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At least 217 records · Page 12
All sourcesNASA NTRSDOE OSTI
NASA NTRS1994

An operations and command systems for the extreme ultraviolet explorer

About 40% of the budget of a scientific spacecraft mission is usually consumed by Mission Operations & Data Analysis (MO&DA) with MO driving these costs. In the current practice, MO is separated from spacecraft design and comes in focus relatively late in the mission life cycle. As a result, spacecraft may be designed that are very difficult to operate. NASA centers have extensive MO expertise but often lessons learned in one mission are not exploited for other parallel or future missions. A significant reduction of MO costs is essential to ensure a continuing and growing access to space for the scientific community. We are addressing some of these issues with a highly automated payload operations and command system for an existing mission, the Extreme Ultraviolet Explorer (EUVE). EUVE is currently operated jointly by the Goddard Space Flight Center (GSFC), responsible for spacecraft operations, and the Center for Extreme Ultraviolet Astrophysics (CEA) of the University of California, Berkeley, which controls the telescopes and scientific instruments aboard the satellite. The new automated system is being developed by a team including personnel from the NASA Ames Research Center (ARC), the Jet Propulsion Laboratory (JPL) and the Center for EUV Astrophysics (CEA). An important goal of the project is to provide AI-based technology that can be easily operated by nonspecialists in AI. Another important goal is the reusability of the techniques for other missions. Models of the EUVE spacecraft need to be built both for planning/scheduling and for monitoring. In both cases, our modeling tools allow the assembly of a spacecraft model from separate sub-models of the various spacecraft subsystems. These sub-models are reusable; therefore, building mission operations systems for another small satellite mission will require choosing pre-existing modules, reparametrizing them with respect to the actual satellite telemetry information, and reassembling them in a new model. We briefly describe the EUVE mission and indicate why it is particularly suitable for the task. Then we briefly outline our current work in mission planning/scheduling and spacecraft and instrument health monitoring.

Muscettola, Nicola↗
NASA NTRS2004

ThermoBuild: Online Method Made Available for Accessing NASA Glenn Thermodynamic Data

The new Web site program "ThermoBuild" allows users to easily access and use the NASA Glenn Thermodynamic Database of over 2000 solid, liquid, and gaseous species. A convenient periodic table allows users to "build" the molecules of interest and designate the temperature range over which thermodynamic functions are to be displayed. ThermoBuild also allows users to build custom databases for use with NASA's Chemical Equilibrium with Applications (CEA) program or other programs that require the NASA format for thermodynamic properties. The NASA Glenn Research Center has long been a leader in the compilation and dissemination of up-to-date thermodynamic data, primarily for use with the NASA CEA program, but increasingly for use with other computer programs.

McBride, Bonnie↗
NASA NTRS2003

Growth Conditions To Reduce Oxalic Acid Content of Spinach

A controlled-environment agricultural (CEA) technique to increase the nutritive value of spinach has been developed. This technique makes it possible to reduce the concentration of oxalic acid in spinach leaves. It is desirable to reduce the oxalic acid content because oxalic acid acts as an anti-nutritive calcium-binding component. More than 30 years ago, an enzyme (an oxidase) that breaks down oxalic acid into CO2 and H2O2 was discovered and found to be naturally present in spinach leaves. However, nitrate, which can also be present because of the use of common nitratebased fertilizers, inactivates the enzyme. In the CEA technique, one cuts off the supply of nitrate and keeps the spinach plants cool while providing sufficient oxygen. This technique provides the precise environment that enables the enzyme to naturally break down oxalate. The result of application of this technique is that the oxalate content is reduced by 2/3 in one week.

Johnson-Rutzke, Corinne↗
NASA NTRS2016

Thermodynamics of Gas Turbine Cycles with Analytic Derivatives in OpenMDAO

A new equilibrium thermodynamics analysis tool was built based on the CEA method using the OpenMDAO framework. The new tool provides forward and adjoint analytic derivatives for use with gradient based optimization algorithms. The new tool was validated against the original CEA code to ensure an accurate analysis and the analytic derivatives were validated against finite-difference approximations. Performance comparisons between analytic and finite difference methods showed a significant speed advantage for the analytic methods. To further test the new analysis tool, a sample optimization was performed to find the optimal air-fuel equivalence ratio, , maximizing combustion temperature for a range of different pressures. Collectively, the results demonstrate the viability of the new tool to serve as the thermodynamic backbone for future work on a full propulsion modeling tool.

