Structural Design Criteria For In-vessel Components Research Articles

Lifetime assessment of the modified grounded grid in the negative ion source SPIDER

In a nuclear fusion reactor, some components of the system are subjected to huge thermal loads. These components, such as grids, neutralizers, calorimeters, ducts, divertors, blankets etc. have to be actively cooled in order to avoid damage, functionality problems and specially to ensure their reliability. The design process of such components requires taking into account several physical phenomena, with the goal of not only satisfying the functionality conditions, but also with a focus on the structural verifications and component lifetime. For assessment of the structural design for components in ITER, the ITER Structural Design Criteria for In-vessel Components (SDC-IC) represent the reference verification criteria.The aim of this work is to investigate the relationship between the cooling system performance and the fatigue life. In this work a generalized framework for the evaluation of cooling system components is presented, taking into account the necessary input parameters, performing numerical analyses, division of component into several interest regions and assessing the fatigue life. The performed numerical analyses are a set of coupled CFD and thermo-structural analyses.A case study is presented for a critical component, the Grounded Grid (GG), inside the SPIDER beam source experiment at Consorzio RFX, Padova, Italy. The presented test case includes cooling system effectiveness analysis, assessment of the temperature distribution on housing during operations, mapping the pressure distribution inside the cooling system and finally assessment of the maximum number of allowed cycles.

Analysis and experimental justification of electrical connector for ITER blanket module

In ITER blanket system, electrical connectors (“E–straps”, ES) are used to form a low impedance electrical path from shield blocks (SB) to the vacuum vessel (VV). Main functions of ES is providing current from SB to VV. ES shall withstand electromagnetic (EM) loads and Joule heating resulted from electrical current with magnitude up to 137kA during 300ms, accommodate cyclic relative displacements SB-to-VV and operate in harsh environment: neutron radiation, thermal cycling and high vacuum.Manufacturing of the ES from one solid blank by electrical discharge machining (EDM) has been developed and proposed by NIKIET instead of the assembling from separate parts. This paper presents design and analyses efforts applied on development of ES made of solid blank. Patterns of temperature, EM loads and stress-strain state, accumulated fatigue damages have been found with numerical simulations for all specified loading conditions. Strength assessment shows that current design of ES meets the applicable design code: structural design criteria for in-vessel components (SDC-IC). Results of cyclic mechanical tests at temperature are presented in this paper.

Two key improvements to enhance the thermo-mechanic performances of accelerator grids for neutral beam injectors

The MITICA experiment (Megavolt ITER Injector & Concept Advancement) is the prototype and the test bed of the Heating and Current Drive Neutral Beam Injectors, which will be necessary for the full-performance exploitation of ITER. MITICA injector experiments shall demonstrate the reliable and accurate emission of a 17 MW beam of neutral particles for duration up to 1 hour, fulfilling ITER specific requirements.The accelerator grids are among the most critical parts of this experiment, because they must fulfill several operational requirements and at the same time satisfy the fatigue verifications according to the ITER Structural Design Criteria for In-vessel Components (SDC-IC). After about two years of continuous development, two design improvements were found to effectively increase the fatigue life of the grids up to the requested values. The first method was to adopt a novel shape of the cooling channels inside the grids, called Nozzle Island Cooling Enhancement (NICE) and able to provide a high performance cooling without exceeding the limits on the pressure drop through the grids. The second, called Stress Relieving Slits (SRS), was to introduce suitable slits in the grids, whose design was iteratively optimized until they were able to significantly reduce the stress/strain peaks due to thermal gradients.The NICE and SRS design solutions, here described in detail, were found to be key improvements in order to obtain a design of the grids able to satisfy all the operating requirements and all the structural verifications according to the ITER criteria.

