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Integration of safety aspects in IFMIF-DONES design is a main objective of EUROfusion and European Commission projects. IFMIF-DONES will be a radioactive facility of the first category, and stringent safety objectives must be achieved and demonstrated. A very low acceptable risk for the worker, the public and the environment is the main principle in the design phase. The progress of safety activities is performed iteratively as detailed engineering develops, taking into account the uniqueness of the facility: a high-power deuterons accelerator (125 mA, 40 MeV), a target of flowing liquid lithium, traps for activation products, a dedicated-design module for irradiated samples, a massive shielding cooled room with confinement function, and a number of conventional systems with safety functions. Several phases are developed: (i) identification of sources and materials at risk, radioactive and nonradioactive, subject to potential mobilization, (ii) failure mode analysis and effects of systems, starting at the functional level, and support with probabilistic analysis, (iii) identification of scenarios leading to unacceptable risk if unmitigated, (iv) proposal of layers of defense by means of safety-credited components and design features, (v) deterministic analysis of scenarios in support of requirements, and (vi) definition and demonstration of safety requirements charged to components. [ABSTRACT FROM AUTHOR]

IFMIF-DONES (International Fusion Material Irradiation Facility- DEMO Oriented NEutron Source) is an accelerator based irradiation facility which is aimed to provide the irradiation data needed for the construction of DEMO. The Test Cell (TC) is the central room of IFMIF-DONES enclosing the target and the test module. This work presents the neutronics analyses conducted to provide the relevant nuclear responses for the bio-shield of the TC, including heating, radiation damage and helium production. The results for the TC steel liners covering the inner TC surface have been obtained with an unstructured mesh based interpolation approach. The results show that the softening of the neutron spectrum at the thin layer of the inner wall increases the nuclear heating as well as the helium production in SS316 L. The helium production might be of major concern if re-welding of the liner is required during the lifetime. The impact of a reduced beam footprint size on the TC nuclear responses has been analyzed and found to be insignificant comparing with the normal footprint. [ABSTRACT FROM AUTHOR]

OpenMC is a community-developed Monte Carlo neutron and photon transport simulation code. It is capable of performing fixed source particle transport simulation based on continuous-energy nuclear cross-section data. In this work, OpenMC has been benchmarked and verified for the application to accelerator-based neutron sources, e.g. the International Fusion Material Irradiation Facility- DEMO Oriented NEutron Source facility (IFMIF-DONES). To this end, the McDeLicious code, which is an extension of the MCNP Monte Carlo code with the capability to simulate the deuterium-lithium neutron source on the basis of evaluated d + 6,7Li cross section data, has been migrated to the OpenMC code. The OpenMC-McDeLicious has been tested for the calculation of uncollided neutron and photon flux and spectra by comparing to results obtained with MCNP6-McDeLicious. SINBAD experimental benchmarks and d -Li experiments have been used for validating OpenMC. These tests and verifications show good agreement between OpenMC and MCNP code, as well as the experimental data. In addition, OpenMC-McDeLicious is used for obtaining the nuclear responses, e.g. Displacement Per Atom (DPA), gas production, nuclear heating, in the irradiation test module of the IFMIF-DONES neutron source. Most OpenMC-McDeLicious results show good agreement with MCNP6-McDeLcious within 5%. This indicates the successful benchmarking and validation of OpenMC code for the applications to accelerator-based neutron source facilities. [ABSTRACT FROM AUTHOR]

• Benchmarks of OpenMC against MCNP was conducted. • A problem has been fixed in the secondary photon transport of the OpenMC code. • Results show the suitability of OpenMC for accelerator-based applications. OpenMC is a community developed Monte Carlo (MC) code for neutron and photon transport calculations. In this work, the OpenMC code has been extended and benchmarked for accelerator-based neutron source applications, such as the IFMIF-DONES (International Fusion Material Irradiation Facility- DEMO Oriented NEutron Source) with neutron energies up to 55 MeV. A spherical computational benchmark and the TIARA high energy neutron shielding experiment has been employed for the benchmarking including also comparisons to the results of the well-validated MC code MCNP as reference. The neutron fluxes computed for spherical benchmarks of selected elements show good agreement between OpenMC and MCNP. A problem has been identified and fixed in the secondary photon transport of the OpenMC code. For the TIARA experiment on iron and concrete shielding, the results of OpenMC are in good agreement with MCNP, and the discrepancies between the experiment results and OpenMC simulated results are considered acceptable. In summary, these benchmarks provide good evidence of the suitability of OpenMC for accelerator-based neutron source applications in the energy range up to 55 MeV, with possibility of further extensions depending on the used nuclear data. [ABSTRACT FROM AUTHOR]

• The activation and shutdown dose analyses of the IFMIF-DONES. • Transport and activation calculations using McDeLicious code. • The dose rate maps in the access cell as well as the shielding assessments is presented. • The analysis of the shielding for activated target of DONES. This paper presents the activation and shutdown dose analyses of the IFMIF-DONES (International Fusion Materials Irradiation Facility – DEMO-Oriented NEutron Source) Target assembly. To this end, a series of coupled transport and activation calculations are performed using McDeLicious code, which is based on the MCNP Monte Carlo code for simulating the neutron production from d-Li reactions, and the FISPACT inventory code. The latest DONES test cell model has been employed and FENDL neutron cross-section data library has been used. Shut-down dose rate calculations were performed using the R2Smesh code system developed at KIT. The obtained dose rate maps in the access cell as well as the shielding assessments will be presented in the paper. [ABSTRACT FROM AUTHOR]

