Neutral Beam Duct Research Articles

A study on irradiated welding joints in ITER materials and its effects on FMECA analysis of components

Abstract ITER (International Thermonuclear Experimental Reactor) is a huge engineering challenge. ITER Project is in construction phase, the machine is very complex and is an assembly of hundreds of different components. Due to its large assembly and huge welding of different materials, it needs to have certain kind of reliability for safe operation. The reliability of machine is achieved in various lifecycle stages from design to manufacturing through ITER RAMI programme. Reliability and failure rate of weld influence availability of machine [1]. In this paper a study is done for the effects of irradiation on weldments and its mechanical properties and understanding various parameters which can affect the reliability of weld due to irradiation. Various papers were studied to understand changes in mechanical properties in weldment due to irradiation. Thus, by understanding various irradiation scenarios of ITER along with FMECA(Failure mode effect and criticality analysis) as tool, this work tries to systematically recognize, evaluate and prioritize the risk and potential failures. FMECA is a part of ITER RAMI programme. Failure rate data is an important input for FMECA analysis. In this study two weld seams, seam 1 and seam 2 were selected respectively in cryostat and neutral beam duct to understand irradiation condition under ITER working scenario. Flux and DPA at their Cartesian coordinates were found out with the help of various softwares for ITER project. FMECA chart is drawn to understand the severity and criticality under different irradiation scenario for these welds. Thus correlating mechanical properties of irradiated material to reliability, with FMECA as a tool will help in assessing availability of machine and lifecycle management of components. This will help in better understanding on failure rate data of non-irradiated conditions to be used for irradiated condition and changing neutron load in fusion reactors.

Integration of the Neutral Beam Injector System Into the DCLL Breeding Blanket for the EU DEMO

The integration of plant systems involving penetrations into the in-vessel components, such as heating and current drive, fuel cycle, and diagnostics, is a complex task constrained by top-level requirements of remote maintainability and high reliability. Within the EUROfusion Power Plant Physics and Technology Program, some activities are ongoing to assess the integration of different systems into the breeding blanket (BB), specifically neutral beam injector (NBI), electron cyclotron resonance heating launchers, diagnostics sightlines, fueling lines, and specific protections for the first wall (like startup limiters). This paper describes the integration of the NBI system into the Dual-Coolant Lithium-Lead (DCLL) BB for the EU DEMO. After identifying the major issues impacting the mechanical, thermal-hydraulic and neutronic behavior of the blanket, the integration efforts have been focused on minimizing the invasiveness of the NBI system and exploring different NBI options for the best compromise between plasma heating and BB performance. This paper describes the adaptation of the DCLL BB design to allocate the neutral beam duct. A particular attention is devoted to the redistribution of breeding and shielding functions, the new path of fluid circuits, and the additional cooling needs. The consequences of design modifications on key neutronic aspects like tritium breeding ratio and shielding capability are addressed. Besides, the thermal loads transferred to the BB walls from the neutral beam and the plasma are discussed, and a preliminary thermal assessment of the proposed solution is presented.

High heat flux test on the thermocouple embedded ITER neutral beam duct liner mock-up

The ITER neutral beam duct liner is located within the tokamak VV port extension and is mounted on the VV port extension flange. The duct liner is made from CuCrZr copper alloy and is actively cooled through deep-drilled channels. A number of thermocouples should be installed on the neutral beam duct liner in order to provide the ability to detect temperature excursions on the surface of the duct liner. Twenty thermocouples have been installed on the neutral beam duct liner full scale mock-up, and a high heat flux test has been performed at the KoHLT-EB test facility, in order to simulate temperature detection in the neutral beam duct liner during ITER operation. For each thermocouple, the fixation method has been verified by high heat flux test with uniform electron beam profile, and the temperature behavior of the neutral beam duct liner has been successfully simulated by Gaussian electron beam profile.

Multi-purpose deployer for ITER in-vessel maintenance

The multi-purpose deployer (MPD) is a general purpose in-vessel remote handling (RH) system in the ITER RH system. The MPD provides the means for deployment and handling of in-vessel tools or components inside the vacuum vessel (VV) for dust and tritium inventory control, in-service inspection, leak localization, and in-vessel diagnostics. It also supports the operation of blanket first wall maintenance and neutral beam duct liner module maintenance operations. This paper describes the concept design of the MPD. The MPD is a cask based system, i.e. it stays in the hot cell building during the machine operation, and is deployed to the VV using the cask system for the in-vessel operations. The main part of the MPD is the articulated transporter which provides transportation and positioning of the in-vessel tools or components. The articulated transporter has nine degrees of freedom with a payload capacity up to 2 tons. The articulated transporter can cover the whole internal surface of the VV by switching between the four equatorial RH ports. Additionally it can use two non-RH equatorial ports to transfer large tools or components. A concept for in-cask tool exchange is developed which minimizes the cask transportation by allowing the MPD to stay in the VV during the tool exchange.