Comparative evaluation of structural integrity for ITER blanket shield block based on SDC-IC and ASME code

The ITER blanket Shield Block is a bulk structure to absorb radiation and to provide thermal shielding to vacuum vessel and external vessel components, therefore the most significant load for Shield Block is the thermal load. In the previous study, the thermo-mechanical analysis has been performed under the inductive operation as representative loading condition. And the fatigue evaluations were conducted to assure structural integrity for Shield Block according to Structural Design Criteria for In-vessel Components (SDC-IC) which provided by ITER Organization (IO) based on the code of RCC-MR.Generally, ASME code (especially, B&PV Sec. III) is widely applied for design of nuclear components, and is usually well known as more conservative than other specific codes. For the view point of the fatigue assessment, ASME code is very conservative compared with SDC-IC in terms of the reflected Ke factor, design fatigue curve and other factors. Therefore, an accurate fatigue assessment comparison is needed to measure of conservatism. The purpose of this study is to provide the fatigue usage comparison resulting from the specified operating conditions shall be evaluated for Shield Block based on both SDC-IC and ASME code, and to discuss the conservatism of the results.

Multiphysics Engineering Analysis for ITER Diagnostic First Wall and Shield Module Design*

ITER diagnostic first walls (DFWs) and diagnostic shield modules (DSMs) inside the port plugs (PPs) are designed to protect diagnostic instrument and components from a harsh plasma environment and provide structural support while allowing for diagnostic access to the plasma. The design of DFWs and DSMs are driven by 1) plasma radiation and nuclear heating during normal operation 2) electromagnetic loads during plasma events and associate component structural responses. A multi-physics engineering analysis protocol for the design has been established at Princeton Plasma Physics Laboratory and it was used for the design of ITER DFWs and DSMs. The analyses were performed to address challenging design issues based on resultant stresses and deflections of the DFW-DSM-PP assembly for the main load cases. ITER Structural Design Criteria for In-Vessel Components (SDC-IC) required for design by analysis and three major issues driving the mechanical design of ITER DFWs are discussed. The general guidelines for the DSM design have been established as a result of design parametric studies.

Multiphysics Engineering Analysis for an Integrated Design of ITER Diagnostic First Wall and Diagnostic Shield Module Design

ITER diagnostic first walls (DFWs) and diagnostic shield modules (DSMs) inside the port plugs (PPs) are designed to protect diagnostic instrument and components from a harsh plasma environment and provide structural support while allowing for diagnostic access to the plasma. The design of DFWs and DSMs are driven by 1) plasma radiation and nuclear heating during normal operation 2) electromagnetic loads during plasma events and associate component structural responses. A multi-physics engineering analysis protocol for the design has been established at Princeton Plasma Physics Laboratory and it was used for the design of ITER DFWs and DSMs. The analyses were performed to address challenging design issues based on resultant stresses and deflections of the DFW-DSM-PP assembly for the main load cases. ITER Structural Design Criteria for In-Vessel Components (SDC-IC) required for design by analysis and three major issues driving the mechanical design of ITER DFWs are discussed. The general guidelines for the DSM design have been established as a result of design parametric studies.

Analysis of the thermomechanical behavior of the IFMIF bayonet target assembly under design loading scenarios

In the framework of the IFMIF Engineering Validation and Engineering Design Activities (IFMIF/EVEDA) phase, ENEA is responsible for the design of the European concept of the IFMIF lithium target system which foresees the possibility to periodically replace only the most irradiated and thus critical component (i.e., the backplate) while continuing to operate the rest of the target for a longer period (the so-called bayonet backplate concept). In this work, the results of the steady state thermomechanical analysis of the IFMIF bayonet target assembly under two different design loading scenarios (a “hot” scenario and a “cold” scenario) are briefly reported highlighting the relevant indications obtained with respect to the fulfillment of the design requirements. In particular, the analyses have shown that in the hot scenario the temperatures reached in the target assembly are within the material acceptable limits while in the cold scenario transition below the ductile to brittle transition temperature (DBTT) cannot be excluded. Moreover, results indicate that the contact between backplate and high flux test module is avoided and that the overall structural integrity of the system is assured in both scenarios. However, stress linearization analysis reveals that ITER Structural Design Criteria for In-vessel Components (SDC-IC) design rules are not always met along the selected paths at backplate middle plane section in the hot scenario, thus suggesting the need of a revision of the backplate design or a change of the operating conditions.

