Advanced Sodium Technological Reactor For Industrial Demonstration Research Articles

Comparison of sodium fast reactor core assembly seismic evaluation using the Japanese and French simulation tools

Launched in 2014, Japan-France collaboration on the ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) project is active both in reactor design and in R&D fields. CEA, Framatome, JAEA, MHI and MFBR are in charge of this collaboration. In the design field, Japan-France evaluates core assemblies with interferences on seismic events. The objectives of this collaboration are to determine the core assembly lattice displacements under seismic load in order to demonstrate the safe behavior of the core regarding control rod insertion, structural integrity, and reactivity variation, etc.From this point of view, the object of this study is to verify the seismic evaluation method on core assemblies between Japan and France by comparing the results. Using their evaluation method applied to common the ASTRID core input data, Japan and France will calculate a collision load, a horizontal displacement, and a vertical displacement for core assemblies.As the first step of this work, the comparison of each evaluation (modelling and calculation results) is carried out considering the horizontal seismic excitation and without including the vertical component.The analysis of this benchmark calculation shows a satisfactory agreement between the Japanese and French tools and the figures show a good behavior of the core in horizontal direction under French seismic condition.This paper describes the summary of horizontal seismic analysis comparison and the future analysis plan.

ASTRID, Back to the Future: Bridging Scales in the Development of Nuclear Infrastructures

Very few papers exist in the field of social sciences that follow and study nuclear infrastructure design projects from the inside. Such a perspective would make it possible to understand the mechanisms of their successes or difficulties at their very origin. At a time when high hopes are placed on civil nuclear energy to solve the climate issue, but when simultaneously, nuclear industry actors are facing major difficulties in a great number of flagship projects, this situated understanding of any given project’s trajectory becomes crucially important. This paper proposes an analysis, from the inside, of a project that, about ten years ago, raised great expectations from both the French and global nuclear industry, but which, in 2019, was finally halted. This project is that of a so-called fourth generation sodium reactor: the Advanced Sodium Technological Reactor for Industrial Demonstration (ASTRID). ASTRID was a new sodium-cooled fast reactor. Begun in 2010, the project’s development was suspended at the end of 2019 by the French authorities. Through an inductive study of the project in the making, conducted from 2015 to 2019, the research team was able to follow the project trajectory and carry out nearly 30 interviews with actors directly involved in the project itself. By studying ASTRID as an infrastructure development project and building on the concepts of scale and infrastructure from the literature, the ASTRID project’s halt can be understood. The project’s suspension was the result of the increasing complexity and ambiguity faced by project members and stakeholders in aligning the local and global scales as the new infrastructure was developed, and more precisely, the ASTRID project infrastructure and the global nuclear infrastructure. Our analyses show that ASTRID’s trajectory gradually drifted as a result of three misalignments between the project’s infrastructure and the global nuclear infrastructure: a temporal, social, and physical misalignment. As a result, the project identity can be described as having been vague and ambiguous. This paper sets out how such a lack of clarity impacted design practices, the project organization as a whole, and the ASTRID trajectory. In consequence, it is crucially important that lessons are learned from the project’s cessation to understand both the difficulties related to the nuclear renaissance and in terms of the field of project management in general.

Design-oriented tool for Unprotected Loss Of Flow simulations in a Sodium Fast Reactor: Validation and application to stability analyses

