F. Ogando

The accomplishment of the Engineering Design Activities of IFMIF/EVEDA: The European-Japanese project towards a Li(d,xn) fusion relevant neutron source

The International Fusion Materials Irradiation Facility (IFMIF), presently in its Engineering Validation and Engineering Design Activities (EVEDA) phase under the frame of the Broader Approach Agreement between Europe and Japan, accomplished in summer 2013, on schedule, its EDA phase with the release of the engineering design report of the IFMIF plant, which is here described. Many improvements of the design from former phases are implemented, particularly a reduction of beam losses and operational costs thanks to the superconducting accelerator concept, the re-location of the quench tank outside the test cell (TC) with a reduction of tritium inventory and a simplification on its replacement in case of failure, the separation of the irradiation modules from the shielding block gaining irradiation flexibility and enhancement of the remote handling equipment reliability and cost reduction, and the water cooling of the liner and biological shielding of the TC, enhancing the efficiency and economy of the related sub-systems. In addition, the maintenance strategy has been modified to allow a shorter yearly stop of the irradiation operations and a more careful management of the irradiated samples. The design of the IFMIF plant is intimately linked with the EVA phase carried out since the entry into force of IFMIF/EVEDA in June 2007. These last activities and their on-going accomplishment have been thoroughly described elsewhere (Knaster J et al [19]), which, combined with the present paper, allows a clear understanding of the maturity of the European–Japanese international efforts. This released IFMIF Intermediate Engineering Design Report (IIEDR), which could be complemented if required concurrently with the outcome of the on-going EVA, will allow decision making on its construction and/or serve as the basis for the definition of the next step, aligned with the evolving needs of our fusion community.

Innovative ignition scheme for ICF—impact fast ignition

A new ignition scheme is proposed, in which the compressed DT main fuel is ignited by impact collision of another fraction of separately imploded DT fuel, which is accelerated in the hollow conical target to super high velocities of about (1–2) × 108 cm s−1. Its kinetic energy is directly converted into thermal energy corresponding to temperatures >5 keV on the collision with the main fuel, and this self-heated portion plays the role of ignitor. The ignitor shell is irradiated typically by nanosecond pulses at intensities well beyond 1015 W cm−2 at such a short laser wavelength as 0.25 µm to exert ablation pressures of 150–300 Mbar. A preliminary two-dimensional hydrodynamic simulation demonstrates substantial heating of 3–5 keV on the impact. Simple physics, potential for high gain designs and low cost—these are the crucial advantages of the present scheme.

First IFMIF/EVEDA Radioprotection Studies for the Preliminary Design of the Accelerator Beam Dump

A preliminary beam dump cartridge design has been proposed recently for the IFMIF-EVEDA accelerator. Copper was the material chosen for the beam stop. In this paper we investigate the possibility of designing a practical shielding for the proposed cartridge so that it can offer an acceptable radioprotection response during both beam-on and beam-off phases. The radioprotection analysis is performed for the whole beam dump component located inside the already designed accelerator vault. A comprehensive methodology has been proposed to deal with the problem. Special attention has been paid to the treatment of the neutron source and a significant effort has been devoted to validation purposes. It is justified that prompt and residual dose rates can be provided with a reasonably conservative margin. A base line shielding consisting of a 1 m water tank and a concrete shield of 1 m thickness in front of the tank can be a good approach to fulfill the radioprotection requirement assigned to the beam-on phase. This approach will not be acceptable for the beam-off but it is seen that a feasible solution can be reached by adding a plug at the entrance of the beam dump.