Propulsion↗
NASA NTRS2021

Coastal California Water Resources II: Utilizing NASA Earth Observations to Detect and Assess the Impacts of Estuarine Breach Events for Improved Coastal Wetland Monitoring and Management

Estuaries are dynamic environments that provide a host of vital ecosystem services. California’s Marine Life Protection Act protects such ecosystems by creating Marine Protected Areas. California has approximately 440,000 acres of estuarine habitats as well as 23 Estuarine Marine Protected Areas (EMPAs); thus, in situ data collection is often difficult due to time and resource constraints. This project used remote sensing to gather data that examined the health and dynamics of California EMPAs in order to supplement ground-based field measurements. Through the use of Landsat 8 Operational Land Imager (OLI) and Thermal Infrared Sensor (TIRS), Sentinel-2 Multispectral Instrument (MSI) and Sentinel-1 C-band Synthetic Aperture Radar (C-SAR), this project assessed mouth state, inundation extent, turbidity, temperature, and tidal measurements for observable estuaries. The Normalized Water Difference Index from Sentinel-2 MSI captured estuary mouth state and inundation extent. Landsat 8 OLI and Sentinel-2 MSI detected differences in water quality metrics that correlated to changes in estuary mouth state (i.e., open or closed). The team’s California Estuary Assessment (CEA) tool in Google Earth Engine was successful in analyzing estuary mouth state, inundation, and water quality. It was most effective when breach events were larger than 10 meters in resolution, water surface was smooth, and imagery was unimpeded by algae or sun glint. The CEA tool will allow the partners, the Ocean Protection Council, Central Coast Wetlands Group, Southern California Coastal Water Research Project, and University of California Los Angeles (UCLA) and Davis (UCD), to better understand estuary dynamics and more effectively conduct in situ estuary monitoring.

Sarah Payne↗
NASA NTRSDate not supplied

Utilizing Remote Sensing to Detect and Assess the Impacts of Estuarine Breach Events for Improved Coastal Wetland Monitoring and Management

Estuaries are extremely dynamic environments that provide a host of vital ecosystem services. California’s Marine Life Protection Act protects such ecosystems by creating Marine Protected Areas (MPAs). California has approximately 440,000 acres of estuarine habitat as well as 23 Estuarine Marine Protected Areas (EMPAs). Thus, in situ data collection is often difficult due to time and resource constraints. This project used remote sensing to gather data that examined the health and dynamics of California EMPAs to supplement ground-based field measurements. Through the use of Landsat 8 Operational Land Imager (OLI) and Thermal Infrared Sensor (TIRS), Sentinel-2 MultiSpectral Instrument (MSI), and Sentinel-1 C-band Synthetic Aperture Radar (C-SAR), this project assessed mouth state, inundation extent, turbidity, temperature, and tidal measurements for observable estuaries. The Normalized Difference Water Index (NDWI) from Sentinel-2 MSI captured estuary mouth state and inundation extent. Landsat 8 OLI and Sentinel-2 MSI detected differences in water quality metrics that correlated to changes in estuary mouth state (i.e., open or closed) through algorithms in Google Earth Engine (GEE). The GEE California Estuary Assessment (CEA) tool was created with partner input throughout development and culminated in a graphical user interface tailored to management needs. The CEA tool will allow the partners, the Ocean Protection Council, Moss Landing Marine Laboratories’ Central Coast Wetlands Group, the Southern California Coastal Water Research Project, and University of California Los Angeles (UCLA) and Davis (UCD), to better understand estuary dynamics and facilitate informed management decisions through satellite-based Earth observations.

Alex Gunnerson↗
DOE OSTI2014

IN-CORE FLUX SENSOR EVALUATIONS AT THE ATR CRITICAL FACILITY.

