The role of spatial and temporal radiation deposition in inertial fusion chambers: the case of HiPER¿

The first wall armour for the reactor chamber of HiPER will have to face short energy pulses of 5 to 20 MJ mostly in the form of x-rays and charged particles at a repetition rate of 5–10 Hz. Armour material and chamber dimensions have to be chosen to avoid/minimize damage to the chamber, ensuring the proper functioning of the facility during its planned lifetime. The maximum energy fluence that the armour can withstand without risk of failure, is determined by temporal and spatial deposition of the radiation energy inside the material. In this paper, simulations on the thermal effect of the radiation–armour interaction are carried out with an increasing definition of the temporal and spatial deposition of energy to prove their influence on the final results. These calculations will lead us to present the first values of the thermo-mechanical behaviour of the tungsten armour designed for the HiPER project under a shock ignition target of 48 MJ. The results will show that only the crossing of the plasticity limit in the first few micrometres might be a threat after thousands of shots for the survivability of the armour

Industrial Applications of Laser Neutron Source

The industrial applications of the intense neutron source have been widely explored because of the unique features of the neutron-matter interaction. Usually, intense neutron sources are assembled with fission reactors or high energy ion accelerators. The big size and high cost of these systems are the bottle neck to promote the industrial applications of intense neutrons. In this paper, we propose the compact laser driven neutron source for the industrial application. As the first step of our project for the versatile applications of laser driven neutron source, Li-neutron and/or Li-proton interactions have been investigated for the application to the development of Li battery

Tritium clouds environmental impact in air into the Western Mediterranean Basin evaluation

The paper considers short-term releases of tritium (mainly but not only tritium hydride (HT)) to the atmosphere from a potential ITER-like fusion reactor located in the Mediterranean Basin and explores if the short range legal exposure limits are exceeded (both locally and downwind). For this, a coupled Lagrangian ECMWF/FLEXPART model has been used to follow real time releases of tritium. This tool was analyzed for nominal tritium operational conditions under selected incidental conditions to determine resultant local and Western Mediterranean effects, together with hourly observations of wind, to provide a short-range approximation of tritium cloud behavior. Since our results cannot be compared with radiological station measurements of tritium in air, we use the NORMTRI Gaussian model. We demonstrate an overestimation of the sequence of tritium concentrations in the atmosphere, close to the reactor, estimated with this model when compared with ECMWF/FLEXPART results. A Gaussian “mesoscale” qualification tool has been used to validate the ECMWF/FLEXPART for winter 2010/spring 2011 with a database of the HT plumes. It is considered that NORMTRI allows evaluation of tritium-in-air-plume patterns and its contribution to doses

The Effect of Cr Concentration on Defect Energies in FeCr Alloys

FeCr ferritic alloys are leading candidates for structural applications in generation IV fission reactors due to their high resistance to swelling and corrosion. The addition of Cr improves the behavior of the steels under irradiation, but this improvement is non-monotonic. Understandig the changes in the FeCr ferritic alloys microstructure induced by irradiation and the role played by the alloying element (Cr) is needed in order to predict the response of these materials under the extreme conditions in the future nuclear plants. In this work, we present the effect of Cr concentration in a bcc Fe matrix on binding and formation energies of vacancy clusters

Atmospheric modelling of Tritium forms transport: review of capabilities and R&D needs for the assessment of fusion facilities environmental impact

Comunicación presentada en: 38ª Reunión Anual de la Sociedad Nuclear Española celebrada del 17 al 19 de octubre de 2012 en Cáceres.This work models in detail the dispersion of tritium forms and dosimetric impact of selected environmental patterns both inland and in-sea using real topography and forecast meteorological data-fields (ECMWF/FLEXPART). We explore specific values of this ratio in different levels and we examine the influence of meteorological conditions in the HTO behavior for 24 hours. For this purpose we have used a tool which consists on a coupled Lagrangian ECMWF/FLEXPART model useful to follow real time releases of tritium at 10, 30 and 60 meters together with hourly observations of wind (and in some cases precipitations) to provide a short-range approximation of tritium cloud behavior. We have assessed inhalation doses. And also HTO/HT ratios in a representative set of cases during winter 2010 and spring 2011 for the 3 air levels

