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• In IFMIF-DONES lithium loop Be-7 and Activated Corrosion Products (ACP) are produced. • Be-7 and ACP represent a complex and heterogeneous decay gamma source. • CAD2CDGS is a new tool that creates decay gamma sources based on CAD geometries. • Tool methodology, inputs and workflow are explained. Two verification tests are done. • Tool applicability for IFMIF-DONES radiation transport simulations is demonstrated. IFMIF-DONES is a neutron source designed to irradiate materials to be used in future fusion power reactors such as DEMO. The facility is based on a deuteron beam impinging onto a liquid lithium jet to generate the neutron flux. Lithium and Corrosion Products will get activated and produce Be-7 and Activated Corrosion Products (ACP). These products will be distributed along the lithium loop, both dissolved in lithium and deposited locally. This complex gamma source should be properly represented to perform radiological safety studies. A new tool, called CAD2CDGS, was created to represent these sources. This tool creates decay gamma sources based on CAD geometries and codifies them into CDGS format. Sources are specified in the CAD model, allowing a user-friendly approach. The tool is based on Open-Source tools (FreeCAD) and EUROfusion codes (cR2S European SDDR tool). In this article the CAD2CDGS tool methodology, inputs and workflow are explained. Two verification tests have been done to check the correctness. Finally, it is demonstrated the applicability of the tool to IFMIF-DONES, with the radiological zoning analysis of the rooms surrounding the Lithium Loop Cell impact by the Be-7 and ACP. [ABSTRACT FROM AUTHOR]

Abstract The In-Vessel Viewing System (IVVS) is a diagnostic tool in ITER to perform in-vessel inspections between plasma pulses or during shutdown. During plasma pulses, the IVVS is parked inside a dedicated port extending from the vacuum vessel (VV) to the port cell (PC). The IVVS design must fulfill specific requirements related to the service lifetime of sensitive parts and to the radiation effects on neighbouring components and the PC. Analyses were performed to determine nuclear responses in the IVVS and neighbor components at plasma operation and after shutdown. The neutron flux map shows improvements due to improved shielding in the neighbouring lower ports and gives insight to further possibilities of reducing nuclear loads to IVVS sensitive components. The activation analysis reveals that the major contributors to the decay photon source are the movable shield block counterweights. [ABSTRACT FROM AUTHOR]

• A new feature of on-the-fly global variance reduction (OTF) to accelerate the MC global coupled neutron photon transportation has been developed. • ITER C model and IFMIF-DONES (International Fusion Materials Irradiation Facility- DEMO Oriented NEutron Source) had been chosen as test cases and calculated with OTF and analogue. • The results of OTF had been compared to analogue • OTF has shown to result in substantial reduction of computational time for global coupled neutron photon shielding calculation. Using Monte Carlo (MC) transport codes for fusion device shielding calculation is very challenging due to the complexity and heavy shielding of a fusion reactor. An effective on-the-fly (OTF) global variance reduction method was developed and verified before. The OTF method is designed to speed up the general-purpose MC global total neutron distribution by generating and updating the global weight window mesh (WWM) internally in the MC code. In this work, this method has been further extended for accelerating global coupled neutron photon transportation calculation. The goal of this paper is to evaluate the applicability of OTF to enhance the efficiency of the global total neutron and photon fluxes calculation of challenging problems. For verification purposes, the ITER C-model and IFMIF-DONES have been chosen as test cases. The OTF has achieved a substantial reduction in computational time comparing with analogue method. [ABSTRACT FROM AUTHOR]

• The OpenMC Weight Window Mesh function has been developed and benchmarked. • The OpenMC WWM supports any WWM file produced for MCNP code. • Results show the ability of OpenMC WWM function for shielding calculations. OpenMC is a community-driven open-source Monte Carlo neutron and photon transport simulation code. It supports geometry modeling using constructive solid geometry, and is capable of performing fixed source particle transport simulation based on continuous-energy and group-wise nuclear cross-section data. The lack of variance reduction technique is one of the main drawbacks of this code, which prevent its use for deep penetration shielding calculations. In this work, a commonly used variance reduction technique – using the Weight Window Mesh (WWM) function has been developed and benchmarked for OpenMC. This enables OpenMC to use any WWM file produced for the MCNP code. For benchmarking, two deep penetration shielding cases, a sphere simulation model and the thickest shielding cases in the TIARA high energy neutron shielding experiment has been employed, using the WWM produced by the ADVANTG code. In addition, the OpenMC WWM function has also been applied on the IFMIF-DONES Test Cell global neutron flux distribution calculations. The successful benchmarking and application demonstrate ability of the newly developed OpenMC WWM function for applications in general purposes deep penetration shielding calculations. [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]