As part of an Idaho State University (ISU)–led Advanced Test Reactor (ATR) National Scientific User Facility (NSUF) collaborative project that includes Idaho National Laboratory (INL) and the French Alternative Energies and Atomic Energy Commission (CEA), flux detector evaluations were completed to compare their accuracy, response time, and longduration performance. Special fixturing, developed by INL, allows real-time flux detectors to be inserted into various Advanced Test Reactor Critical Facility (ATRC) core positions to perform lobe power measurements, axial flux profile measurements, and detector crosscalibrations. Detectors initially evaluated in this program included miniature fission chambers, specialized self-powered neutron detectors (SPNDs), and specially developed commercial SPNDs. Results from this program provide important insights related to flux detector accuracy and resolution for subsequent ATR and CEA experiments and yield new flux data required for benchmarking models in the ATR Life Extension Program (LEP) Modeling Update Project.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2016

Analysis Capability and Data Needs Identified During the Evaluation of the SILENE CAAS Benchmarks

In October 2010, a series of benchmark experiments was conducted at the Commissariat à l’Énergie Atomique et aux Énergies Alternatives (CEA) Valduc SILENE facility. These experiments were a joint effort between the US Department of Energy (DOE) and the French CEA to create three benchmarks for verification and validation of radiation transport codes, as well as evaluated nuclear data, used to analyze criticality accident alarm systems (CAASs). Many preliminary details of these experiments and their evaluation have been discussed in previous publications. The first benchmark evaluation was published in 2015, and the second and third evaluations will be published at the end of 2016.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2019

Results of the Phenix Dissymmetric Test Benchmark Exercise

In the context of the European H2020 project SESAME, a work package was dedicated to the validation of code coupling for the thermal-hydraulics of pool type reactors. To this end, CEA proposed a benchmark exercise of the so-called dissymmetric test performed at the end of life of the PHENIX reactor. The PHENIX reactor was an experimental reactor operated by CEA in France from 1974 until 2009. The knowledge and data acquired over the lifetime of this reactor are of paramount importance for the development of future sodium-cooled generation IV reactors. At the beginning of this test, the reactor is operated at nominal power (350 MWth) with 3 primary pumps and 4 intermediate heat exchangers. During this test, a pump trip occurs at the secondary side of one of the two loops (one loop provides coolant to two intermediate heat exchangers). The chain reaction is then stopped while the pumps remain operating at nominal speed. Dissymmetric flow patterns are then observed in the cold pool as hot sodium leaves the intermediate heat exchangers that are no longer cooled. The failed secondary pump is then restarted while the rotational speed of the other is reduced. Eventually, a stable homogeneous state is reached. This transient is challenging for both modelers and codes since this exercise tackles an industrial scale reactor featuring complex flows and interactions between different scales. The physical phenomena observed include thermal stratification, flow mixing and buoyancy effects. In order to accurately model these phenomena for a reasonable computational costs, the partners used System Thermal Hydraulics codes coupled with Computational Fluid Dynamics codes and one even included a subchannel code. In this paper, the PHENIX reactor and its behavior during the dissymmetric test are described. Then, a comparison of the different modeling options selected by the partners is offered. Finally, the results are compared to the experimental data and discussed and modelling recommendations are proposed.

code coupling↗
DOE OSTI2012

Determination of the kinetic parameters of the CALIBAN metallic core reactor from stochastic neutron measurements

Several experimental devices are operated by the Criticality and Neutron Science Research Dept. of the CEA Valduc Laboratory. One of these is the Caliban metallic core reactor. The purpose of this study is to develop and perform experiments allowing to determinate some of fundamental kinetic parameters of the reactor. The prompt neutron decay constant and particularly its value at criticality can be measured with reactor noise techniques such as Rossi-{alpha} and Feynman variance-to-mean methods. Subcritical, critical, and even supercritical experiments were performed. Fission chambers detectors were put nearby the core and measurements were analyzed with the Rossi-{alpha} technique. A new value of the prompt neutron decay constant at criticality was determined, which allows, using the Nelson number method, new evaluations of the effective delayed neutron fraction and the in core neutron lifetime. As an introduction of this paper, some motivations of this work are given in part 1. In part 2, principles of the noise measurements experiments performed at the CEA Valduc Laboratory are reminded. The Caliban reactor is described in part 3. Stochastic neutron measurements analysis techniques used in this study are then presented in part 4. Results of fission chamber experiments are summarized in part 5. Part 6 is devoted to the current work, improvement of the experimental device using He 3 neutron detectors and first results obtained with it. Finally, conclusions and perspectives are given in part 7. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Experimental power density distribution benchmark in the TRIGA Mark II reactor