Within the framework of the Generation IV Sodium-cooled Fast Reactor (SFR) R&D program of CEA (French Commissariat à l’Energie Atomique et aux Energies Alternatives), the reactor behavior in case of severe accidents is assessed through experiments and simulations. Such accidents are usually simulated with mechanistic calculation tools (such as SAS-SFR and SIMMER-III). As a complement to these codes, which give reference accidental transient results, a new Best Estimated Plus Uncertainty approach has been developed by CEA; its final objective being to derive the variability of the main outputs of interest for the safety. This approach involves a fast-running description of extended accident sequences coupling physical models for the main phenomena with advanced statistical analysis techniques. It enables to perform a large number of simulations in a reasonable computational time and to describe all the possible bifurcations of the accident transient. In this context, this paper presents a physical tool (models and results assessment) dedicated to the primary phase of an Unprotected Loss Of Flow accident (ULOF) in a SFR, i.e. before the first hexagonal can failure. This paper focuses on the modeling of neutronic phenomena and of the pin degradation and on their validations, the two-phase sodium thermal–hydraulic behavior having largely been assessed in a previous paper. The physical tool called MACARENa is described before presenting the comparison of its results with experimental tests results for the pin degradation behavior modelling and with mechanistic SIMMER-III code results for the global core behavior evolution during an ULOF transient. This tool is demonstrated to be capable of reproducing the mass-flow rate, reactivity, power evolution, and global pin degradation behaviour during an ULOF with a low discrepancy regarding the same transient simulated with SIMMER-III. In particular, observed discrepancies concerning mass flow rates and power evolutions are under 3% before boiling occurs and, later during the transient, the power excursion starts only 2s earlier in the MACARENa simulation than in the SIMMER-III one. The fast-running tool has then been used for safety-informed design and flow stability analyses of fast reactor systems, more specifically dealing with the ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) core concept. It allowed to emphasize the main dominant phenomena and the significant trends for safety assessment. This work has pointed out the strong impact of the modelling of reactivity feedback due to thermal expansion on the resulting core degradation during an ULOF. It has also highlighted the main scenario parameters and models influencing this transient evolution.

The knowledge management on the design of a generation IV sodium fast reactor project at CEA. The case and methodology applied on the Astrid project

From 2010 to 2019, the French Alternative Energies and Atomic Commission (CEA) associated with industrial partners realized the Basic Design of a prototype Sodium Fast Reactor. This project was called ASTRID (ASTRID for Advanced Sodium Technological Reactor for Industrial Demonstration). ASTRID design studies were financed through governmental funds until the end of the basic design. These funds covered also the design studies for the core manufacturing workshop, the refurbishment or construction of large test loops. One year before the term of this Basic Design phase (in 2018), industrial partners, CEA and the French State conducted a review of fast neutrons reactors and fuel cycle strategy. The review which is now translated into the Multiannual Energy Program concluded that the perspective of industrial deployment of Fast Reactors is more distant. Yet it has been concluded to keep this option open, requiring to maintain competences, and to progress on technological barriers and further develop know-how. The strategy for complete closure of nuclear fuel cycle is maintained as a long-term sustainability objective (in the second half of the 21st century). Therefore, as a direct consequence of this decision, the ASTRID project stopped at the end of 2019 at its Basic Design phase. Quickly the question raised on the Knowledge Management (KM) and Know-How capitalization of the huge amount of studies and results realized during ten years (around 23 000 technical documents). Moreover the challenge was to realize this KM process in less than one year, before the ASTRID project team definitive split. The paper is presenting an innovative KM methodology which has been created and specifically performed on the ASTRID project. It is based on a series of interviews and video recordings, all transformed into some New KM tools called “MOOK” (MOOK for Management of Organized Online Knowledge). All these MOOKs considered as “data rich contents” are then inter-connected and linked by the ASTRID Product Breakdown Structure to some fundamental documents, for a comprehensive and quick mapping of the project. They finally form an efficient KM tool recorded in a PLM Software (PLM for Product Lifecycle Management). Thus the ASTRID project team has realized a high level and easy-to-use “GPS” (Global Positioning System) tool to keep the ASTRID history, context, knowledge and know-how for years. This KM methodology can be easily adapted to other nuclear projects and needs.

