[en] The Helium Cooled Pebble Bed (HCPB) blanket is one of the two concepts to be further developed for a European DEMO power reactor. Tritium leakage due to the permeation through the main structural materials constitutes one of the main problems of the fusion technology. In the present work, the tritium permeation through and inventory in the first wall between the plasma and the first wall coolant has been evaluated. The influence of temperature gradient, surface conditions, isotopic effect, and trapping in ion- and neutron-induced defects on the hydrogen isotope permeation and inventory has been considered

[en] Since the beginning of space activities, solar and nuclear energy have been identified as the only available options for extended missions according to present knowledge. Both types have been used extensively for missions on Earth orbit, interplanetary space and planetary/lunar surface. However, the intensity of solar irradiation decreases with the square of the distance from the Sun. Future scientific and human explorations will take benefits of using a safe in-space nuclear reactor for providing both sufficient electric energy and efficient performance for a space propulsion. The first part of the paper presents a brief status of the different types of nuclear power sources, their characteristics and their field of applications. Previous CEA's projects of space nuclear fission reactors that have been studied in the past will also be discussed; the ERATO project in the 80's (design of a Nuclear Electric Propulsion system of 20 to 200 kWe) and the MAPS project in the 90's (definition of a Nuclear Thermal Propulsion system of 300 MWth for 72 kN of thrust). According to the recent road-maps, CEA decided to maintain a waking state in its spatial nuclear activities by carrying out some conceptual design studies of Nuclear Electric Power systems in the range of 100-500 kWe. The second part of the paper describes the main characteristics of this Optimized Propulsion Unit System (OPUS studies) and its different components. These characteristics, the basic options of the OPUS system that have been selected and the reasons associated to those choices are examined. Especially, the nuclear reactor has been defined considering the possible synergies with the next generation of terrestrial nuclear reactor (International Generation IV Forum). After two successive sets of studies, two different versions of this nuclear system have emerged. The first one is a fast, high-temperature helium cooled reactor, coupled to a direct reheated Brayton cycle. This version is technically the easiest one: it uses simple, well-documented elements, an inert working fluid, and could run at relatively lower temperature if it had to be built next year. However, even if the mass-to-power ratio is acceptable for a 100 kWe, (about 24 kg/kWe) and is expected to decrease significantly with the power, the problematic size of the radiator used for heat rejection limits the power to 100 kWe for non-deployable radiators. Still, this first version must be taken into consideration because of its feasibility, regarding the actual advance in materials technology. A more ambitious version is now being designed and is expected, among other progresses, to dramatically reduce the size of the radiator. Such benefits come with other problems: one of the studied working fluids is the sulfur, used at both vapor and liquid states, which is very poorly documented compared to other fluids used in nuclear technology. Hence, if the specific mass and volume of this version are theoretically lower, this project would need some technical progress in materials science before its execution. At the same time, it is shown that the maximum allowable neutron fluence for both fuel and core materials remains very acceptable in both versions, which suggests that the OPUS project could also be interesting for higher power levels. The paper address the following issues: The nuclear power sources for space; The RHU and RTG radioisotope systems; The reactor system; Previous studies in France; Review of the design considerations for nuclear reactor-based systems; The OPUS approach; The choices for the reactor core; The choices for the conversion cycle; The OPUS basic features; Reactivity control; The shield unit; The radiator; The conversion system; An innovative option for OPUS. In conclusion, the present OPUS studies document a point design of a progressive space power nuclear system that possesses adequate flexibility to accommodate a broad range of science missions and that is compatible with available technology. The point design relies on either Brayton or Hirn conversion cycle and is built around a same fast reactor concept. Such a system is being able to achieved quite acceptable specific mass (α parameter in kg/kWe) in the power range 100-500 kWe. Starting with an α of 24 kg/kWe with the Brayton version and 18 kg/kWe with the more ambitious version based on the Hirn cycle, the OPUS system allows envisaging more powerful units with mass-to-power ratios well below 10 kg/kWe. Moreover, the slightly oversized reactor core presented in this study does not penalize drastically the global α parameter and let to reach long operating times in addition to a high power level