In order to improve the power calibration process and to benchmark the existing computational model of the TRIGA Mark II reactor at the Josef Stefan Inst. (JSI), a bilateral project was started as part of the agreement between the French Commissariat a l'energie atomique et aux energies alternatives (CEA) and the Ministry of higher education, science and technology of Slovenia. One of the objectives of the project was to analyze and improve the power calibration process of the JSI TRIGA reactor (procedural improvement and uncertainty reduction) by using absolutely calibrated CEA fission chambers (FCs). This is one of the few available power density distribution benchmarks for testing not only the fission rate distribution but also the absolute values of the fission rates. Our preliminary calculations indicate that the total experimental uncertainty of the measured reaction rate is sufficiently low that the experiments could be considered as benchmark experiments. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

The pre-conceptual design of the nuclear island of ASTRID

The CEA is involved in a substantial effort on the ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) pre-conceptual design in cooperation with EDF, as experienced Sodium-cooled Fast Reactor (SFR) operator, AREVA, as experienced SFR Nuclear Island engineering company and components designer, ALSTOM POWER as energy conversion system designer and COMEX NUCLEAIRE as mechanical systems designer. The CEA is looking for other partnerships, in France and abroad. The ASTRID preliminary design is based on a sodium-cooled pool reactor of 1500 MWth generating about 600 MWe, which is required to guarantee the representativeness of the reactor core and the main components with regard to future commercial reactors. ASTRID lifetime target is 60 years. Two Energy Conversion Systems are studied in parallel until the end of 2012: Rankine steam cycle or Brayton gas based energy conversion cycle. ASTRID design is guided by the following major objectives: improved safety, simplification of structures, improved In Service Inspection and Repair (ISIR), improved manufacturing conditions for cost reduction and increased quality, reduction of risks related to sodium fires and water/sodium reaction, and improved robustness against external hazards. The core is supported by a diagrid, which lay on a strong back to transfer the weight to the main vessel. AREVA is involved in a substantial effort in order to improve the core support structure in particular regarding the ISIR and the connection to primary pump. In the preliminary design, the primary system is formed by the main vessel and the upper closure comprising the reactor roof, two rotating plugs - used for fuel handling - and the components plugs located in the roof penetrations. The Above Core Structure deflects the sodium flow in the hot pool and provides support to core instrumentation and guidance of the control rod drive mechanisms. The number of the major components in the main vessel, primary pumps, Intermediate Heat Exchangers, and Decay Heat Exchangers are now under consideration. Under normal conditions, power release is achieved using the steam/water plant (in case of Rankine steam cycle) or the gas plant (in case of Brayton gas cycle). The diverse design and operating modes of Decay Heat Removal systems provide protection against common cause failures. A Decay Heat Removal system through the reactor vault is in particular studied with the objective to complement Direct Reactor Cooling systems. At this stage of the studies, the secondary system comprises four independent sodium loops (two and three sodium loops configurations are also investigated). Each loop includes one mechanical pump (or a large capacity Annular Linear Induction Electromagnetic Pump), and three modular Steam Generator Units characterized by once through straight tube units with a ferritic tube bundle; nevertheless, helical coil steam generator with tubes made of Alloy 800, and inverted type steam generator with a ferritic tube bundle are also investigated. The limited power of each modular Steam Generator Unit allows the whole secondary loop to withstand a large water/sodium reaction consecutive to the postulated simultaneous rupture of all the heat exchange tubes of one module. The arrangement of the components is based on the 'Regain' concept, in which the secondary pump is situated at a low level in the circuit; conventional arrangement, as SUPERPHENIX type, is a back-up option. Alternative arrangements based on gas cycles are also studied together with Na-gas heat exchanger design. This paper presents a status of the ASTRID pre-conceptual design. The most promising options are highlighted as well as less risky and back-up options. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Uncertainties propagation in the framework of a Rod Ejection Accident modeling based on a multi-physics approach

The control of uncertainties in the field of reactor physics and their propagation in best-estimate modeling are a major issue in safety analysis. In this framework, the CEA develops a methodology to perform multi-physics simulations including uncertainties analysis. The present paper aims to present and apply this methodology for the analysis of an accidental situation such as REA (Rod Ejection Accident). This accident is characterized by a strong interaction between the different areas of the reactor physics (neutronic, fuel thermal and thermal hydraulic). The modeling is performed with CRONOS2 code. The uncertainties analysis has been conducted with the URANIE platform developed by the CEA: For each identified response from the modeling (output) and considering a set of key parameters with their uncertainties (input), a surrogate model in the form of a neural network has been produced. The set of neural networks is then used to carry out a sensitivity analysis which consists on a global variance analysis with the determination of the Sobol indices for all responses. The sensitivity indices are obtained for the input parameters by an approach based on the use of polynomial chaos. The present exercise helped to develop a methodological flow scheme, to consolidate the use of URANIE tool in the framework of parallel calculations. Finally, the use of polynomial chaos allowed computing high order sensitivity indices and thus highlighting and classifying the influence of identified uncertainties on each response of the analysis (single and interaction effects). (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Calculation of the Phenix end-of-life test in natural circulation with the CATHARE code