CFD analysis of the flow in the MICAS experimental facility, a water model of the hot pool of a sodium cooled fast reactor

Sodium Cooled Fast Reactors (SFRs) have been developed in France for about 50 years including the research reactors Rapsodie and Phenix as well as the Superphenix plant. In the framework of Generation IV reactor deployment, the development the Advanced Sodium Technological Reactor for Industrial Demonstration (ASTRID) was started in 2006 under the leadership of CEA. Current activities on the use of Computational Fluid Dynamics (CFD) in the ASTRID project are focused on the validation of computational tools to predict the thermal–hydraulic behavior of sodium in the reactor vessel. As thermal hydraulics is recognized as key scientific subject in the development of SFRs, new experiments are realized for both qualification of design options and validation of code calculations. Large scale tests are conducted in water models such as the MICAS facility, which represents the hot pool of the ASTRID reactor in 1/6 scale including all internals. The facility is built in transparent polymer for accurate optical measurements of the flow field under various flow conditions.The flow in the MICAS facility is analyzed with the CEA in-house code TrioCFD, the reference CFD code of the Nuclear Reactor Division of the CEA. The use of HPC (High Performance Computing) allows the access to flow fields in complex geometrical structures with a high resolution in time and space. The validation of the modelling approach is shown on the example of the MICAS experiment by comparing calculation results (velocity fields) to LDV measurements. Detailed information is given on the test facility, the used modelling approach and the comparison between experiment and calculation. Although the calculation is globally in accordance to the experiment, differences exist locally due to the complexity of the flow topology in the hot pool.

Modeling representative Gen-IV molten fuel reactivity effects in the ZEPHYR ZPR - LFR analysis

© 2018 John Wiley & Sons, Ltd. Studies related to severe core accidents constitute a crucial element in the safety design of Gen-IV systems. A new experimental program, related to severe core accidents studies, is proposed for the zero-power experimental physics reactor (ZEPHYR) future reactor. The innovative program aims at studying reactivity effects at high temperature during degradation of Gen-IV cores by using critical facilities and surrogate models. The current study introduces the European lead-cooled system (ELSY) as an additional Gen-IV system into the representativity arsenal of the ZEPHYR, in addition to the sodium-cooled fast reactors. Furthermore, this study constitutes yet another step towards the ultimate goal of studying severe core accidents on a full core scale. The representation of the various systems is enabled by optimizing the content of plutonium oxide in the ZEPHYR fuel assembly. The study focuses on representing reactivity variation from 900°C at nominal state to 3000°C at a degraded state in both ELSY and Advanced Sodium Technological Reactor for Industrial Demonstration (ASTRID) cores. The study utilizes the previously developed calculation scheme, which is based on the coupling of stochastic optimization process and Serpent 2 code for sensitivity analysis. Two covariance data are used: the ENDF 175 groups for ELSY and the Covariance Matrix Cadarache (COMAC) 33 groups for ASTRID. The effect of the energy group structure of the covariance data on the representativity process is found to be significant. The results for single degraded ELSY fuel assembly demonstrate high representativity factor (>0.95) for reactivity variation and for the criticality level. Also, it is shown that the finer energy group structure of the covariance matrices results in dramatic improvement in the representation level of reactivity variations.

Conceptual design of ASTRID radial shielding sub-assemblies

The French 600 MWe Advanced Sodium Technological Reactor for Industrial Demonstration (ASTRID) project reached in 2015 the end of its conceptual design phase. The core design studies are being conducted by the CEA with support from AREVA NP and EDF. Innovative design choices for the core have been made to comply with the GEN IV reactor objectives, marking a breakthrough with the former Phénix and SuperPhénix Sodium Fast Reactors.One of the biggest challenges of the last five years was to propose a consistent design for the reflectors and neutron shielding sub-assemblies surrounding the fuel core in order to fulfill ASTRID requirements of minimising the secondary sodium activity level. Heavy iterative studies on both core and sub-assemblies were necessary to propose and evaluate different solutions following a strict value analysis process considering neutron shielding performances, life duration, maturity levels, washing and manufacturing capability, and qualification needs. Evaluated options were reflectors sub-assemblies made of steel or MgO rods, and radial neutron shielding sub-assemblies made of B4C or borated steel, with different configurations in the design and in the core layout.This paper presents the iterative engineering studies, conducted by CEA and performed by AREVA NP, concerning the radial shielding sub-assemblies for ASTRID core, from the selection of possible solutions to a final consistent conceptual design.

