[en] The specifications of permeation barriers, tritium recovery process maintaining a very low tritium activity in the coolant, and control of the coolant chemistry, required the evaluation of the tritium losses through the steam generators and include the definition of its operating conditions by thermodynamic cycle calculations and its thermal-hydraulic design. For both tasks specific computer tools were developed. The obtained geometry, surface area, and temperature profiles along the heat exchanger tubes were then used to estimate the daily tritium permeation into the steam cycle. Steam oxidized Incoloy 800 austenitic stainless steel was identified as the best suited existing material; in nominal steady-state operation, the tritium escape into the steam cycle could be restricted to less than 10 Ci/d. Tritium permeation during temperature and pressure transients in the steam generator (destruction and possible self-healing of the permeation barrier) is identified to bear a large tritium release potential. Solutions are proposed. (from authors). 4 figs., 1 tab

[en] The potential sources of tritium contamination of the helium-coolant of ceramic breeder blankets have been evaluated in a previous paper for the specific case of the European BIT DEMO blanket. This evaluation associated with a rough assessment of the permeability to tritium of the tubing of helium-heated steam generators confirmed that the control of tritium losses to the steam circuit is a critical issue for this class of blanket requiring developments in three areas: (1) permeation barriers, (2) tritium recovery processes maintaining a very low concentration in tritiated species in the coolant, and (3) methods for controlling the chemistry of the coolant. Consequently, in order to define the specifications of these developments, a detailed evaluation of the permeability to tritium of helium-heated steam generators (SGs) was performed, which will be reported in this paper. This study includes the definition of the thermal-hydraulic operating conditions of the SGs through thermodynamic cycle calculations, and its thermal-hydraulic design. The obtained geometry, area and temperature profiles along the tubes are then used to estimate, based on relevant permeability data, the tritium permeation through the SG as a function of the composition in tritiated species of the coolant. The implications of these results, in terms of requirements for the considered tritium control methods, will also be discussed on the basis of expected limits in tritium release to the steam circuit

[en] The CEA R and D program on advanced Gas Cooled Reactors (GCR) relies on different concepts: modular High Temperature Reactor (HTR), its evolution dedicated to hydrogen production (Very High Temperature Reactor) and Gas Cooled Fast Reactors (GCFR). Some key safety questions are related to decay heat removal during potential accident. This is strongly connected to passive natural convection (including gas injection of Helium, CO², Nitrogen or Argon) or forced convection using active safety systems (gas blowers, heat exchangers). To support this effort, thermal-hydraulics computer codes will be necessary tools to design, enhance the performance and ensure a high safety level of the different reactors. Accurate and efficient modeling of heat transfer by conduction, convection or thermal radiation as well as energy storage are necessary requirements to obtain a high level of confidence in the thermal-hydraulic simulations. To achieve that goal a thorough validation process has to ve conducted. CEA's CAST3M code dedicated to GCR thermal-hydraulics has been validated against different test cases: academic interaction between natural convection and thermal radiation, small scale in-house THERCE experiments and large scale High Temperature Test Reactor benchmarks such as HTTR-VC benchmark. Coupling with neutronics is also an important modeling aspect for the determination of neutronic parameters such as neutronic coefficient (Doppler, moderator,...), critical position of control rods...CEA's CAST3M and CRONOS2 computer codes allow this coupling and a first example of coupled thermal-hydraulics/neutronics calculations has been performed. Comparison with experimental data will be the next step with High Temperature Test Reactor experimental results at nominal power

[en] In the frame of the European contract HTR-N, a work package is devoted to the code validation and method improvements as far as the high temperature gas-cooled reactor (HTGR) core modelling is concerned. Institutions from three countries are involved in this work package: FZJ in Germany, NRG and IRI in the Netherlands, and CEA in France. The present work is based on a benchmark problem proposed by JAERI through the IAEA. It concerns the HTTR's start-up core physics experiments that were a good opportunity for the European partners to validate their calculational tools and methods. The number of fuel columns necessary to achieve the first criticality and the excess reactivity for 18, 24, and 30 fuel columns in the core had to be evaluated. Pre-test and post-test calculational results, obtained by the partners, are compared with each other and with the experiment. Parts of the discrepancies between experiment and pre-test predictions are analysed and tackled by different treatments. In the case of the Monte Carlo code TRIPOLI4, used by CEA, the discrepancy between measurement and calculation at the first criticality is reduced to Δk/k∼0.85%, when considering the revised data of the HTTR benchmark [Fujimoto, private communication]. In the case of the diffusion codes, this discrepancy is reduced to Δk/k∼0.8% (FZJ) and 2.7 or 1.8% (CEA)

[en] Following the present tendency of the international community, the French industrial partners are greatly interested in the technical and economical potential performances of the HTGR (High Temperature Gas cooled Reactor) concept that appears as a promising reactor for the future nuclear power applications. However, the next nuclear program related studies will rely on a major part on the intensive use of computational calculations rather than experimental results. Therefore, the codes will have to be reliable and well prepared to calculate a wide range of applications. The computational tools will have to serve as well for conceptual studies and industrial calculations as for best estimate and reference calculations. Taking advantages of the benchmark problems of the HTTR's start-up core physics experiments initially proposed by JAERI through the IAEA in a Coordinated Research Program, CEA has performed calculations in order to validate and qualify the codes that will serve to evaluate the future HTGR generation. For the analysis, two different calculation schemes based on a deterministic or a probabilistic approach are used. It turns out that all calculation results obtained for the fully loaded core configuration fit well each other and with the experiment, considering the experimental uncertainties. As far as the diffusion calculations are concerned, the number of fuel columns needed to achieve criticality increases by about 7 or 8 in comparison with the results obtained with a simplified Transport - Diffusion calculation scheme (homogenized fuel elements and no streaming effect taken into account in the diffusion calculations). (author)

