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In the IFMIF-DONES facility of the future, the back-plate behind the Li target will receive strong irradiation from high-energy neutrons. The potential use of the back-plate for material specimens is attractive with respect to providing complementary irradiation data for Eurofer. In this work, DPA (displacement per atom) and gas production rates as well as DPA gradients and temperature distributions have been studied for the center segment of the back-plate, using both a nominal beam and a reduced beam footprint. It is shown that specimens can be produced with high DPA in similar conditions to the DEMO first-wall. Based on the size of the SSTT (small specimen test technology) specimens, the limited number of samples obtainable from the adopted arrangement scheme is driven by a major constraint: the thickness of the back-plate. A parametric study of the back-plate's thickness provides an alternative arrangement scheme; thus, the DPA and gradient of the specimens are remarkably improved. [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]

• 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 NASCA-OTF method has been developed for SDDR analysis of mobile activated components in fusion reactor. • It supports on the fly modification of geometry structure and global variance reduction during the MC run, and generation of statistically convergent SDDR maps when activated components stay at different positions. • The NASCA-OTF method has been verified through the calculation on self-designed model and CFETR model. During the plasma operation period of fusion reactor facilities, neutrons generated by DT reaction can penetrate deeply into structures and induce strong activation. Highly activated components have to be transferred from the on-load position to the hot cell for the purpose of maintenance. This movement produces a high-dose field, originating from the mobile radiation source, which could affect the functionality of important devices or pose risks to the workers. Therefore, it is imperative to compute the shutdown dose rate (SDDR) maps while transporting activated components in the facility and develop an optimized solution for this calculation process. In this study, we established a new extension of the NASCA method combined with the OTF GVR method to compute SDDR maps. NASCA-OTF is a method developed based on the shutdown dose simulation code NASCA to ensure automatic modification of the geometry structure on the fly. In the meantime, it uses OTF method for speeding up the Calculation. Thus, this modified method produces convergent SDDR maps corresponding to different positions of activated components within a single Monte Carlo (MC) simulation. The NASCA-OTF was verified by comparison with R2S method through benchmark analysis. The method was also used for SDDR analysis of the mobile blanket sector for the China Fusion Engineering Test Reactor (CFETR), and the results indicated that the NASCA-OTF is feasible for computing the SDDR resulting from the movement of activated components. [ABSTRACT FROM AUTHOR]

• An automatic Global Variance Reduction has been developed & implemented in OpenMC. • The GVR enhances computation efficiency than flux-based GVR for point-like source. • Results show the ability of global shielding calculations with OpenMC only. Shielding calculation is an important but challenging part of safety analyses for nuclear facilities due to the complex geometry and heavy shielding. In this work, an automatic Global Variance Reduction (GVR) method has been developed and implemented in OpenMC, an open-source Monte Carlo (MC) code for neutron and photon transport simulation. This GVR method is based on flux-based GVR method, improving it by defining a novel "pseudo-source", which enhances remarkably computation efficiency. In addition, the OpenMC parallel calculation is also improved with the asynchronous scheduling function to mitigate the unbalance loading on the processes. For benchmarking this automatic GVR method, a simple thick solid sphere benchmark and a heavy shielding cuboid benchmark has been employed. An application on IFMIF-DONES Test Cell has been also adopted to demonstrate the ability of this GVR method on shielding calculations of heavy-shielded facility. These benchmarks and application show the enhancement on computation efficiency, as well as some limitations, of this newly developed GVR method. [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]