APOLLO3 homogenization techniques for transport core calculations—application to the ASTRID CFV core

Abstract This paper presents a comparison of homogenization techniques implemented in the APOLLO3 platform for transport core calculations: standard scalar flux weighting and new flux–moment homogenization, in different combinations with (or without) leakage models. Besides the historical B1 - homogeneous model, a new B-heterogeneous one has indeed been implemented recently in the two/three-dimensional-transport solver using the method of characteristics. First analyses have been performed on a very simple Sodium Fast Reactor core with a regular hexagonal lattice. They show that using the heterogeneous leakage model in association with flux-moment homogenization strongly improves the prediction of k eff and void reactivity effects. These good results are confirmed when the application is done to the fissile assemblies of the more complex CFV (Low Void Effect) core of the ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) project of sodium-cooled fast breeder reactor (Generation IV).

SIMMER extension for multigroup energy structure search using genetic algorithm with different fitness functions

Abstract The multigroup transport theory is the basis for many neutronics modules. A significant point of the cross-section (XS) generation procedure is the choice of the energy groups' boundaries in the XS libraries, which must be carefully selected as an unsuitable energy meshing can easily lead to inaccurate results. This decision can require considerable effort and is particularly difficult for the common user, especially if not well-versed in reactor physics. This work investigates a genetic algorithm-based tool which selects an appropriate XS energy structure (ES) specific for the considered problem, to be used for the condensation of a fine multigroup library. The procedure is accelerated by results storage and fitness calculation speed-up and can be easily parallelized. The extension is applied to the coupled code SIMMER and tested on the European Sustainable Nuclear Industrial Initiative (ESNII+) Advanced Sodium Technological Reactor for Industrial Demonstration (ASTRID)-like reactor system with different fitness functions. The results show that, when the libraries are condensed based on the ESs suggested by the algorithm, the code actually returns the correct multiplication factor, in both reference and voided conditions. The computational effort reduction obtained by using the condensed library rather than the fine one is assessed and is much higher than the time required for the ES search.

Conceptual design of ASTRID fuel sub-assemblies

The French 600MWe Advanced Sodium Technological Reactor for Industrial Demonstration (ASTRID) project has reached the end of its Conceptual Design phase. The core design studies are being conducted by the CEA with support from AREVA and EDF. Innovative design choices for the core have been made to comply with the GEN IV reactor objectives, marking a break with the former Phénix and SuperPhénix Sodium Fast Reactors.The main objective to improve safety compared with current GEN II or III reactors led to a core design that demonstrates intrinsically safe behaviour. A negative sodium void worth is achieved thanks to a new fuel sub-assembly design including (U,Pu)O2 and UO2 axially heterogeneous fuel pins, a large cladding/small spacer wire bundle, a sodium plenum above the fuel pins, and upper neutron shielding with both enriched and natural boron carbide (B4C) which also maintain a low secondary sodium activity level. As these Na-bonded B4C pins can lead to the retention of unacceptable amounts of sodium, the whole upper neutron shielding has been made removable on-line through the sub-assembly head just before the washing operations. Finite elements calculations have been performed to increase the stiffness of the stamped spacer pads in order to analyse its effect on the core mechanical behaviour during hypothetical radial core flowering and compaction events. More generally, all design choices for ASTRID have been made with the permanent objective of minimising the sub-assembly height to decrease the overall costs of the reactor and the fuel cycle.This paper describes the fuel sub-assembly design for the ASTRID CFV v4 core at the end of the Conceptual Design phase and studies performed to optimise the performances. Focus is placed on innovations and specificities in the design compared with former French SFRs.

