F. Moro

Combined unfolding and spatial inversion of neutron camera measurements for ion temperature profile determination in ITER

Measurements of the core ion temperature profile are required in ITER with ?10% accuracy and 100?ms time resolution. The ITER radial neutron camera (RNC), with 45 collimated lines of sight (spanning almost completely a poloidal plasma section) equipped with compact neutron spectrometers, has the potential to provide such spatially resolved temperature information. In this paper a novel technique for the ion temperature profile measurement based on the combination of unfolding and spatial inversion of RNC measurements is presented and applied to the ITER full power standard H-mode deuterium?tritium scenario, assuming the RNC to be equipped with liquid scintillators. Results based on synthetic data suggest that the target accuracy could be reached with a time resolution ?200?ms.

Neutron measurements in ITER using the Radial Neutron Camera

The Radial Neutron Camera (RNC) is one of the key diagnostic systems of the ITER international fusion experiment. It is designed to measure the uncollided 14 MeV and 2.5 MeV neutrons from deuterium-tritium (DT) and deuterium-deuterium (DD) fusion reactions taking place in the ITER plasma through an array of 45 detectors positioned along collimated lines of sight. Scintillators and diamonds coupled to fast digital acquisition electronics are among the detectors presently considered for the RNC. The RNC will provide spatially resolved measurements of several plasma parameters needed for fusion power estimation, plasma control and plasma physics studies. The line-integrated RNC neutron fluxes are used to evaluate the local profile of the neutron emission (neutron emissivity, s−1m−3) and therefore the total neutron yield and the birth profile of the alpha particles. The temperature profile of the bulk ions can be derived from the Doppler broadened widths of the RNC line-integrated spectra, that also provide insight on the supra-thermal ions produced by the injection in the plasma of electromagnetic waves and neutral particles. The RNC emissivity and temperature measurements can be employed to estimate the composition of the ITER fuel, namely the ratio between the tritium and deuterium densities. Data processing techniques involving spatial inversion and spectra unfolding are necessary to deduce the profile quantities from the line-integrated RNC measurements. The expected performances of the RNC as a diagnostic for the neutron emissivity/ion temperature/fuel ratio profile (measurement range, time resolution, accuracy, precision) have been estimated by means of synthetic data simulating actual RNC measurements. The results of the simulations, together with an overall description of the diagnostic and of the measurement techniques, are presented.

Neutronic analysis of FAST

As part of the Fusion Advanced Studies Torus (FAST) project, a neutronic analysis has been performed, aimed to design optimization and radiological safety assessment. The neutron emissivity source foreseen for various FAST scenarios has been calculated and used as input for Monte Carlo calculations. The shielding analysis and nuclear heating calculations have been carried out with MCNP5 using a detailed 3-D model of the machine. The energy and spatial distributions of neutron fluxes have been used to perform activation analysis by means of the FISPACT code for safety assessment. The implications of the results on the design of the machine and on safety issues are presented and discussed.

Progress on the Integration of ITER Diagnostics Equatorial Port Plugs in Europe

Diagnostics in ITER are supported by big structures called port plugs, the second main function of which is to ensure a sufficient shielding against neutrons and gammas. Regarding the integration of diagnostics in equatorial port plugs, a new approach is under study, which consists in installing the diagnostics in “drawers”. This paper describes the recent work which has been performed in Europe on the integration of diagnostics in drawers in the Equatorial Port Plug 1 (EPP1). First the methodology which has been followed to progress on the integration of the diagnostics in this port plug is described and the resulting arrangement of diagnostics is shown. Then a special attention is paid to the integration of the two main diagnostics of EPP1, namely the visible/infrared wide angle viewing system and the radial neutron camera. Finally the preliminary design of the drawers of EPP1, in particular the shielding modules around the diagnostics, is presented, and the preliminary results of the analyses performed to validate this design are provided.

Progress on the integration of ITER diagnostics equatorial port plugs in Europe

Diagnostics in International Thermonuclear Experimental Reactor (ITER) are supported by big structures called port plugs, the second main function of which is to ensure a sufficient shielding against neutrons and gammas. Regarding the integration of diagnostics in equatorial port plugs, a new approach is under study, which consists in installing the diagnostics in “drawers”. This paper describes the recent work which has been performed in Europe on the integration of diagnostics in drawers in the Equatorial Port Plug 1 (EPP1). First, the methodology which has been followed to progress on the integration of the diagnostics in this port plug is described, and the resulting arrangement of diagnostics is shown. Then, a special attention is paid to the integration of the two main diagnostics of EPP1, namely the visible/infrared wide angle viewing system and the radial neutron camera. Finally, the preliminary design of the drawers of EPP1, in particular the shielding modules around the diagnostics, is presented, and the preliminary results of the analyses performed to validate this design are provided.