The Inst. of Radiological Protection and Nuclear Safety (IRSN) acts as technical support to French public authorities. As such, IRSN is in charge of safety assessment of operating and under construction reactors, as well as future projects. In this framework, one current objective of IRSN is to evaluate the ability and accuracy of numerical tools to foresee consequences of accidents. One of the advantages pointed up for fast reactors cooled by heavy liquid metal is the possibility of decay heat removal based on natural convection. The promotion of this passive cooling mode in future safety demonstrations will involve the use of adapted and validated numerical codes. After the final shutdown of the Phenix sodium cooled fast reactor in 2009, a set of tests covering different areas was conducted for code validation, including a natural circulation test in the primary circuit. Experimental data were issued by CEA to organize a benchmark exercise in the frame of an IAEA Coordinated Research Project (CRP), with the objective to assess the system-codes capability in simulating the thermal-hydraulics behavior of sodium cooled fast reactors in such accidental conditions. IRSN participated to this benchmark with the CATHARE code. This code, co-developed by CEA, EDF, AREVA and IRSN and widely used for PWR safety studies, was recently extended for sodium applications. This paper presents the CATHARE modeling of the Phenix primary circuit and the results obtained. A relatively good agreement was found with available measurements considering the experimental uncertainties. This work stressed the local aspects of phenomena occurring during the natural convection establishment and the limits of a 0D/1D approach. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Analysis of the KROTOS KFC test by coupling X-Ray image analysis and MC3D calculations

During a hypothetical severe accident sequence in a Pressurized Water Reactor (PWR), the hot molten materials (corium) issuing from the degraded reactor core may generate a steam explosion if they come in contact with water and may damage the structures and threaten the reactor integrity. The SERENA program is an international OECD project that aims at helping the understanding of this phenomenon also called Fuel Coolant Interaction (FCI) by providing data. CEA takes part in this program by performing tests in its KROTOS facility where steam explosions using prototypic corium can be triggered. Data about the different phases in the premixing are extracted from the KROTOS X-Ray radioscopy images by using KIWI software (KROTOS Image analysis of Water-corium Interaction) currently developed by CEA. The MC3D code, developed by IRSN, is a thermal-hydraulic multiphase code mainly dedicated to FCI studies. It is composed of two applications: premixing and explosion. An overall FCI calculation with MC3D requires a premixing calculation followed by an explosion calculation. The present paper proposes an alternative approach in which all the features of the premixing are extracted from the X-Ray pictures using the KIWI software and transferred to an MC3D dataset for a direct simulation of the explosion. The main hypothesis are discussed as well as the first explosion results obtained with MC3D for the KROTOS KFC test. These results are rather encouraging and are analyzed on the basis of comparisons with the experimental data. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

The technical and economic impact of minor actinide transmutation in a sodium fast reactor

Within the frame work of the French National Act of June 28, 2006 pertaining to the management of high activity, long-lived radioactive waste, one of the proposed processes consists in transmuting the Minor Actinides (MA) in the radial blankets of a Sodium Fast Reactor (SFR). With this option, we may assess the additional cost of the reactor by comparing two SFR designs, one with no Minor Actinides, and the other involving their transmutation. To perform this exercise, we define a reference design called SFRref, of 1500 MWe that is considered to be representative of the Reactor System. The SFRref mainly features a pool architecture with three pumps, six loops with one steam generator per loop. The reference core is the V2B core that was defined by the CEA a few years ago for the Reactor System. This architecture is designed to meet current safety requirements. In the case of transmutation, for this exercise we consider that the fertile blanket is replaced by two rows of assemblies having either 20% of Minor Actinides or 20% of Americium. The assessment work is performed in two phases. - The first consists in identifying and quantifying the technical differences between the two designs: the reference design without Minor Actinides and the design with Minor Actinides. The main differences are located in the reactor vessel, in the fuel handling system and in the intermediate storage area for spent fuel. An assessment of the availability is also performed so that the impact of the transmutation can be known. - The second consists in making an economic appraisal of the two designs. This work is performed using the CEA's SEMER code. The economic results are shown in relative values. For a transmutation of 20% of MA in the assemblies (S/As) and a hypothesis of 4 kW allowable for the washing device, there is a large external storage demanding a very long cooling time of the S/As. In this case, the economic impact may reach 5% on the capital part of the Levelized Unit Electricity Cost (LUEC). A diminished concentration at 10% of MA, reduces the size of the external storage and the cooling time of the assemblies becomes compatible with the management of the irradiated fuel. Even with a low allowable power for the washing device, the economic impact on the capital cost is less than 2.5%. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Behavior under LOCA conditions on enhanced accident tolerant chromium coated zircaloy-4 claddings