Control Rod Calculation in Axially Heterogeneous Fast Reactors—I: Influence of the Absorber Environment

In axially heterogeneous fast reactor concepts, such as the Advanced Sodium Technological Reactor for Industrial Demonstration (ASTRID) CFV (French acronym of Cœur à Faible effet de Vide sodium, meaning low sodium void effect core) core, the accurate neutronic prediction of control rods is a challenge. In such cores, the performance of the classical two-dimensional (2-D) equivalence procedure, used for control rod homogenization in homogeneous fast reactors, is questionable.In this work (part I of two companion papers), a number of axially heterogeneous environments, representative of a CFV-type core, are investigated using 2-D (X-Z) models, with the objective to distinguish regions where the classical equivalence procedure is valid from those where it is not.It is found that the environments that affect the control rod absorber the most, and are likely to invalidate the procedure, are the internal control rod interfaces, such as the absorber/follower interface and the interface between zones of different boron enrichments. The range of the main spectral impact could be seen within 0 to 10 cm from the material interfaces studied.In the companion paper (part II), a full-core investigation is performed that builds upon the results of this paper.

Control Rod Calculation in Axially Heterogeneous Fast Reactors—I: Influence of the Absorber Environment

In axially heterogeneous fast reactor concepts, such as the Advanced Sodium Technological Reactor for Industrial Demonstration (ASTRID) CFV (French acronym of Cœur à Faible effet de Vide sodium, meaning low sodium void effect core) core, the accurate neutronic prediction of control rods is a challenge. In such cores, the performance of the classical two-dimensional (2-D) equivalence procedure, used for control rod homogenization in homogeneous fast reactors, is questionable.In this work (part I of two companion papers), a number of axially heterogeneous environments, representative of a CFV-type core, are investigated using 2-D (X-Z) models, with the objective to distinguish regions where the classical equivalence procedure is valid from those where it is not.It is found that the environments that affect the control rod absorber the most, and are likely to invalidate the procedure, are the internal control rod interfaces, such as the absorber/follower interface and the interface between zones of different boron enrichments. The range of the main spectral impact could be seen within 0 to 10 cm from the material interfaces studied.In the companion paper (part II), a full-core investigation is performed that builds upon the results of this paper.

Control Rod Calculation in Axially Heterogeneous Fast Reactors—II: Impact of Three-Dimensional Homogenization on Core Parameters

Advanced sodium-cooled fast reactors with improved safety features such as the French Advanced Sodium Technological Reactor for Industrial Demonstration (ASTRID) CFV (French acronym of Cœur à Faible effet de Vide sodium, meaning low sodium void effect core) core concept are characterized by an axial heterogeneous core that will present a challenge for the homogenization procedures used today, taking into account all the different axial material transitions. Reliable modeling of the control rod and accurate prediction of the control rod worth are essential to determining the shutdown margins and to ensuring safe operation.In this work (part II of two companion papers), two different homogenization schemes are compared. One is based on the traditional reactivity-equivalence procedure in two dimensions, and the other is a newly implemented three-dimensional (3-D) version of the reactivity-equivalence procedure, with approximations based on the results in the companion paper. The deterministic results are compared with a Monte Carlo reference.Both cross-section sets from the two homogenization schemes yielded results within the requested ±5% error margin in reactivity. The largest discrepancy was found for the classical procedure for the case with a slightly inserted control rod (normal operating conditions). Both cross-section sets yielded similar power profiles in the fuel subassembly neighboring the control rod within the 2σ Monte Carlo standard deviation. Neither of the cross-section sets was able to predict the large gradients in capture rates close to the internal control rod interfaces.The study showed that the traditional two-dimensional (2-D) reactivity-equivalence procedure produces homogenized cross sections that yield reliable results in a CFV-type core. One exception from this was found for slightly inserted control rods, where the effect of the follower-absorber interface could not be fully captured by the 2-D scheme, and for such cases, 3-D modeling is recommended.