The methodological approach applied for the nuclear analyses of the IFMIF-DONES facility, developed within the Early Neutron Source (ENS) project of EUROfusion, is presented. It utilizes dedicated tools, developed and adapted to the specific needs and peculiarities of an accelerator based d-Li neutron source. Major elements relate to the simulation of deuteron interactions with the Li target and the accelerator structures, the generation and transport of neutrons and photons, and the production of radio-active nuclides with the subsequent emission and transport of decay gamma radiation. With such an approach, all neutronics data can be provided which are required to design, optimize and evaluate the systems/sub-systems of the facility with regard to their nuclear and shielding performance. This is shown on the example of nuclear analyses conducted for the latest design of IFMIF-DONES. [ABSTRACT FROM AUTHOR]

Abstract The IFMIF-DONES (International Fusion Materials Irradiation Facility-Demo Oriented NEutron Source) test cell (TC) biological shielding is responsible of providing sufficient protection for the surrounding rooms and cells from intensive neutron irradiation and gamma radiation from inside of the TC. The TC biological shielding is composed of surrounding shielding walls, two top shielding plugs, six piping and cabling plugs (PCPs), TC floor, and removable shielding materials below the TC floor. Based on the previous TC configuration, the design of the lower shielding plug has been further developed by defining the specifications of embedded helium active cooling systems with the support of neutronic simulations, CFD analysis, and remote handling requirements. The PCP design is updated to provide sufficient space for pipe and cable connections. A preliminary piping of the water cooling system inside the heavy concrete shielding walls has been performed to qualify the capability of removing volumetric nuclear heating during irradiation experiments. The geometry of TC surrounding shielding walls are adjusted according to the latest design of PCPs and LSP as well as the requirements of the active cooling pipes which are embedded in the TC surrounding walls. [ABSTRACT FROM AUTHOR]

Computer aided design (CAD) is an important industrial way to produce high quality designs. Therefore, CAD geometries are in general used for engineering and the design of complex facilities like the ITER tokamak. Although Monte Carlo codes like MCNP are well suited to handle the complex 3D geometry of ITER for transport calculations, they rely on their own geometry description and are in general not able to directly use the CAD geometry. In this paper, three different approaches for the use of CAD geometries with MCNP calculations are investigated and assessed with regard to calculation performance and user-friendliness. The first method is the conversion of the CAD geometry into MCNP geometry employing the conversion software McCad developed by KIT. The second approach utilizes the MCNP6 mesh geometry feature for the particle tracking and relies on the conversion of the CAD geometry into a mesh model. The third method employs DAGMC, developed by the University of Wisconsin-Madison, for the direct particle tracking on the CAD geometry using a patched version of MCNP. The obtained results show that each method has its advantages depending on the complexity and size of the model, the calculation problem considered, and the expertise of the user. [ABSTRACT FROM AUTHOR]

The IFMIF Test Cell (TC) design was developed and optimized in the EVEDA phase, and finally the reference TC design was proposed. The present study is devoted to further investigations of open issues on the reference TC design. In order to examine the neutron streaming effect caused by pipe penetrations and gaps around removable shielding plugs, a new geometry model for neutronic analyses has been prepared directly from engineering CAD data by utilizing the McCad conversion software. All removable shielding plugs are separately described in the model and a detailed description of pipes was incorporated into the model. The calculation result suggests that the streaming effect is mitigated if the pipe penetration is designed appropriately, while the gaps around the shielding plugs above the TC have large impact on the radiation dose in the access cell. The concept of the reference TC design has been basically validated from the neutronics point of view, although the streaming effect should be compensated by the shielding capability of the test cell cover plate so that occupational workers can access to the access cell during operation. [ABSTRACT FROM AUTHOR]

Abstract: Neutronics calculations were performed to analyse the parameters of blanket energy multiplication factor (M) and tritium breeding ratio (TBR) in a fusion–fission hybrid reactor for energy production named FDS (Fusion-Driven hybrid System)-EM (Energy Multiplier) blanket. The most significant and main goal of the FDS-EM blanket is to achieve the energy gain of about 1GWe with self-sustaining tritium, i.e. the M factor is expected to be ∼90. Four different fission materials were taken into account to evaluate M in subcritical blanket: (i) depleted uranium, (ii) natural uranium, (iii) enriched uranium, and (iv) Nuclear Waste (transuranic from 33000 MWD/MTU PWR (Pressurized Water Reactor) and depleted uranium) oxide. These calculations and analyses were performed using nuclear data library HENDL (Hybrid Evaluated Nuclear Data Library) and a home-developed code VisualBUS. The results showed that the performance of the blanket loaded with Nuclear Waste was most attractive and it could be promising to effectively obtain tritium self-sufficiency and a high-energy multiplication. [ABSTRACT FROM AUTHOR]