dedicated to tritium decay - 15164

Considering the high mobility of tritium through the package in which it is contained, the new 50-year storage concepts proposed by the CEA currently provide a solution adapted to the management of waste with high tritium concentrations, associated with adapted disposal sites. The 50-year intermediate storage corresponds to 4 tritium radioactive periods i.e. a tritium reduction by a factor 16. The waste acceptance criteria (WAC) for this interim storage concept will take into account the safety of the facility, the reference scheme for the management of tritiated waste in France, and the criteria applicable at Andra disposal centers. This will lead to define a set of waste specifications that describe the generic criteria such as acceptable waste forms, general principles and specific issues, e.g. conditioning, radioactive content, tritium content, waste tracking system, and quality control. This paper details how the work by the CEA has led to a storage facility project that not only takes into account the specificity of tritium, but also the producers' needs and Andra's disposal requirements. (authors)

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

Management of HL deposits for the D and D of UP1 reprocessing plant - retrieval and treatment scenarios - 15295

Built in 1958 on the Marcoule CEA site, the UP1 facility, was the 1. French industrial (500 t U / year) reprocessing plant based on the PUREX process. The plant, which has stopped operation in 1997, constitutes with its auxiliary workshop (vitrification workshop, glass canister storage), one of the largest and most challenging decommissioning project in France. In about 40 years of operation, various type of fuels elements such as metallic uranium, U-Al fuel rods, Phenix breeder elements or PWR fuel have been reprocessed, thereby requiring several plant and process adaptations. Resulting from these complex process chemistry and operation history, significant fouling of the chemical plant circuits and components has occurred. The fission product (FP) concentration process is particularly concerned. Preliminary investigation performed in 2010 in the evaporator's feeding tanks revealed the presence of significant amounts of sludge and dose rates as high as 250 Gy/h. Therefore a thorough investigation campaign, including visual inspection, volume quantification, deposits sampling, and radiological characterization of the components of this workshop has been initiated in 2013. Chemical analysis performed so far show that the solids are mainly composed of zirconium phosphates, 'MoZr' compounds and nitrate salt precipitates. But the presence of insoluble fines arising from the dissolution workshop and organic residues resulting from the TBP decomposition has also been evidenced in some of the tanks. The numerous constraints associated with the management of such highly active sludge, with Cs-137 specific activity ranging from 1 to 10 GBq/g, and a complex physico-chemical composition, lead CEA to investigate the implementation of an in situ conditioning treatment able to produce a waste package complying with existing evacuation routes and/or on-site storage facilities. The full paper presents an overview of the UP1 D and D project and then focuses on the possible dismantling scenarios for the FP concentration workshop, including the evaluation of different strategies for the highly active deposits management. The investigations and deposits characterization program will be developed so as the sampling methodologies that has been specifically tuned. In the light of the best actual knowledge of the deposits properties, the recovery and conditioning treatment scenarios will be exposed, considering, as a case study, the development of a specific vitrification tool to be implemented In Situ and specifically designed to match the requirements and constraints of the D and D project: adequately dimensioned, compact, flexible with respect to the feed composition, for a limited duration of use and designed to be dismantled. (authors)

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

In-Can Melter for Vitrification of Waste Coming from Decommissioning and Dismantling Operations - 19020

Some waste streams from decommissioning and dismantling operations are identified as an issue in terms of handling, transportation, adapted conditioning matrix or evacuation outlet within the existing regulations. In this context, the consortium of the DEM and MELT project which gathers Orano, the CEA and ECM technologies are currently developing and qualifying a specific vitrification tool designed to match the requirements and constraints of D and D projects. The main characteristics of this In-Can vitrification process are presented hereafter: - Suitable for high or medium level waste; - Flexible enough to accommodate uncertainties on waste composition; - Adequately dimensioned and compact to be implemented and operated inside an existing cell or close to the waste to be treated; - Designed for a short duration of use, as a decommissioning tool, aimed to be dismantled just after the treatment operation; - Designed in order to produce a small amount of secondary waste; - Requiring low investment and operation costs. The In-Can melting process has already been developed by the CEA for an internal application (defense waste vitrification) and can be rated TRL 7. That means it has been tested at full scale at a prototypical demonstration level with a range of simulants in an operational environment. This paper presents the In-Can vitrification process and its performances for a typical D and D waste highly loaded with Cs. The vitrification tests results will be in particular discussed in terms of process parameters, glass properties and cesium volatility. This paper also highlights the methodology and the organization (material, process, technological and long term behavior teams) deployed in order to adapt and qualify the In-Can process for all kind of wastes (liquid, solid, highly loaded with Tc and Cs, waste with no route, ...) from the matrix design definition to the full qualification. (authors)

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

In-can Melter for Fukushima Waste: From Laboratory Scale Tests to the Industrial Design Definition - 19021

