Search

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]

• 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]

• 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: 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]