Control Rod Calculation in Axially Heterogeneous Fast Reactors—II: Impact of Three-Dimensional Homogenization on Core Parameters

Advanced sodium-cooled fast reactors with improved safety features such as the French Advanced Sodium Technological Reactor for Industrial Demonstration (ASTRID) CFV (French acronym of Cœur à Faible effet de Vide sodium, meaning low sodium void effect core) core concept are characterized by an axial heterogeneous core that will present a challenge for the homogenization procedures used today, taking into account all the different axial material transitions. Reliable modeling of the control rod and accurate prediction of the control rod worth are essential to determining the shutdown margins and to ensuring safe operation.In this work (part II of two companion papers), two different homogenization schemes are compared. One is based on the traditional reactivity-equivalence procedure in two dimensions, and the other is a newly implemented three-dimensional (3-D) version of the reactivity-equivalence procedure, with approximations based on the results in the companion paper. The deterministic results are compared with a Monte Carlo reference.Both cross-section sets from the two homogenization schemes yielded results within the requested ±5% error margin in reactivity. The largest discrepancy was found for the classical procedure for the case with a slightly inserted control rod (normal operating conditions). Both cross-section sets yielded similar power profiles in the fuel subassembly neighboring the control rod within the 2σ Monte Carlo standard deviation. Neither of the cross-section sets was able to predict the large gradients in capture rates close to the internal control rod interfaces.The study showed that the traditional two-dimensional (2-D) reactivity-equivalence procedure produces homogenized cross sections that yield reliable results in a CFV-type core. One exception from this was found for slightly inserted control rods, where the effect of the follower-absorber interface could not be fully captured by the 2-D scheme, and for such cases, 3-D modeling is recommended.

Nuclear data sensitivity and uncertainty assessment of sodium voiding reactivity coefficients of an ASTRID-like sodium fast reactor

The EU 7th Framework ESNII+ project was launched in 2013 with the strategic orientation of preparing ESNII for Horizon 2020. ESNII stands for the European Industrial Initiative on Nuclear Energy, created by the European Commission in 2010 to promote the development of a new generation of nuclear systems in order to provide a sustainable solution to cope with Europe’s growing energy needs while meeting the greenhouse gas emissions reduction target. The designs selected by the ESNII+ project are technological demonstrators of Generation-IV systems. The prototype for the sodium cooled fast reactor technology is ASTRID (standing for Advanced Sodium Technological Reactor for Industrial Demonstration), which detailed design phase is foreseen to be initiated in 2019. The ASTRID core has a peculiar design which was created in order to tackle the main neutronic challenge of sodium cooled fast reactors: the inherent overall positive reactivity feedback in case of sodium voiding occurring in the core. Indeed, the core is claimed by its designers to have an overall negative reactivity feedback in this scenario. This feature was demonstrated for an ASTRID-like core within the ESNII+ framework studies performed by nine European institutions. In order to shift the paradigm towards best-estimate plus uncertainties, the nuclear data sensitivity analysis and uncertainty propagation on reactivity coefficients has to be carried out. The goal of this work is to assess the impact of nuclear data uncertainties on sodium voiding reactivity feedback coefficients in order to get a more complete picture of the actual safety margins of the ASTRID low void-core design. The nuclear data sensitivity analysis is performed in parallel using SCALE TSUNAMI-3D and the newly developed GPT SERPENT 2 module. A comparison is carried out between the two methodologies. Uncertainty on the sodium reactivity feedbacks is then calculated using TSAR module of SCALE and the necessary safety margins conclusions are drawn.