IFMIF-DONES (International Fusion Materials Irradiation Facility – DEMO Oriented Neutron Source) is a fusion materials testing facility that is currently being designed under the framework of a work package of the EUROfusion Consortium. It will use a 125 mA at 40 MeV deuteron beam to generate a high neutron flux through L i (d , x n) stripping nuclear reactions in a liquid lithium target. The High Energy Beam Transport line (HEBT), the most upstream system of the IFMIF-DONES accelerator, is responsible for the guidance and shaping of the beam towards the target. Additionally, during commissioning periods, the HEBT is also responsible for diverting the beam, through the Beam Dump Transport Line, to the Beam Dump for testing purposes. The HEBT is spread along different rooms and zones: the Accelerator Vault, the Radiation Interface Zone (RIZ), and Target Interface Room (TIR). The engineering design of the HEBT components situated within the TIR and RIZ has been updated to satisfy new requirements, with a focus on ensuring the protecting of the Fast Isolation Valve (FIV) from the backscattered radiation from the target. These modifications include relocating the FIV from the TIR to the RIZ, adjusting the building layout to accommodate the new FIV module, configuring an enclosure cabinet for the RIZ, and adding local shielding to extend the lifetime of the FIV seal actuators. This work describes the current status of these TIR and RIZ engineering design, including radioprotection, commissioning and maintenance plan, beam diagnostics devices, beam dynamics and new remote handling approaches, as well as the layout and integration of the required components along the beamline. The TIR and RIZ are critical areas for IFMIF-DONES , and their design and operation must be compliant with functional, reliability and safety requirements. The updated design addresses potential issues and enhances the facility's overall functionality. • Collimator redesigned with new coolant fluid for the near TS safety measures. • TIR layout shift to two LRU frees space for new elements. • TIR-RIZ wall thickness reduced by one meter for space. • Enclosure cabinet in RIZ stores argon atmosphere for the FIV. • FIV lifespan doubled by implementing local shielding strategies. [ABSTRACT FROM AUTHOR]

• We investigated a set of dosimetry reactions for measurement of the neutron fluence in the High-Flux Test Modules of DONES. • Foil stacks irradiated with a neutron source at NPI Rez providing white neutron spectrum covering reactions of interest. • All expected induced activities could be determined by gamma spectroscopy utilizing a HPGe detector. • Further work will require to adapt the set of dosimetry reactions to the very high fluences in the HFTM.. The High Flux Test Module (HFTM) carries the specimens to be irradiated with the Early Neutron Source. The knowledge of the total neutron fluence on the specimen will be necessary for analysis. Neutron activation is a widely used method for neutron fluence measurement. We have tested a set of activation materials which seem to be suitable for application in the HFTM considering half-lives of the induced activities, the environmental conditions and the deployment time in the HFTM. Test foils were irradiated for 9.5 h with a cyclotron of NPI Řež in a white neutron field with energies up to 32 MeV. The induced activities were determined 61 days after the irradiation. We present the resulting saturation activities with estimated uncertainties and draw conclusions on the application of the selected activation materials in the HFTM. [ABSTRACT FROM AUTHOR]

Abstract A novel approach has been developed for accurate and efficient shutdown dose rate (SDDR) calculations on fusion reactor devices. The new approach is based on an assumption of linear relationship between the mono-energetic neutron flux and the decay gamma source, thus the coefficients between them can be obtained for every materials and neutron energy groups through an inventory calculation. These coefficients are used in an extended version of MCNP code to perform decay gamma transport calculation. Decay gamma generated in multiple cooling time steps can be simulated together with the neutron transport calculation. This new approach has been successfully verified through the comparison with the JET shutdown dose rate measurements, R2S and D1S results. The results obtained demonstrate the reliability and efficiency of the new approach in estimation of the SDDR in fusion devices. [ABSTRACT FROM AUTHOR]