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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]

• 7Be has no radiological implications in the hydrogen trap room of future IFMIF-DONES during normal operation and maintenance operations. • In normal operation maximum ambient dose equivalent rates of 15.1 mSv/h at contact with the hydrogen trap and 0.31 mSv/h at 1 m has been assessed. • In maintenance operation maximum dose rates of 1.42 mSv/h has been estimated. • A concrete wall 30 cm thick located at 1 m of the hydrogen trap would reduce the dose rate at allowed radiation dose limits levels for workers. The radiological impact that 7Be and activated corrosion products (ACP) dissolved in Li could have in normal operation and maintenance in the surroundings of the hydrogen trap (HT) of future IFMIF-DONES facility has been assessed. Ambient dose equivalent rates were calculated with radiation transport program MCNP at half height of trap. Four different Y contents were considered in the trap: 1, 3, 5 and 8 kg. In normal operation dose rates in the range 13.7–15.1 mSv/h have been found in contact with the trap and 0.28–0.31 mSv/h at 1 m of the trap. 7Be has a very low contribution to dose (3–3.7 µSv/h). Dose simulations have shown that a concrete wall 30 cm thick can reduce dose rates to acceptable dose limits. In maintenance operation the trap is drained and Li film remains attached to Y pebbles. Three possible Li film thickness have been considered: 10, 50 and 100 µm. Estimated 7Be activity is below exemption limits during maintenance. Assessed dose rates are in the range 34–1420 µSv/h at contact with the trap and a maximum of 10.9 µSv/h at 1 m of the trap. [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]

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]

McCad is a geometry conversion tool developed at the Karlsruhe Institute of Technology (KIT) for the automatic conversion of CAD models into the semi-algebraic geometry representations which can then be used in Monte Carlo (MC) particle transport simulations applied in design analyses of fusion and fission reactors. New algorithms for the decomposition of complex geometry models have been designed and implemented in the McCad code. With these improvements, a weakness of the original McCad was mitigated which is due to the instability of the applied graphic kernel and the original decomposition algorithm. The improvements have been verified with some representative test models and a generic model of a DEMO fusion power reactor. The results show that the advanced McCad version with the new decomposition algorithms is more robust and provides more accurate and less complex conversion results. The decomposition process is efficient and produces correct results which are consistent with the original model. [ABSTRACT FROM AUTHOR]

The quench tank (QT) location in the test cell is an open issue for IFMIF-DONES (International Fusion Material Irradiation Facility- DEMO Oriented NEutron Source) design. Neutronics assessments have been carried out on two QT location options. A parametric study of the neutron streaming in the lithium chute shows a quasi-linear dependence of the neutron flux on the void thickness. For both options, activation calculations at key locations of the QT system indicate that the contact dose rate is higher than the hands-on dose limit over the whole maintenance period. In addition, the shutdown dose rates in the lithium facility room after 1-day shutdown exceed the hands-on dose limit in different levels. Therefore, remote handling is required for the maintenances of the QT in both considered locations. [ABSTRACT FROM AUTHOR]

Coupled multi-physics analyses on fusion reactor devices require high-fidelity neutronic models, and flexible, accurate data exchanging between various calculation codes. An integrated coupling approach has been developed to enable the conversion of CAD, mesh, or hybrid geometries for Monte Carlo (MC) codes MCNP5/6, TRIPOLI-4, and translation of nuclear heating data for CFD codes Fluent, CFX and structural mechanical software ANSYS Workbench. The coupling approach has been implemented based on SALOME platform with CAD modeling, mesh generation and data visualization capabilities. A novel meshing approach has been developed for generating suitable meshes for MC geometry descriptions. The coupling approach has been concluded to be reliable and efficient after verification calculations of several application cases. [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]

• 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]