Transient behavior of ASTRID with a gas power conversion system

The present article is dedicated to preliminary transient studies carried out for the analysis of the system overall behavior of the ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) demonstrator developed in France by CEA and its industrial partners. ASTRID is foreseen to demonstrate the progress made in SFR technology at an industrial scale by qualifying innovative options, some of which still remain open in the areas requiring improvements, especially safety and operability. Among the innovative options, a gas power conversion systems (PCS) is envisaged. In this innovative PCS, the working gas is nitrogen whose flow rate delivers power to a turbine driving with the same shaft two compressors (low and high pressure) separated by an intercooler. The other part of the work delivered by the gas is used to drive the alternator that produces electricity. The main objective of such a PCS consists in avoiding physically the possibility of a sodium/water reaction with the secondary circuit but the impact of this PCS on the control of incidental and accidental transients has also been studied. The main purpose of the studies presented in the paper is to assess the dynamic behavior of ASTRID including a gas PCS with the CATHARE2 code. The first transient presented deals with a loss of off-site power and has been calculated for the gas PCS but also for a classical steam/water PCS for comparison purpose. Then typical transients of gas system have been investigated. Several families of initiating events affecting the PCS are studied: breaches in the ternary circuit, a loss of power supply, a loss of main heat sink and the spurious opening of the turbine by-pass valve. Regarding this last transient, sensitivity studies have shown that an adequate design of the turbomachine by-pass lines allows a control of the event without shutdown. Moreover, regulating actions following the scram are also investigated in order to take benefit of the PCS in order to remove the decay heat.

A comparison between oxide and metallic fueled ASTRID-like reactors

In order to limit the greenhouse emissions and ensure energy security, the global climate change perspective implies a modification in the future energy sector supply. Nuclear energy is an alternative that would decrease the amount of greenhouse gas emissions, and specifically, the fourth generation reactors that present a favorable outlook for clean energy supply. In particular, the Advanced Sodium Technological Reactor for Industrial Demonstration (ASTRID) is a fourth generation sodium-cooled pool type fast nuclear reactor of 1500MWth that seems to be a promising design into the commercial fast nuclear reactors. One of the different core designs for ASTRID is a low void effect core (CFV: Cœur à Faible effet de Vidange sodium) due to the reactor configuration and the radial and axial position of the fuel subassemblies. The current paper presents a heterogeneous ASTRID model developed in MCNP6 with JEFF-3.2 cross sections library, in which three neutronic parameters (Keff at End of Cycle, Doppler constants and coolant void worth) are validated with the European Benchmark on the ASTRID-like low-void-effect core characterization, presented in the ICAPP conference in Nice, France, on May 2015. Validation of the ASTRID model was successfully accomplished, having slightly differences with the benchmark participants in the neutronic parameters. Once the validation was done, a metallic fueled ASTRID-like design is proposed and compared with the oxide fueled reactor. Criticality analysis was done in order to understand the reactor behavior during one cycle of operation as well as Doppler constants and coolant void worth comparison. Actinide inventory was analyzed for both models and 239Pu was tracked with the purpose of observing the breeding achieved in the fertile blankets. Reactivity at the End of Cycle had only a 3pcm delta Keff difference in comparison with the oxide design, Doppler constants behaved as expected and coolant void worth for the whole core has a negative reactivity effect, which is the essential characteristic of the CFV core. Metallic model performed 8.57% more breeding in isotope 239Pu than the conventional oxide fueled design in the fertile zone, and in the global actinide inventory achieved 3% more 239Pu. Metallic design enhances the plutonium management because, besides that the reactor can be sustained in a supercritical state with less initial Pu enrichment, breeding has a better performance.

An innovative plate heat exchanger of enhanced compactness

In the framework of CEA R&D program to develop the Advanced Sodium Technological Reactor for Industrial Demonstration (ASTRID), the present work aims to demonstrate the industrial interest of an innovative compact heat exchanger technology. In fact, one of the main innovations of the ASTRID reactor could be the use of a Brayton Gas-power conversion system, in order to avoid the energetic sodium–water interaction that might occur if a traditional Rankine cycle was used. The present work aims to study the thermal-hydraulic performance of the innovative compact heat exchanger concept. Hence, thanks to a trustful numerical model, friction factor and heat transfer correlations are obtained. Then, a global compactness comparison strategy is proposed, taking into account design constraints. Finally, it is demonstrated that the innovative heat exchanger concept is more compact then other already industrial technologies of interest, showing that is can be considered to warrant serious consideration for future ASTRID design as well as for any industrial application that needs very compact heat exchanger technologies.