For enhanced accident tolerant fuels for light water reactors application, chromium coatings on zirconium based nuclear fuel claddings are developed and studied at CEA in the framework of the French CEA-EDF-AREVA collaborative program. The results obtained so far, mainly on Zircaloy-4 substrate, show very good corrosion resistance in nominal conditions and significant enhancement of the resistance of the material to oxidation in steam at high temperature (HT), up to 1300 Celsius degrees, with a drastic decrease of hydrogen release and/or pick-up. The present paper reports some new results obtained on chromium coated Zircaloy-4 claddings tested in loss-of-coolant accident (LOCA) conditions. In order to investigate the potential effect of the coating on the cladding mechanical behavior at HT and the capacity of the coating to sustain significant substrate deformation (i.e., during ballooning until burst occurrence) without generalized cracking/peeling, a preliminary limited set of internal pressure creep and temperature ramp tests have been performed in steam environment thanks to the EDGAR facility. The thermal-mechanical tests were done for testing/burst temperatures ranging from 600 C. degrees (α{sub Zr} phase domain) up to 1000 C. degrees (β{sub Zr} phase domain) on 50 cm long low-tin Zircaloy-4 cladding samples with a 15 μm thick outer chromium coating. It is shown that: -) whatever the applied temperature/pressure values, the chromium coating is still fully adherent after having experienced ballooning and burst, including at the vicinity of the burst opening where the Zircaloy-4 clad substrate is highly deformed; -) a HT strengthening effect of the coating on the overall creep clad behavior is evidenced when compared to uncoated Zircaloy-4 cladding materials tested in the same conditions; -) as a consequence, it is observed that, in the 600-750 C. degrees temperature range (α{sub Zr} phase domain) and after burst occurrence, the balloon sizes (i.e., 'uniform' and maximum hoop strains) are generally reduced when compared to uncoated materials; -) regarding the burst mechanism in the β{sub Zr} phase temperature range (1000 C. degrees), it is interesting to observe that, even if some ballooning occurred prior to the cladding failure, the actual burst openings are generally very small (in the order of 1 mm{sub 2} or less), reducing the risk of fuel fragments dispersal in the coolant.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Nuclear Liquid Wastes Calcination: The High-Level French Experience - 17184

Calcination is an efficient process for volume reduction and stabilization of nuclear liquid wastes. It is often the first step of waste containment processes such as vitrification or grouting. The CEA (French Alternative Energies and Atomic Energy Commission) and AREVA have acquired a high-level experience in the field of calcination through more than forty years of R and D and industrial operation. During this period, a broad range of liquid wastes have been studied and treated. R and D work and associated engineering studies allowed defining the design and sizing of the equipment and the process control parameters to be applied for the calcination of all kinds of solutions. The calcination process has been qualified for the following effluents amongst others: - Liquid waste from UOx fuel reprocessing; - Liquid waste from UMo fuel reprocessing (high molybdenum and phosphorus contents); - Hanford HLW; - Liquid waste from D and D operation (sodium bearing waste). This paper presents several studies led by the Joint Vitrification Laboratory (L.C.V), a common research laboratory between CEA and AREVA in charge of qualifying new processes and matrices for waste containment. These studies allowed defining the calcination rules, in particular for very hard to process solutions such as effluents with high boron, molybdenum and sodium contents. The feedback from industrial operation of the calcination process with fission products and effluents from D and D operations, highly loaded with sodium, is also presented. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