100-kV feedthrough for the Indian Test Facility (INTF) — design and analysis

The Indian Test Facility (INTF) high-voltage (HV) feedthrough has been designed to connect all the required feedlines carried by the transmission line from the HV deck to the beam source. It forms the primary vacuum boundary and provides 100-kV isolation for INTF beam operation. Thermomechanical and electrostatic analyses have been carried out to validate the design. The observed thermomechanical stress and deflections are within acceptable limits. The results are validated using Structural Design Criteria for In-vessel Components (SDC-IC) criteria. The present manuscript describes the design of and the analysis results for the INTF HV feedthrough.

DNB exit scraper—Concept and engineering

The 100kV negative hydrogen ion source based Diagnostic Neutral Beam (DNB) injector is part of the Indian procurement for ITER. The DNB is designed to deliver an 18–20A hydrogen neutral beam to the ITER plasma. The exit scraper (ES) defines the profile of the beam after it leaves the calorimeter, the last Beam Line Components (BLCs) during operation. BTR11Developed by the Russian Research Centre, Kurchatov Institute, Moscow, Russia. and PDP1 codes are used to obtain the optimum entry and exit opening dimensions of the ES, so as to deliver the maximum power to the tokamak plasma, and to limit the power to downstream components due to beam interception, to protect front end components from the interception of re-ionized devious particles. Each horizontal heat transfer (HT) panel of ES receives a total power of 85kW for each symmetric 10mrad beam (worst case scenario) with peak power density of 1.17MW/m2. The thermo hydraulic design is carried out to withstand the heat flux due to the beam interception on the ES. The mechanical design of the ES is carried out by considering the spatial constraints in the DNB system and the remote handling (RH) system. In the present design, the copper heat transfer panels are directly bolted to the structural member. Protection shall be incorporated to ensure that there is no exposure of the bolt head. The thermo-mechanical analysis has been carried out for the normal, off-normal (accidental/worst-case) events of the ITER DNB. During normal operation, the maximum possible von Mises stress of the order of 62.4MPa can be expected on the bottom heat transfer panel, and the corresponding deflection of the structure is of the order of 0.62mm. All the stress checks were carried out with respect to the Structural Design Criteria for In-vessel Components (SDC-IC) criterion. The paper shall present the ES design highlighting its salient features.

Development of design Criteria for ITER In-vessel Components

Absrtract The components located inside the ITER vacuum chamber (in-vessel components – IC), due to their specific nature and the environments they are exposed to (neutron radiation, high heat fluxes, electromagnetic forces, etc.), have specific design criteria which are, in this paper, referred as Structural Design Criteria for In-vessel Components (SDC-IC). The development of these criteria started in the very early phase of the ITER design and followed closely the criteria of the RCC-MR code. Specific rules to include the effect of neutron irradiation were implemented. In 2008 the need of an update of the SDC-IC was identified to add missing specifications, to implement improvements, to modernise rules including recent evolutions in international codes and regulations (i.e. PED). Collaboration was set up between ITER Organization (IO), European (EUDA) and Russian Federation (RFDA) Domestic Agencies to generate a new version of SDC-IC. A Peer Review Group (PRG) composed by members of the ITER Organization and all ITER Domestic Agencies and code experts was set-up to review the proposed modifications, to provide comments, contributions and recommendations.

Specification of properties and design allowables for copper alloys used in HHF components of ITER

Two types of copper alloys, precipitation hardened (PH) Cu (CuCrZr-IG) and dispersion strengthened (DS) Cu (CuAl25-IG), are proposed as heat sink materials for the high heat flux (HHF) components of ITER. However, copper alloys are not included in any national codes, and properties of both CuCrZr and CuAl25 are not yet fully characterised. The performed R&D gives a basis for the specification of physical and mechanical properties required for the design analysis in accordance with the ITER Structural Design Criteria for In-vessel Components (SDC-IC). For both CuCrZr-IG and CuAl25-IG alloys, the statistical evaluation of available experimental data has been used to calculate the temperature dependence of the average value and of the 95% confidence limit of tensile properties. The stress limits, S m, S e, and S d, have been estimated on the basis of available data. The procedure used for specification of the temporary design limits of Cu alloys is described in the paper.