R&D in Support of ASTRID and JSFR: Cross-Analysis and Identification of Possible Areas of Cooperation

The Commissariat à L’Énergie Atomique et aux Énergies Alternatives (CEA) and the Japan Atomic Energy Agency (JAEA) intend to develop prototype or demonstration sodium-cooled fast reactors (SFRs) within two decades. The common final goal of their respective programs is SFR commercialization. The target of a commercial SFR for CEA and JAEA is basically consistent with Generation IV goals. Based on their industrial backgrounds and feedback from past and existing reactor experiences, CEA and JAEA have selected pool and loop configurations for the Advanced Sodium Technological Reactor for Industrial Demonstration (ASTRID) and the Japanese Sodium-Cooled Fast Reactor (JSFR), respectively. CEA and JAEA have cross-analyzed both pool and loop concepts (ASTRID and demonstration JSFR, respectively). The analysis results show that both concepts are technologically feasible and meet design goals. From the viewpoint of collaboration, the analysis identifies a wide range of collaborative areas in several fields: design principles (e.g., design target and design standard); development of components and systems; development of component parts, devices, or subsystems; design methods; simulation tools; etc.

Sodium-Cooled Fast Reactors: the ASTRID Plant Project

ln the framework of the French Act, the 28th of June 2006, about nuclear materials and waste management, an actinides incineration demonstration prototype is to be commissioned in the 2020 decade.CEA has launched this prototype, called ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) to demonstrate the performances and the progress made in SFR technology on an industrial scale by qualifying innovative options to meet the objectives of the fourth generation reactors; some of these options still remain open in the areas requiring research and development programs, especially in safety and reliability.ASTRID will consider the inspection and repair from the beginning of the design and will also be used as a test bench for advanced techniques in this field.More specifically, we are aiming for a level of safety that is at least equivalent to that of the third generation reactors, with progress made in SFR-specific fields.To meet the requirements of the above-mentioned act, ASTRID will also be designed to investigate and to demonstrate the feasibility of transmutation of radwaste: it will be replacing the PHENIX plant in terms of an irradiation reactor so as to test homogenous and heterogeneous minor actinide recycling modes, whose industrial interest is currently being assessed in the frame of the Act. The first deadline set by the French Act of 28 June 2006 is for late 2012, at which time the authorities must be provided with an assessment of the industrial perspectives on actinides transmutation. To remain consistent with the general schedule fixed by the Act, the CEA decided to launch the phase of the preliminary design of the ASTRID prototype in 2010 in order to be able to present at the same occasion (late 2012) preliminary budgetary evaluations (especially in terms of amount investment schedule), and technical facts so that a decision on the following steps can be reached. An assessment of the fixed and still open-ended options will be established at this date, together with the planning and cost evaluation (including R&D) of next phases.Phase 1 of the preliminary design also aims at obtaining an initial opinion from the French Nuclear Safety Authority (ASN) on the safety orientations recommended for ASTRID, in order to check their suitability.Phase 2 of the preliminary design will take place between 2013 and 2014, including the drafting and submittal to ASN of the Safety Options Report, a deeper assessment of retained options and hard points, and a consolidation of cost and planning figures.ln order to benefit from the unique feedback existing in France on Sodium Fast Reactors, the CEA partnered, among others, with EDF, as experienced SFR operator and general architect, and with AREVA, as experienced SFR Nuclear Island engineer and component’s designer.This paper presents the current state of the project, the first design options and the main milestones.