Interstitial defects stability in FeCr alloys vs Cr concentration

Finding adequate materials to withstand the demanding conditions in the future fusion and fission reactors is a real challenge in the development of these technologies. Structural materials need to sustain high irradiation doses and temperatures that will change the microstructure over time. A better understanding of the changes produced by the irradiation will allow for a better choice of materials, ensuring a safer and reliable future power plants. High-Cr ferritic/martensitic steels head the list of structural materials due to their high resistance to swelling and corrosion. However, it is well known that these alloys present a problem of embrittlement, which could be caused by the presence of defects created by irradiation as these defects act as obstacles for dislocation motion. Therefore, the mechanical response of these materials will depend on the type of defects created during irradiation. In this work, we address a study of the effect Cr concentration has on single interstitial defect formation energies in FeCr alloys

Comparison of the Tritium permeated from ITER Blanket in normal operation and its short range impact of HT over France, Swiss or Spain

Comunicación presentada en: 41ª Reunión Anual de la Sociedad Nuclear Española celebrada del 23 al 25 de septiembre de 2015 en la ciudad de La Coruña.In this paper we consider hydrogen isotope (HT) permeation from a liquid metal (LM) ITER breeder blanket (assuming normal operation and a LM as DCLL or HCLL blanket) as a possible source of a tritium leak or release, as the main,, but not the only, source. The paper presents modeling of short-range low-impact HT gas activity over France, Swiss or Spain after release from ITER for sample 2014 and 2015 local weather conditions. The permeation of hydrogen isotopes is an important experimental issue that needs to be taken into account for the development of a Tritium Breeder Module for ITER [1]. Tritium cannot be confined - without an uncertainty of 5% in the flux permeation - and therefore HT can be detected (e.g. by ionization chambers) as it permeates though the structure of RAFM steel towards the coolant [1]. HT arising from Pb15.7Li, and permeated through Eurofer97, can contaminate other parts of the system and may be transported though the normal-vent detritiation system (NVDS). Real-time forecasts of the transport of tritium in air from the fusion reactor towards off-site far-downwind locations though extended tritium clouds into low levels of the atmosphere is calculated for the short range (up to 24 hours) by the coupling of 2 models the European Centre for Medium Range Weather Forecast (ECMWF) [2] model and the FLEXPART lagrangian dispersion model [3] verified with NORMTRI simulation [4] and implemented in many different cases and scenarios [5, 6, 7]. As a function of daily weather conditions, a release may affect just France or it may move towards Switzerland, under cyclonic circulation, or towards the Iberian Peninsula or Balearic Islands (Spain) when a HIGH produces anticyclonic circulation of air over the Mediterranean Sea

The effect of Cr concentration on single interstitials stability in FeCr alloys

Finding adequate materials to withstand the demanding conditions in future fusion and fission reactors is a real challenge in the development of these technologies. Structural materials are going to be subjected to high irradiation doses and operating temperatures which will affect and modify material properties at a microstructural level. Understanding the changes in the microstructure induced by irradiation is needed in order to predict the response of these materials, ensuring safe and reliable future power plants. High-Cr ferritic/martensitic steels are preferred candidate structural materials due to their high resistance to radiation effects and their good resistance against corrosion. On the other hand, it is well known that these alloys present a problem of embrittlement, which could be caused by the presence of defects created by irradiation as these defects act as obstacles for dislocation motion. Therefore, the mechanical response of these materials will depend on the type of defects created during irradiation. In this work, we address a study of the effect of Cr concentration on single interstitial defect formation energies in FeCr alloys

Deffect energetics in FeCr from empirical interatomic potentials

We present a comparative study between two empirical potentials for the study of FeCr alloys. Vacancy, self- and mixed interstitials formation energies are investigated for both potentials in pure bcc Fe and pure bcc Cr. These results are compared to DFT values reported in literature. Some small Cr clusters in substitutional positions have also been studied in bcc Fe. Finally we have performed some calculations of the vacancy formation energy (Ev) depending on Cr concentration

IFMIF suitability for evaluation of fusion functional materials

The International FusionMaterials Irradiation Facility (IFMIF) is a future neutron source based on the D-Li stripping reaction, planned to test candidate fusionmaterials at relevant fusion irradiation conditions. During the design of IFMIF special attention was paid to the structural materials for the blanket and first wall, because they will be exposed to the most severe irradiation conditions in a fusion reactor. Also the irradiation of candidate materials for solid breeder blankets is planned in the IFMIF reference design. This paper focuses on the assessment of the suitability of IFMIF irradiation conditions for testing functionalmaterials to be used in liquid blankets and diagnostics systems, since they are been also considered within IFMIF objectives. The study has been based on the analysis and comparison of the main expected irradiation parameters in IFMIF and DEMO reactor