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.

Optimization of SPIDER Grounded Grid Segment Design

Source for the production of ions of deuterium extracted from a radio frequency plasma (SPIDER) is a full-size negative ion source and a multibeamlet 100-kV accelerator, operational since 2018. It serves as the test bed for the full development of the heating neutral beam system for megavolt ITER injector and concept advancement (ITER). In 2020, the third grounded grid segment broke at the electrodeposited copper layer near one of the four feeders causing leakage. This article reports the redesign development of the SPIDER GG segment. In the first part of the article, several design configurations are proposed and compared to each other. The comparison serves to identify a suitable redesign configuration. A set of coupled analyses, including computational fluid dynamic (CFD) and finite-element method (FEM) analyses, is then performed to finally evaluate the redesigned component in accordance with the ITER structural design criteria for in-vessel components (ITER SDC-IC) criteria.

Steady state thermo-mechanics and material property definition framework for analyzing DCLL blanket in the fusion nuclear science facility

A thermo-mechanics model that relies on creating the material property definition framework (MPDF) and multiphysics coupling of the heat transfer and the solid mechanics modules is developed to determine the structural integrity of the recently designed dual cooled lead lithium (DCLL) inboard blanket (IB) for the Fusion Nuclear Science Facility under steady state loads. The MPDF is called to supply fusion relevant neutron irradiation and temperature induced changes in material properties during multiphysics finite element runs, and PbLi temperature profiles are used to approximate Magnetohydrodynamics effect and the nuclear volumetric heating on the PbLi. Neutron irradiation and temperature induced reduction of the yield and ultimate strengths of F82H steel at the first wall (FW) are quantified for one year. A blanket in an assembly with gaps between blanket sectors and another blanket in an assembly with no gaps between blanket sectors, both exposed to radiation damage that lasted for one year are analyzed. Analysis using the elastic ITER structural design criteria for in-vessel components (ITER SDC-IC) design rules and a linear isotropic-hardening-type elastoplastic material model are used where most appropriate. The IB blanket with gaps between blanket sectors will withstand the steady state combined thermal and coolant loads for one year operational period but will fail if no gaps are allowed between blanket sectors. It is recommended that a gap of about 7.62 mm should be provided between IB blanket sectors during assembly which would close up during service, stop neutron streaming, reduce stresses and reduce bending of the FW into the scrape-off layer.

Effects of stress triaxiality and strain rate on the fracture of a CuCrZr alloy

Precipitation hardened Copper-Chromium-Zirconium (CuCrZr) alloy is a prime candidate for divertor components in future European DEMOnstration (DEMO) fusion reactors. To develop the DEMO Design Criteria for In-vessel Components (DDC-IC), the failure criterion of CuCrZr needs to be investigated. Hence, the effects of stress triaxiality and loading strain rate on the fracture of solution heat treated and annealed CuCrZr alloys were studied using digital-imaging-correlation and electron microscopy techniques. It was found that an increase in the stress triaxiality caused a significant decrease of over 50% in the equivalent strain to fracture. On the other hand, increasing applied strain rate from 1.3 × 10−5 s−1to 8.6 × 10−3 s−1 had no considerable effects on the yield stress and elongation. However, higher flow stresses and a larger number density of voids were observed under the highest applied strain rate of 8.6 × 10−3 s−1. Fracture surface analysis showed that the failure model was dominated by void growth and coalescence for all the tests. This work has been performed within the Engineering Data and Design Integration (EDDI) sub-project of the EUROfusion Materials work package and aimed to contribute to the development of the DEMO Design Criteria for In-vessel Components.

Modal and response spectrum analyses of an ITER divertor module

The present study is to investigate resistance of an ITER divertor module against typical seismic loads. Two kinds of complex finite element (FE) models, which consists of cassette body, inner and outer vertical targets, dome and stabilizers, were developed; one is 1- or 2-layer simplified models without consideration of inner tubes. The other is a 3-layer detailed model with the inner tubes as well as cooling water. Modal analyses to predict dynamic characteristics were conducted in use of block-Lanczos algorithm for all the FE models. Subsequently, response spectrum analyses were performed by employing three modal combination techniques and taking into account different seismic magnitudes. Analysis results from the detailed model led to higher von-Mises stresses than those obtained from the simplified model. Moreover, calculated stress intensities at critical locations were compared with corresponding design stress intensities, according to ITER structural design criteria for in-vessel components, which satisfied sufficiently the structural assessment requirement.

Thermo-structural design of the European DEMO water-cooled blanket with a multiscale-multiphysics framework

The Water-Cooled Lithium-Lead (WCLL) blanket concept is one of four possible embodiments under consideration for the European Demonstration Power Plant DEMO, beyond ITER. It is based on a water-cooled first wall and blanket structure with the RAFMS (Reduced Activation Ferritic/Martensitic Steel, Eurofer 97) as structural material and PbLi as neutron multiplier and tritium breeder. We first develop a multiphysics modeling framework in order to optimize the design and achieve long blanket lifetime, maintainability, and high reliability. The approach is demonstrated by coupling Computational Fluid Dynamics (CFD), heat transfer and solid mechanics. The multiphysics framework is then utilized to provide the loading conditions for a multiscale design approach particularly focused on modeling the effects of both steady-state and operational transients on the structural integrity of the First Wall (FW). The methodology builds on the ITER Structural Design Criteria for In-Vessel Components (ISDC-IC), then extending the criteria to progressively incorporate continuum plasticity models and microstructure-based representation of deformation and fracture. We finally introduce advanced fracture mechanics concepts based on a Materials-Specific Failure Assessment Diagram (MS-FAD), in which precise calculations of the J-integral of an elasto-plastic material is required. This is especially important for the water-cooled design because of the limited fracture toughness of Eurofer 97 below 300 °C. The methodology has been applied to screen the FW/Blanket performance of a range of steady state heat flux assumed to be applied during the plasma pulse flat top, and up to 0.8 MW/m2. It has become clear that the radiation steady state load will be lower than the values screened here, while the local heat flux due to particle loads will lead to global higher values than the values investigated here. The methodology will be further fine-tuned to specific loading conditions in order to assess actual performance limits of the DEMO WCLL BB.