The In-Can melter process is being developed and qualified to treat waste arising from decommissioning and dismantling operations. This robust, simple and versatile in situ vitrification process is designed for high or medium level waste and is compact enough to be implemented in a decommissioned cell or close to the waste to be treated. It is developed to treat liquid and solid waste, to produce a small amount of secondary waste and to minimize investment and operating costs. The In-Can vitrification process is also developed with a modular design in order to be adapted regarding to nuclear operators' needs and requirements. The close collaboration between the CEA and Orano constitutes a unique structure able to provide a full and integrated solution, from the scenarios studies and matrix design definitions to the industrial implementation. The evaluation of the In-Can process for Fukushima water treatment secondary waste is a project led by Orano, CEA and ANADEC. This paper presents the first results of this project from the scenarios and glass formulation studies to the feasibility evaluation to Fukushima Daiichi Nuclear Power Station implementation. (authors)

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

Sorbmatech{sup R}Cs, an Innovative Highly Selective Cs Sorbent for Wastewater Decontamination - 20065

The treatment of radioactive wastewater from the operating nuclear industry (energy production, reprocessing of spent fuel, decommissioning of shut-down facilities, etc.) or in case of incidental or accidental situation (such as Fukushima Daichii NPP accident) is an important issue requiring constant improvement of the technologies and techniques used to prevent radioactive releases into the environment and/or to reduce the production of secondary waste. CEA, Orano and CTI are collaborating since 2014, with the support of the French government's program 'investments for the future', to develop Sorbmatech{sup R}Cs, an innovative and patented selective sorbent (Grandjean and Barre 2014) to deal with the removal of radioactive {sup 137}Cs, one of the most hazardous isotope to be managed in nuclear waste, highly efficient either in agitated reactor or in column process. Sorbmatech{sup R}Cs sorbent consists of mesoporous silica particles loaded with potassium-copper hexacyanoferrate nanoparticles (KCu-HCF). This peculiar structure enables much faster sorption kinetics compared to denser crystalline structures, such as Silicotitanate or Zeolite sorbents, whose drawback consists in hindering the diffusion process through the material, and whose larger sorbent particles slow down the ion exchange process. The equilibrium of Sorbmatech{sup R}Cs is reached in few minutes, even at high flow rates in a fixed-bed column process (up to 10 m.h{sup -1} Darcy velocity) with low drop pressure. Moreover, Sorbmatech{sup R}Cs can be used over a wide pH range from 2 to 10 and has a very high selectivity towards {sup 137}Cs even in brine solution (saline, mineral or sea water). This new sorbent also has a K{sub d} higher than 10{sup 5} mL.g{sup -1} depending on the salinity of the wastewater, enabling a highly efficient removal of Cs trace in radioactive wastewater compared to Cs selective sorbent such as Zeolites and Silicotitanates. Its radiological resistance enables the treatment of high, medium and low activity wastewater. The CEA-Orano-CTI/ALSYS partnership leads to the development of a simple and low-cost synthetic route, which demonstrated large-scale manufacturing of the product on a batch of 100 kg. Two products have been developed and are ready for short run market introduction so far with grain size well adapted either for column processes or agitated reactors.. Finally, the industrial implementation has been studied and largely benefits from the properties of the sorbent. Indeed, its high sorption kinetics enable high flow rates resulting in a decrease of operation times, unit sizes, and volumes of secondary waste produced. Moreover, Sorbmatech{sup R}Cs is designed to be combined into a full and integrated solution, ranging from preliminary optimization of operating conditions through numerical dimensioning model, industrial implementation and treatment of the wastewater (through cartridge or preconditioned columns systems developed thanks to Orano's extensive expertise), to long-term stabilization of generated waste (under progress). (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Analysis of hafnium configuration of FLUOLE-2 program - 25409

The purpose of this paper is to present an analysis of the hafnium configuration of FLUOLE-2 program. FLUOLE-2 is a set of benchmark-type experiments dedicated to neutron attenuation analysis with the aim of improving the TRIPOLI-4{sup R} Monte Carlo code validation and industrial neutron physics scheme. The CEA developed this program, with the support of EDF, in order to be representative of 900 and 1450 MWe Pressurized Water Reactors operated in France. For that purpose, different stainless steel structures have been designed and appropriately positioned inside the EOLE facility located at Cadarache CEA center. EOLE is a pool type zero power reactor, composed of a cylindrical aluminum vessel with an over-structure of stainless steel, that is able to contain various types of core and related structures. The FLUOLE-2 core has been designed as a 29 x 29 pins square lattice of fuel rods. Different kind of dosimeters (cobalt, gold, tin, rhodium, indium, iron, nickel, titanium, aluminum, and vanadium) were irradiated inside and outside the core. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