The McCad Code for the Automatic Generation of MCNP 3-D Models: Applications in Fusion Neutronics

The Monte Carlo (MC) code MCNP is the reference tool in fusion neutronics, allowing the description and analysis of full and detailed 3-D geometry of a tokamak machine. The geometrical models of the components used are typically available through computer aided design (CAD) files: the main benefits of this system are related to its portability and compatibility with several tools commonly used in engineering analyses. However, at the present stage, the information contained in CAD files cannot be directly provided to MC as inputs, because of the different representation scheme used. This issue leads to the necessity to develop interfaces that can translate them into the correct MC geometrical description. McCad is a software developed by the Karlsruhe Institute of Technology, dedicated to the fully automated generation of the MCNP geometrical models from CAD files (step, iges, and brep formats): it is provided with a graphical user interface allowing the visualization of the geometries and tools for data exchange and modeling. This paper summarizes the results of some benchmark tests performed on JET components and a DEMO reactor aimed at the assessment of the suitability of McCad for fusion neutronic applications. The reliability of the conversion algorithm has been evaluated comparing the results of stochastic MCNP volume calculations carried out using the generated models, and the corresponding volumes provided by the CAD kernel of the interface program. Moreover, the consistency of a converted DEMO MCNP model has been verified through particle transport calculations for the estimation of the neutron wall loading poloidal distribution. Several aspects related to the use of the code have been evaluated such as its portability, performances, and impact of the geometric approximation introduced on the neutronic analyses. Furthermore, a useful feedback for the optimization and enhancement of the McCad interface has been provided.

Applications of McCad for the automatic generation of MCNP 3D models in fusion neutronics

The Monte Carlo (MC) code MCNP is the reference tool in fusion neutronics, allowing the description and analysis of full and detailed 3D geometry of a tokamak machine. The geometrical models of the components used are typically available through computer aided design (CAD) files: the main benefits of this system are related to its portability and compatibility with several tools commonly used in engineering analyses. However, at the present stage, the information contained in CAD files cannot be directly provided to MC as inputs, because of the different representation scheme used. This issue leads to the necessity to develop interfaces that can translate them into the correct MC geometrical description. McCad is a software developed by the Karlsruhe Institute of Technology (KIT), dedicated to the fully automated generation of MCNP geometrical models from CAD files (STEP, IGES and BREP formats): its provided with a graphical user interface (GUI) allowing the visualization of the geometries and tools for data exchange and modelling. The present paper summarizes the results of some benchmark tests performed on JET components and a DEMO reactor aimed at the assessment of the suitability of McCad for fusion neutronic applications. The reliability of the conversion algorithm has been evaluated comparing the results of stochastic MCNP volume calculations carried out using the generated models, and the corresponding volumes provided by the CAD kernel of the interface program. Moreover, the consistency of a converted DEMO MCNP model has been verified through particle transport calculations for the estimation of the neutron wall loading poloidal distribution. Several aspects related to the use of the code have been evaluated such as its portability, performances and the impact of the geometric approximation introduced on the neutronic analyses. Furthermore a useful feedback for the optimization and enhancement of the McCad interface has been provided.

A Neutronics Shielding Mock-Up Experiment for Reduction of Uncertainty on the Prediction of the ITER-TFC Nuclear Heating

Abstract A mock-up of the inboard shield and vacuum vessel of ITER was set up at ENEA Frascati and irradiated with 14-MeV neutrons produced by the Frascati Neutron Generator. The mock-up includes the coil region, and its dimensions and materials composition are consistent with the latest ITER design. The objective of the experiment is to validate the calculations of nuclear heating in the ITER toroidal field coil performed with the Monte Carlo code MCNP-5 and the FENDL-2.1 library, through measurement performed by thermoluminescent dosimeters. The goal was to reach a calculated-to-experiment accuracy of less than or equal to ±10%, possibly as low as ±8%.