Materials engineering and design for fusion—Towards DEMO design criteria

Future fusion reactors will operate under unprecedented environmental conditions and will rely on the performance of complex in-vessel components during long term operation. These new reactors have key operational differences compared to existing nuclear systems and require new materials and new design criteria for in-vessel components. Within the EUROfusion consortium, several approaches are being taken in parallel: modification of existing frameworks and development of new databases, handbooks and design criteria specific for EU-DEMO operation. These adaptations and advances need to be considered within the context of the novel materials and multi-material interfaces being considered for the unprecedented fusion in-vessel components operational environment. The materials properties required for qualification in conventional nuclear sites are reviewed along with proposed pragmatic adaptations required for DEMO reactors arising from the potentially sparse fusion irradiation spectrum materials database. The efforts ongoing within EUROfusion to assess the links between materials engineering and design for fusion reactors are highlighted focusing on developments of new DEMO specific materials databases and handbook and DEMO Design Criteria. The efforts in EUROfusion are shown to form key steps towards new fusion reactor specific design codes.

Multiphysics modeling of the FW/Blanket of the U.S. fusion nuclear science facility (FNSF)

The dual coolant lead-lithium (DCLL) blanket concept, which is utilized in the Fusion Nuclear Science Facility (FNSF) conceptual design, is based on a helium-cooled first wall and blanket structure with RAFS (Reduced Activation Ferritic Steel) and a self-cooled LiPb breeding zone. The objective of this work is to develop a multiphysics modeling process in order to optimize the design and achieve long lifetime, maintainability, and high reliability. 3D finite element multiphysics modeling of the DCLL first wall and blanket (midplane of one sector) has been performed using COMSOL 5.2. The multiphysics aspect of the design is demonstrated via coupling of Computational Fluid Dynamics (CFD), conjugate heat transfer and solid mechanics. Both normal and off-normal loading conditions have been analyzed. The results of velocity, pressure, and temperature distributions of helium flow, as well as the primary and thermal stress of the structure were obtained. This was followed by determination of the factors of safety along three critical paths based on the ITER Structural Design Criteria for In-vessel Components (ISDC-IC). We show here that the structural design meets the ITER-ISDC design rules under both normal and off-normal operating conditions, though the safety factors under off-normal condition with 8MPa helium pressure are marginal. Thus simple design optimization was conducted based on a parametric study on first wall dimensions to improve the design.

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.

Diagnostic Neutral Beam for ITER—Concept to Engineering

Following the allocation of the procurement of the diagnostic neutral beam (DNB) to the Indian DA, a series of tasks have been undertaken to first assess the DNB configuration and arrive at an optimal beam-line configuration folding in the gas-feed and vacuum-pumping requirements. Specific emphasis is placed on the thermal, structural, and electrical designs of beam-line components, in order to ensure their compatibility with the criteria specified for ITER in vessel components, i.e., Structural Design Criteria for In-Vessel Components. The detailed assessment of manufacturing technologies and their compatibility with the ITER standards forms an integral part of the design. A common approach to manufacturing for DNB and heating-and-current-drive NB components shall be undertaken through a comprehensive prototyping phase which shall lead to built-to-print specifications. In addition to safety and remote-handling issues, the design also addresses the requirements of interfaces related to other systems such as cryo, hydraulic, pneumatic, vacuum pumping, gas feed, civil, power supplies and transmission, CODAC, etc. The successful delivery of DNB is dependent on two critical R&D aspects: 1) the production of a uniform low-divergence beam from the beam source and 2) a well-controlled transmission through lengths of ~ 22 m. The first shall primarily be a subject of the Ion Source Test Facility-SPIDER [part of NB test facility (MITICA in Padova)]-where India is involved as a collaborator and the Indian test bed, where issues for DNB beam source which were not resolved in the SPIDER would be taken up. The second shall form one of the primary objectives of the Indian test bed to characterize the DNB. This paper presents the progress in DNB from the concept level to an engineered system along with the plans for system integration and an R&D intensive implementation.

Disruption Design Criteria for Joint European Torus In-Vessel Components

In view of the modification to the Joint European Torus (JET) plasma facing components foreseen for the 2004 shutdown, the disruption design criteria for in-vessel components have been updated building on the operational experience with divertor plasmas gained since the early 1990s. In fast disruptions the largest contribution to the electromechanical loads comes from currents induced by the poloidal field change. This is proportional to the plasma current decay rate, the maximum of which is observed to be linear with the predisruption plasma current, as if the current quench in the fastest events has a fixed duration, around 10 ms. Usually vertical displacement events (VDEs) take place on a longer timescale. In these cases halo currents determine the worst loading condition. Analysis of recent VDE data confirmed the previously observed magnitude of asymmetries: toroidal peaking factor times ratio of average poloidal halo to initial plasma current up to 0.42.Experimental evidence to justify the new criteria and procedures for applying them to JET are included. The revised design criteria are discussed and compared with those used for the components already present in the JET vessel.

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.