Development of the R2SUNED Code System for Shutdown Dose Rate Calculations

R2SUNED is a code system implementing the mesh-based Rigorous-two-step method for shutdown dose rates calculations, making use of MCNP and ACAB codes. In addition to the most relevant features of state-of-the-art R2S systems, novel and unique features have been implemented in R2SUNED to overcome limitations common to the current mesh-based R2S implementations. One of particular interest is the cell-under-voxel approach intended to address most of the issues associated with the conventional averaging technique used in the neutron flux determination. The underlying idea is to identify the cells enclosed in each voxel and calculate the average value of the neutron flux within each cell fraction. Subsequently, the activation of each material, filling the cell delimited by the voxel, is calculated using the neutron flux evaluated in the corresponding cell. This capability enables to properly resolve the strong spatial gradients of the neutron flux independently of geometrical considerations. Another relevant feature is that a flexible decay gamma source sampling has been incorporated to assure an efficient decay gamma transport in a vast diversity of problems. A verification of R2SUNED has been addressed on the ITER computational shutdown dose rate benchmark. It is highlighted both the limitations of voxel averaged neutron flux standard approach, and the ability of R2SUNED to overcome this issue. The excellent agreement of the results with those obtained using a mesh matching perfectly the geometry tells that R2SUNED performs correctly. Finally, a description of the most representative applications carried out with R2SUNED is provided for fusion relevant facilities.

Activation Analysis for a He/LiPb dual Coolant Blanket for DEMO Reactor

Abstract The objective of the Spanish national project TECNO_FUS is to generate a conceptual design of a DCLL (Dual-Coolant Lithium-Lead) blanket for the DEMO fusion reactor. The dually-cooled breeding zone is composed of He/Pb-15.7 6Li and SiC as liquid metal flow channel inserts. Structural materials are ferritic-martensitic steel (Eurofer-97) for the blanket and austenitic steel (316LN) for the Vacuum Vessel (VV). The goal of this work is to analyze the radioactive waste production by the neutron-induced activation and the back-end of the blanket and the VV (SS316LN) materials (Eurofer, SiC, LiPb, and SS316LN). Furthermore, the radioactive waste production in the cryostat (SS316LN) and the bioshielding (concrete) has been estimated. Following the current approach to the back-end of the materials in fusion facilities, the radioactive waste has been subdivided according to the activity-level classification (EW, exempted waste, LILW, low and intermediate level waste, and HLW, high level waste) and according to the radiological complexity of operations (handling and cooling). The activation calculations have been carried out with the ACAB code.

Radioprotection Analysis for the High Energy Beam Transport Line of the Accelerator Facility of IFMIF

Abstract The safety analysis for the design of the IFMIF accelerator facility is an ongoing task within the IFMIF project, and the radioprotection analysis is one of the main sections of this task. The high energy beam transport line is a very sensitive section since it contains intense radiation sources. This work presents the radiological analysis of its current layout, and the design of local radiation shields, in order to meet project dose limits. These limits are established to protect workers in neighboring rooms during operation and to ensure unrestricted access for maintenance operations.

Sensitivity to Nuclear Data in the Radioprotection Design of the LIPAC (IFMIF/EVEDA) Beam Dump

Abstract Linear IFMIF Prototype Accelerator (LIPAC) is the prototype accelerator of the Engineering Validation and Engineering Design Activities (EVEDA) phase of the International Fusion Materials Irradiation Facility (IFMIF) project. The EVEDA phase is a first IFMIF step devoted to the construction of prototypes of the main units. The deuteron beam of LIPAC (125 mA, 9 MeV) is stopped by a conical copper beam stop, giving rise to neutron and photon sources that must be shielded to comply with dose requirements. A reliable characterization of these secondary sources is a mandatory task. The built-in-semi-analytical nuclear models used by advanced Monte Carlo transport codes as Monte Carlo N-Particle eXtended (MCNPX) or Particle and Heavy Ion Transport code System (PHITS) have been demonstrated as unreliable for describing deuteron interactions and secondary particle production at these low energies. The use of reliable external nuclear data is consequently necessary in the design of the LIPAC shielding. In particular, the TENDL-2010 library has been compared with recently published experimental data demonstrating its reliability for deuteron interaction on copper at 9 MeV. The Monte Carlo Universidad Nacional de Educación a Distancia (MCUNED) code has been developed to make use of external nuclear data, and its use with the TENDL-2010 library has proven very satisfactory for LIPAC radioprotection analysis. The impact on radioprotection tasks in LIPAC when the unreliable nuclear models mentioned above are used is discussed.