Fabrication study on the cooling module of the ITER neutral beam duct liner

Recently the new concept of the ITER neutral beam (NB) duct liner has been developed to improve thermo-mechanical performance and satisfy the requirements for remote handling and maintenance. The design concept of cooling module located inside neutron shield structure is to use deep-drilled panels instead of the original design concept of the casting-modularized component with tubes. In this study, the manufacturing feasibility has been investigated through the fabrication of small size coupons and full scale mock-up. Firstly, the small size coupons are for developing the electron beam welding processes. Secondly, the full scale mock-up which has 6 holes for cooling passage has been fabricated in order to develop the main fabrication processes such as deep drilling, bending and machining. In addition, the pressure and the leak tests have been carried out to check the required performance for completed cooling panel. Although some improvement is required, but the Electron Beam Welding (EBW) has been successfully achieved and generally the deep drilling and bending process also shown good results in dimensional control.

Feasibility of a fast optical pressure interlock for the ITER neutral beam injectors

The feasibility of using Balmer- α emission for a high-speed pressure diagnostic and beam interlock for the ITER neutral beam heating system is investigated. An interlock is needed to prevent excessive re-ionisation of the neutral beam when rapid excursions of pressure occur in either the electrostatic residual ion dump (ERID), or the neutral beam duct (NBD). The re-ionised fraction of the beam, will be deflected by stray tokamak fields, potentially causing excessive thermal loads on beam line components. Experience from JET indicates that a response time of order 100 μ s is required in order to prevent fast pressure excursions. Fast penning gauges have a time response of around 30–50 ms, however, a faster response (around 1 μ s) is possible by monitoring the H α (656.3 nm)/D α (656.1 nm) emission from collisional excitation of the background gas and neutral beam. Published total cross-sections are used to calculate a signal of 3.5 × 1 0 13 – 3.0 × 1 0 17 photons s −1 m −2 sr −1 for normal conditions. This signal must be distinguished from the background light of the tokamak plasma (line emission and bremsstrahlung). The beam emission is Doppler shifted by up to 21 nm (D operation) and up to 27 nm (H operation) depending on angle of observation and this can be used to help distinguish against background line emission. The distribution of background light along the beam line is calculated with a two-dimensional radiosity code, solving the equilibrium energy balance within the beam line enclosure. The Balmer- α signal and background signal due to bremsstrahlung are compared for a 500-MW reference plasma.

ITER neutral beam system

Since the main features of the design of the neutral beam (NB) system for the International Thermonuclear Experimental Reactor (ITER) were first reported, integration with the tokamak and with the rest of the plant has been the main priority. Moreover, operational requirements and maintainability have been considered in the evolution of the design. Each of the three NB injectors is connected to the tokamak vacuum vessel with the NB duct on an equatorial port. The article describes the integration of the NB port/duct with the blanket, the vacuum vessel, the toroidal field and poloidal field coils, the cryostat and the bioshield. Two main design modifications are reported. The insulation of the source, originally done with compressed gas, is now achieved with vacuum to limit the power losses caused by the radiation induced conductivity. Large cylindrical insulators are still required but their inner diameter has been reduced from 2.7 to 1.8 m. The improvements on the compensation system needed to reduce the magnetic field in the NB volume are also described. Finally, the progress in R&D for the ITER NB system is reported, including an overview of the achievements in the critical areas of negative ion production at high current density (tests of a large size, low pressure, steady state caesiated ion source), acceleration up to 1 MV (tests of two alternative accelerator concepts) and neutralization (tests of an experimental plasma neutralizer to investigate it as an alternative to the gas target neutralizer).

Mars Axicell Radiation Damage and Shielding Analysis

Radiation effects and shield requirements are analyzed for the MARS axicell region. Dominant coil life limiting effects in the normal insert coils are found to be swelling in the spinel insulation and resistivity changes of the conductor due to transmutations. The shield between the insert and superconducting coils is optimized and reduces all radiation responses to acceptable levels. However, streaming in the neutral beam duct resulted in an unacceptable dpa rate in the Cu stabilizer and has resulted in a redesign of this region without a neutral beam and with a single insert coil.

Nuclear Analysis of a Tokamak Experimental Power Reactor Conceptual Design

Detailed nuclear analysis of a reference conceptual design for a tokamak experimental power reactor (EPR) is presented. The reference EPR has a 6.25-m major radius and a 2.1-m minor radius circular plasma with a nominal neutron wall loading of 0.5 MW/m/sup 2/. A 0.28-m-thick blanket of stainless steel surrounds a stainless-steel vacuum vessel. The inner shield consists of stainless steel and B/sub 4/C and is 0.58 m thick. The 0.97-m-thick outer shield employs lead mortar, stainless steel, and graphite. The neutronics results in the first wall and blanket vary significantly in the poloidal direction due to an outward shift in the deuterium-tritium neutron source distribution and the toroidal curvature. The infinite cylinder approximation overestimates response rates in the first wall compared with toroidal geometry calculations. Neutral beam lines, vacuum ducts, and other penetrations of the blanket and bulk shield represent large (approximately 0.6- to 1.0-m/sup 2/ cross section) streaming paths for neutrons and require special shielding. A special 0.75-m-thick annular shield surrounds the neutral beam duct after it exists from the bulk shield and extends beyond the toroidal field coil and out to the beam injectors. A pneumatically operated movable shield plug, opening during the pumpdown phase and closing duringmore » the plasma burn, is selected as the principal design option for shielding the evacuation ducts.« less