• A variance reduction approach for shutdown dose rate calculations has been proposed. • Verifications of a new shutdown dose simulation package has been carried out by comprehensive comparisons. • The shutdown dose rate in the IFMIF-DONES lithium loop cell has been assessed based on the updated design layout. Shutdown dose rate (SDR) calculations for the lithium loop cell (LLC) of the IFMIF-DONES (International Fusion Material Irradiation Facility- DEMO Oriented NEutron Source) have been performed in this work. A new version of Rigorous 2-Step code R2Smesh-3.0 was utilized, which shows improvements on efficiency, robustness, and reliability in SDR simulations. The quality of the dose map in the LLC has been improved using a weight-window mesh, combining with an intensity-based source sampling technique. The dose rate in most of the area of the LLC is below 0.5 μSv/h at one day after shutdown. Strong gamma streaming through the empty lithium pipe is a possible concern for the lithium loop maintenance. [ABSTRACT FROM AUTHOR]

• A novel GVR for MC shielding calculation which is convenient and highly efficient is proposed and developed in this paper. • A method to deal with long history problem is proposed in this paper. • Using simple cuboid and ITER model to test the performance of on-the-fly GVR method. This paper presents an improved on-the-fly global variance reduction (GVR) method for Monte Carlo radiation shielding calculations. This method generates the weight window (WW) mesh internally in the Monte Carlo (MC) code, and updates automatically the WW value on-the-fly as the particle flux of a mesh element has a certain level of statistical error. This method has many advantages, such as, it does not depend on third party codes; it only needs one MC run to obtained the WW mesh together with the global flux map; and it has no limitation on the source definition and particle types. In addition, a method to deal with long history problem, which is caused by excessive splitting of tracks, is proposed in this paper. The on-the-fly GVR method has been implemented in the MCNP code and verified using a simple 2D model and the ITER benchmark model. The results show significant acceleration effects on the MC shielding calculation against analog simulations. [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]

• 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]

Abstract: As the main interface components between plasma and vacuum vessel, the divertor is foreseen to be removed to the hot cell for refurbishment during the 20 years of ITER operation. During this process, the activated divertor will cause a large increase of radiation in the ITER building. 3D analysis has been performed to assess the radiation maps throughout the ITER building for assisting the shielding design for personnel and sensitive equipment. The activation of the divertor has been determined by coupled neutron transport and inventory calculations, radiation maps have been obtained from gamma transport calculations. The neutron and gamma transport calculations have been performed by MCNP5 code with FENDL2.1library. The inventory calculations have been performed by FISPACT2007 code with EAF-2007 library. The results of these 3D decay gamma radiation maps are presented by pictures in this paper, including the biological dose maps and maps of heat deposition in electronic equipment. [Copyright &y& Elsevier]

• 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 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]

• 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]

Helium flows at low pressure (3 bar) are used to cool the specimen capsules and the structure of the neutron irradiated High Flux Test Module (HFTM) of the DEMO-Oriented Neutron Source (DONES). The flow path includes inlet and outlet ducts with large cross sections, but also mini-channels with gap widths less than 1 mm, where a high velocity low Reynolds number flow influences the temperature of the irradiated specimens. The aim of the study was the achievement of thermal requirements to the HFTM. The large span of Reynolds numbers from laminar to fully turbulent is a significant challenge for the simulation of the complete HFTM. Several turbulence models were tested using experimental results obtained in the ITHEX (IFMIF Thermal-Hydraulic Experiment) experimental facility. Reynolds Stress (RSM) and k-ω Shear Stress Transport (SST) models are able to reproduce the heat transfer within the Reynolds number range between 4500 and 10,000. Simulations show that in case of 100% nuclear heating conditions the prescribed temperature of specimen can be achieved by justification of electrical power and variation of helium mass flow rate for each HFTM compartment. [ABSTRACT FROM AUTHOR]

A maintainable test cell (TC) concept, which targets replacing TC key components in case of damages, for the IFMIF-DONES (International Fusion Materials Irradiation Facility – Demo Oriented NEutron Source) facility is proposed to facilitate the maintenance of key containment and biological shielding components, i.e. the liner and the surrounding concrete walls. Unlike the current TC reference design which features permanent liner and concrete shielding walls, the new proposal decouples the components which suffer relative high failure possibilities from the ones which can stay as permanent construction structures. The former ones, that are expected to be replaced in case of damage, include removable sections of the liner and part of the biological shielding materials. Fundamental mechanical solution of this configuration is briefly described, and technical feasibilities are preliminarily analyzed for the aspects of independent liner enclosure structure, maintenance scenarios using remote handling systems, etc. [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]