[en] AREVA's joint subsidiary with Siemens, Framatome ANP, has launched the ANTARES Program (AREVA New Technology based on Advanced gas-cooled Reactor for Energy Supply) for the development of an advanced commercial HTR reactor for electricity generation and process heat supply. In this context, specific development and qualification R and D programs were established with CEA to set-up NEPHTIS (Neutronics Process for HTR Innovating System), a new industrial neutronics code system for the computation of V/HTR reactors. After a brief overview of the ANTARES Program, this paper aims to present the characteristics of the HTR neutronics code system NEPHTIS and to show the validation performed against reference (Monte-Carlo) codes and versus experimental results. Nephis is a deterministic calculation scheme dedicated to the computation of prismatic block-type HTR cores. It is based on a usual two-step approach. First, the fuel element is calculated in 2D transport theory in its true heterogeneous representation (including fuel particles double-heterogeneity) using the APOLLO2 spectral code with a very fine energy meshing (172 groups). Then, this calculation provides condensed (8 groups) and homogenized cross sections to the core simulator CRONOS2, which performs 3D calculations in diffusion theory with finite-elements. Validations were performed on several geometry configurations (assembly/core; 2D/3D) for UOX fuel. The two-steps approach used in NEPHTIS is validated by comparisons with an APOLLO2 reference calculation (whole core computed in 2D transport theory) and with Monte-Carlo calculations. Specific complementary validations (e.g. control rod worth), using TRIPOLI and MCNP calculations, are presented as well in this paper. Calculations vs. experiment (C/E) comparisons have been performed on the two main existing block-type HTR reactors: HTTR and Fort Saint-Vrain (FSV). NEPHTIS results were compared with the HTTR benchmark, proposed by IAEA in 1999, on the first criticality. Several configurations from annular to full core geometry were computed with NEPHTIS and successfully compared to reactivity Monte-Carlo and experimental results. NEPHTIS also proved its capability to accurately calculate the first criticality results of the FSV reactor, and to predict the discharged isotopic content of one standard element after 32 GWd/T of irradiation in FSV (C/E lower than 3% on the main nuclides). (authors)

[en] In a High Temperature Reactor, tritium is produced by a number of mechanisms. Due to its high mobility, some of this tritium ends up in the primary helium cooling circuit from where it can be extracted by the coolant purification system to keep the partial pressure of tritiated compounds low. The remaining partial pressure of tritium in the coolant is the driving force for permeation across the heat exchanger from the primary cooling system into the secondary cooling system. From there the contamination may further propagate and ultimately escape into the environment. This paper summarizes a study on the different tritium control options capable of meeting possible future safety requirements. Our results indicate that compliance with plausible tritium control requirements can indeed be achieved with reasonable effort both for electricity generation using a closed steam cycle and for process steam generation with an open steam cycle. However, for new-build HTR, definite country-specific licensing requirements (e.g. chronic and accidental tritium release) are yet to be determined and will shape the required tritium control strategy. (author)

[en] In a high temperature reactor, tritium is produced by a number of mechanisms. Due to its high mobility, some of this tritium ends up in the primary helium cooling circuit from where it can be extracted by the coolant purification system to keep the partial pressure of tritiated compounds low. The remaining partial pressure of tritium in the coolant is the driving force for permeation across the heat exchanger from the primary cooling system into the secondary cooling system. From there the contamination may further propagate and ultimately escape into the environment. This paper summarizes a study on the different tritium control options capable of meeting possible future safety requirements. Our results indicate that compliance with plausible tritium control requirements can indeed be achieved with reasonable effort both for electricity generation using a closed steam cycle and for process steam generation with an open steam cycle. However, for new-build HTR, definite country-specific licensing requirements (e.g. chronic and accidental tritium release) are yet to be determined and will shape the required tritium control strategy.

[en] Recent progress in the field of high-temperature materials and components has put these types of reactors again in the forefront of research and development. After the work already initiated on Magnox reactors, gas-cooled reactors have been considerably modernized and could eventually challenge the current supremacy of PWRs, at least in certain areas or specific applications. Industrial-scale implementations of slow-neutron versions are foreseeable in the medium term. Fast-neutron versions (more prospective) offer additional possibilities for energy-efficient use of natural uranium resources by using a fuel cycle that minimises final waste production and proliferation risks. This monograph covers the current research on these types of reactors, describing the research challenges involved, the recent results achieved at the CEA and the obstacles that remain to be overcome

[en] This reference book addresses the study of neutron transport in matter, the study of conditions for a chain reaction and the study of modifications of matter composition due to nuclear reactions. This book presents the main nuclear data, their measurement, assessment and processing, and the spallation. It proposes an overview of methods applied for the study of neutron transport: basic equations and their derived forms, deterministic methods and Monte Carlo method of resolution of the Boltzmann equation, methods of resolution of generalized Bateman equations, methods of time resolution of space kinetics coupled equations. It presents the main calculation codes, discusses the qualification and experimental aspects, and gives an overview of neutron transport applications: neutron transport calculation of reactors, neutron transport coupled with other disciplines, physics of fuel cycle, criticality

[en] This bibliographical note presents a reference book which addresses the study of neutron transport in matter, the study of conditions for a chain reaction and the study of modifications of matter composition due to nuclear reactions. This book presents the main nuclear data, their measurement, assessment and processing, and the spallation. It proposes an overview of methods applied for the study of neutron transport: basic equations and their derived forms, deterministic methods and Monte Carlo method of resolution of the Boltzmann equation, methods of resolution of generalized Bateman equations, methods of time resolution of space kinetics coupled equations. It presents the main calculation codes, discusses the qualification and experimental aspects, and gives an overview of neutron transport applications: neutron transport calculation of reactors, neutron transport coupled with other disciplines, physics of fuel cycle, criticality