Interpolation for momentum conservation in 3D toroidal gyrokinetic particle simulation of plasmas

Abstract A linear momentum conserving interpolation for the electric field in a three-dimensional toroidal particle-in-cell gyrokinetic plasma simulation in a tokamak configuration is found unstable due to the false divergence of E → × B → flow of particles. A paradigm 8-point finite difference for interpolation of the radial electric field on the poloidal plane is proposed which stabilizes the simulation.

Neutronic Assessment of Candidate Materials for TF Coils Shielding in a DEMO Fusion Reactor Based on a DCLL Blanket

Abstract Under the Spanish Breeding Blanket Technology Program TECNO_FUS, a conceptual design of a dual-coolant lithium-lead (DCLL) blanket for DEMO is being revisited. In this work, different shielding candidate materials are assessed in their ability to satisfy the radiation load requirements that must be fulfilled in the toroidal field (TF) coils: absorbed dose in the insulator (Epoxy), peak fast neutron fluence in the superconductor (Nb3Sn), peak nuclear heating in the winding pack and maximum neutron fluence in the cooper stabilizer. Furthermore, the impact of the material choice on waste management requirements of both shielding and vacuum vessel (VV) materials is evaluated, and the performance of candidate materials is examined in terms of the helium production in the VV SS316LN material and its implications in reweldability. Materials discussed for the High Temperature Shield are Eurofer, graphite, B4C, WC and WB4C, while the metal hydrides ZrH2, Zr(BH4)4, and TiH2 are discussed for the Low Temperature Shield. In the case of DEMO irradiation scenario, all the analyzed material combinations fulfill the design requirements for the waste management of the shield and VV, He production in the VV wall and TF coils radiation loads requirements.

Advances in implosion physics, alternative targets design, and neutron effects on heavy ion fusion reactors

The coupling of a new radiation transport (RT) solver with an existing multimaterial fluid dynamics code (ARWEN) using Adaptive Mesh Refinement named DAFNE, has been completed. In addition, improvements were made to ARWEN in order to work properly with the RT code, and to make it user-friendlier, including new treatment of Equations of State, and graphical tools for visualization. The evaluation of the code has been performed, comparing it with other existing RT codes (including the one used in DAFNE, but in the single-grid version). These comparisons consist in problems with real input parameters (mainly opacities and geometry parameters). Important advances in Atomic Physics, Opacity calculations and NLTE atomic physics calculations, with participation in significant experiments in this area, have been obtained. Early published calculations showed that a DTx fuel with a small tritium initial content (x<3%) could work in a catalytic regime in Inertial Fusion Targets, at very high burning temperatures (⪢100 keV). Otherwise, the cross-section of DT remains much higher than that of DD and no internal breeding of tritium can take place. Improvements in the calculation model allow to properly simulate the effect of inverse Compton scattering which tends to lower Te and to enhance radiation losses, reducing the plasma temperature, Ti. The neutron activation of all natural elements in First Structural Wall (FSW) component of an Inertial Fusion Energy (IFE) reactor for waste management, and the analysis of activation of target debris in NIF-type facilities has been completed. Using an original efficient modeling for pulse activation, the FSW behavior in inertial fusion has been studied. A radiological dose library coupled to the ACAB code is being generated for assessing impact of environmental releases, and atmospheric dispersion analysis from HIF reactors indicate the uncertainty in tritium release parameters. The first recognition of recombination barriers in SiC, modify the understanding of the calculation of displacement per atom, dpa, to quantify the collisional damage. An important analysis has been the confirmation, using Molecular Dynamics (MD) with an astonishing agreement, of the experimental evidence of low-temperature amorphization by damage accumulation in SiC, which could modify extensively its viability as a candidate material for IFE (fusion in general) applications. The radiation damage pulse effect has also been assessed using MD and Kinetic Monte Carlo diffusion of defects, showing the dose and driver frequency dependences.