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]

The ITER In-Vessel Viewing System (IVVS) consists of six identical units located at the B1 level of the Tokamak complex. In this work relevant nuclear quantities in the Port Cell (PC) have been calculated by means of the MCNP Monte Carlo code in a full 3-D geometry, including the IVVS and its shielding blocks. A comprehensive MCNP model of the PC has been developed including a detailed description of the Bioshield plug, pipes, penetrations, cask rails and PC door. The neutron and gamma sources needed to perform the nuclear analyses include both contributions from the radiation streaming through the Lower Port and activated water. Monte Carlo calculations have been performed to assess the radiation field inside the PC through maps of neutron and gamma fluxes. Absorbed dose on sensitive components over the ITER lifetime has been calculated in the PC area, in order to estimate the nuclear loads that the installed equipment has to withstand. Moreover, the shutdown dose rate distribution at 1 day after shutdown has been calculated. [ABSTRACT FROM AUTHOR]

The test cell (TC) in IFMIF-DONES (International Fusion Material Irradiation Facility – Demo Oriented NEutron Source) is the meeting area of the three major systems (test systems, lithium systems, and accelerator systems) to host the irradiation test module, lithium target assembly and the end section of the accelerator line. Although the design of the TC, to a large extent, inherits that of the reference IFMIF-EVEDA (Engineering Validation and Engineering Design Activities) TC, design justifications are required to solve the existing major open issues such as cavitation of the lithium flow, activation and maintainability of key components, and etc. In this paper, these major open issues of the reference TC layout are addressed and a TC reconfiguration proposal, featuring arranging the lithium quench tank inside the TC, is introduced. The two concepts are investigated, analyzed, and compared on the aspects of lithium flow stability, tritium generation rate, key component maintenance scenarios, irradiation shielding, and in-TC components arrangement. [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]

Abstract: The Experimental Advanced Superconducting Tokamak (EAST) device began operation in 2006. A radiation protection work has been paid more and more attention with discharge pulse length increase. A system named radiation virtual simulation for dosimetry (RVIS) using virtual reality (VR) technology has been developed to improve the efficiency of maintenance work, and decrease the radiation exposure of maintenance personnel. The RVIS can be used to evaluate workload in radiation environments, optimize the maintenance plan and visualize three-dimensional (3D) dose rate distribution. The original 3D dose rate distribution in shielded hall was calculated rapidly to use as a basis for dose evaluation and 3D dose field visualization. The RVIS improves on the original work to give more effective dose evaluation and dynamic visualization of the dose field during EAST experiment campaigns. [ABSTRACT FROM AUTHOR]

Abstract: A multigroup library HENDL2.1/SS (Hybrid Evaluated Nuclear Data Library/Self-Shielding) based on ENDF/B-VII.0 evaluate data has been generated using Bondarenko and flux calculator method for the correction of self-shielding effect of neutronics analyses. To validate the reliability of the multigroup library HENDL2.1/SS, transport calculations for fusion–fission hybrid system FDS-I were performed in this paper. It was verified that the calculations with the HENDL2.1/SS gave almost the same results with MCNP calculations and were better than calculations with the HENDL2.0/MG which is another multigroup library without self-shielding correction. The test results also showed that neglecting resonance self-shielding caused underestimation of the K eff, neutron fluxes and waste transmutation ratios in the multigroup calculations of FDS-I. [ABSTRACT FROM AUTHOR]

Abstract: SNAM is the abbreviation of “SN Auto-Modeling”, which is a CAD interface program for SN codes based on CAD technology. Making full use of CAD technology is SNAM''s major characteristic. In brief, CAD technology is utilized in the following two ways: firstly, SNAM makes it possible to create SN geometry models in convenient CAD environments; secondly, the creation of SN geometry models is remarkably accelerated by inheriting existing CAD models. Main capabilities of SNAM include: (1) geometry modeling; (2) ability to convert CAD models into SN geometry models; (3) ability to convert SN geometry models into CAD models. This paper describes the system architecture and main features of SNAM. SNAM-based SN modeling process and preliminary applications for real-world engineering models, such as ITER and EAST, are also presented. [Copyright &y& Elsevier]