[en] One of the very attractive HTGR reactor characteristics is its highly versatile and flexible core that can fulfil a wide range of diverse fuel cycles. Based on a GTMHR-600 MWth reactor, analyses of several fuel cycles were carried out without taking into account common fuel particle performance limits (burnup, fast fluence, temperature). These values are, however, indicated in each case. Fuel derived from uranium, thorium and a wide variety of plutonium grades has been considered. Long-lived actinide production and total residual decay heat were evaluated for the various types of fuel. The results presented in this papers provide a comparison of the potential and limits of each fuel cycle and allow to define specific cycles offering lowest actinide production and residual heat associated with a long life cycle. (author)

[en] The CEA, in collaboration with EDF and AREVA-NP, is developing a core modelling tool called NEPHTIS, for Neutronic Process for HTGR Innovating Systems and dedicated at present day to the prismatic block-type HTGR (High Temperature Gas-Cooled Reactors). Due to the lack of usable HTGR experimental results, the confidence in this neutronic computational tool relies essentially on comparisons to reference or best-estimate calculations. In the present analysis, the Aleppo deterministic transport code has been selected as reference for validating core depletion simulations carried out within NEPHTIS. These reference calculations were performed on fully detailed 2D core configurations using the Method of Characteristics. The latter has been validated versus Monte Carlo method for different static core configurations [1], [2] and [3]. All the presented results come from an annular HTGR core loaded with uranium-based fuel (15% enrichment). During the core depletion validation, reactivity, reaction rates distributions and nuclei concentrations have been compared. In addition, the impact of various physical and geometrical parameters such as the core loading (one-through or batch-wise reloading) and the amount of burnable poison has been investigated during the validation phases. The results confirm that NEPHTIS is able to predict the core reactivity with uncertainties of ±350 pcm. At the end of the core irradiation, the U-235 consumption is calculated within ± 0, 7 % while the plutonium mass discharged from the core is calculated within ±1 %. As far as the core power distributions are concerned, small discrepancies ( and < 2.3 %) can be observed on the fuel block-averaged power distribution in the core. (authors)

[en] Since a few years now, Cea decided to maintain a waking state in its space nuclear activities by carrying out some conceptual studies of embarked nuclear power systems in the range of 100-500 kWe. Results stemming from these ongoing studies are gathered in the project OPUS -Optimized Propulsion Unit System-. This nuclear power system relies on a fast gas-cooled reactor concept coupled either to a Brayton cycle or to a more ambitious energy conversion system using a Hirn cycle to dramatically reduce the size of the radiator. The OPUS reactor core consists of an arrangement of enriched graphite elements of hexagonal cross-section. Their length is equal to the core diameter (48 cm). Coated fuel particles containing enriched (93%) uranium are embedded in these fuel elements. Each fuel element is designed with a centered axial channel through which flows the working fluid: a mixture of helium and xenon gas. This reactor is expected to have an operating life of over 2000 days at full power. In fact the main questions remain on the fuel element manufacturing and on the mechanical design (type and size of particles, packing fraction in the matrix, final core diameter and mass). Especially, the nuclear reactor has been defined considering the possible synergies with the next generation of terrestrial nuclear reactor (International Generation IV Forum). Based on relatively short-term technologies, the same reactor is designed to cover a wide range of power: 100 to 500 kWe without core design modification. The final reactor design presented in this paper is the result of a coupled analysis between the thermomechanical and the neutronic aspects

[en] The very high temperature reactor (VHTR) is one of the six projects selected in the Generation-IV nuclear power plant research program. The VHTR is a graphite-moderated, helium-cooled reactor with a once-through uranium fuel cycle. It is designed to be a high-efficiency power production system, flexible to adopt uranium/plutonium fuel cycle, minimize waste production and retaining the desirable safety characteristics offered by its modularity. The bundling of dispersed fuel made in coated particles and the graphite as moderator leads to special core physics. The strong neutronic and thermal-hydraulics spatial dependency is one of the most important consequences of this bundling. This paper is a contribution to VHTR core studies. It presents the improvements made on the neutronic and thermal-hydraulics coupling system of the French Atomic Energy Agency (CEA). Especially the use of spatial de-homogenization temperature model and the possibility to insert the operating and start-up control systems into the coupling model. In addition, it exposes some preliminary results of VHTR core depletion calculation. The paper discusses also the reflector thermal effect and the location of both neutronic and thermal hottest area during the fuel cycle and it is consequence on core safety. (authors)

[en] Whatever the future of the civil nuclear programme in France may be, the plutonium reprocessing and recycling option has been chosen 14 years ago and the control of the plutonium inventory appears today as a major R and D issue. Many studies in progress at Cea attempt to improve plutonium recycling in PWR by increasing the amount of plutonium fed in the core, using inert matrix, new design. Moreover, in spite of their good performances and safe behaviour, innovative reactor concepts considered at the present time must also demonstrate their capacity to use at best the plutonium matter that represents at the same time a great energetic potential and strong radio-toxic source in spent fuel. In this context and with regard to the renewed interest in the High Temperature Gas-cooled Reactor (HTGR) concept, the problem of the mastery of the plutonium stock with the help of the HTGR has been undertaken at Cea in collaboration with Framatome. (author)