Conceptual study of vertical sector transport maintenance for DEMO fusion reactor

The maintenance scheme is a critical issue for DEMO design, and requires high availability of the reactor. The SlimCS, designed in JAEA, adopts the horizontal sector transport hot cell maintenance scheme. In order to determine the most appropriate DEMO reactor maintenance scheme, it is important to assess the various maintenance schemes. In this paper the vertical sector transport maintenance concept is proposed for the first time. In the sector maintenance scheme, the amount of cutting/re-welding of the piping is minimized. The sector including blanket modules and a high-temperature shield was divided into 10° segments in a toroidal direction. The sectors are designed to be removed and re-inserted through upper alternate vertical maintenance ports. In the case of the vertical sector transport maintenance scheme, inter-coil structures could be adopted for use against turnover force in toroidal field (TF) coils. This shows an advantage in DEMO maintenance compared with horizontal sector transport.     

Waste management scenario in the hot cell and waste storage for DEMO

Management strategy of radioactive waste generated in periodic replacement may be important in the point of view of fusion reactor design, because it has a large impact on the design of the hot cell and waste storage in the plant. In the replacement period of a fusion power reactor, the assembly of blanket or divertor modules needs to be removed from the reactor in order to minimize remote maintenance in the vacuum vessel and to attain reasonable plant availability. In the hot cell, the modules will be removed from the back plate of the assembly. Here, note that the active cooling must be done by a way that does not cause contamination of the hot cell environment due to dispersion of tritium and tungsten dust. In this sense, the cooling scenario is adopted that the existing pipe of cooling water in the assembly is connected to a different cooling water system in the hot cell. On the other hand, it is assumed that the structural material (F82H) of the blanket and divertor is not recycled due to its high contact dose rate. It should be crushed into small pieces to reduce volume of the waste and required storage space. In this paper, the basic idea of the waste management scenario and the conceptual design in the hot cell and waste storage for DEMO has been proposed.     

Pre-conceptual design assessment of DEMO remote maintenance

EDFA, as part of the Power Plant Physics and Technology programme, has been working on the pre-conceptual design of a Demonstration Power Plant (DEMO). As part of this programme, a review of the remote maintenance strategy considered maintenance solutions compatible with expected environmental conditions, whilst showing potential for meeting the plant availability targets. A key finding was that, for practical purposes, the expected radiation levels prohibit the use of complex remote handling operations to replace the first wall. In 2012/2013, these remote maintenance activities were further extended, providing an insight into the requirements, constraints and challenges. In particular, the assessment of blanket and divertor maintenance, in light of the expected radiation conditions and availability, has elaborated the need for a very different approach from that of ITER. This activity has produced some very informative virtual reality simulations of the blanket segments and pipe removal that are exceptionally valuable in communicating the complexity and scale of the required operations. Through these simulations, estimates of the maintenance task durations have been possible demonstrating that a full replacement of the blankets within 6 months could be achieved. The design of the first wall, including the need to use sacrificial limiters must still be investigated. In support of the maintenance operations, a first indication of the requirements of an Active Maintenance Facility (AMF) has been elaborated.     

DEMO divertor maintenance

This paper presents a DEMO divertor segmentation and maintenance scheme together with maintenance time estimations. As far as it is known, it is the first such study for DEMO, and the work was coordinated by EFDA. The approach of the study was that DEMO divertor and its handling shall be very similar to that of ITER, therefore ITER divertor segmentation and maintenance schemes were used as starting point. The maintenance scheme for both ITER and DEMO is the following: the divertor is segmented into cassettes. For maintenance 3 divertor ports are used through which all equipment and cassettes are inserted and removed. A toroidal rail based machine transports the element toroidally in front of the port, from where a radial tractor is used that move it between cask and VV. The main difference between DEMO and ITER is the increase in major radius from 6.2 to 7.5 m, and the cassette segmentation changed from 6.66° to 11.25°. These two factors increase the cassette weight from 12 to 25 tons. Therefore the cantilevered radial transportation of ITER cassettes was not adapted, but instead, the first design of a new, bottom supported radial machine is proposed. For the entire blanket maintenance 70 days were calculated instead of the 179 days in ITER. It was also concluded, that if in-vessel toroidal machine is used for blanket maintenance, there are no real possibilities for parallel blanket/divertor replacement.     

Preliminary concept design of the divertor remote handling system for DEMO power plant

This paper is based on the remote maintenance system project (WPRM) for the demonstration fusion power reactor (DEMO). Following ITER, DEMO aims to confirm the capability of generating several hundred of MW of net electricity by 2050. The main objective of these activities is to develop an efficient and reliable remote handling (RH) system for replacing the divertor cassettes.This paper presents the preliminary results of the concept design of the divertor RH system. The proposed divertor mover is a hydraulic telescopic boom driven from the transportation cask through the maintenance tunnel of the reactor. The boom is divided in three sections of 4 m each, and it is driving an end-effector in order to perform the scheduled operations of maintenance inside the vacuum vessel.Two alternative design of the end effector to grip and manipulate the divertor cassette are also presented in this work. Both the concepts are hydraulically actuated, basing on the ITER previous studies. The divertor cassette end-effector consists of a lifting arm linked to the divertor mover, a tilting plate, a cantilever arm and a hook-plate.The main objective of this paper is to illustrate the feasibility of DEMO divertor remote maintenance operations.     

Dynamic thermal-hydraulic modelling of the EU DEMO HCPB breeding blanket cooling loops

A global, system-level thermal-hydraulic model of the EU DEMO tokamak fusion reactor is currently under development and implementation in a suitable software at Politecnico di Torino, including the relevant heat transfer and fluid dynamics phenomena, which affect the performance of the different cooling circuits and components and their integration in a consistent design. The model is based on an object-oriented approach using the Modelica language, which easily allows to preserve the high modularity required at this stage of the design. The first module of the global model will simulate the blanket cooling system and will be able to investigate different coolant options and different cooling schemes, to be adapted to the different blanket systems currently under development in the Breeding Blanket (BB) project. The paper presents the Helium-Cooled Pebble Bed (HCPB) module of the EU DEMO blanket cooling loops system model. The model is used to compare different schemes for the cooling of the different components of the HCPB BB, and to suggest improvements aimed at optimizing the pumping power required by the cooling system. The model is then used to analyse a pulsed scenario, characteristic of the EU DEMO operation.     

Preliminary neutronic assessment of a helium-cooled Li8PbO6 breeding blanket design for DEMO

A preliminary neutronic assessment of the performances of a helium-cooled Li₈PbO₆ breeding blanket for the conceptual design of a DEMO fusion reactor is given. The study mainly focuses on TBR, power density responses and shielding factor optimization to estimate the feasibility of the design under the prescribed radiation deposition limits at TF-coils superconducting magnets. Computational analyses are based on three-dimensional 30° sector using the Monte Carlo code MCNPX 2.6. The scoping interest of helium-cooled Li₈PbO₆ blanket designs is based on a large potential minimization of the amount of Be required and the strong relaxation of ⁶Li enrichment requirements for this solution when compared to other solid breeder blanket options.     

Attachment concept for the “multi module segments” in a pulsed DEMO reactor

The pulsed thermonuclear demonstration reactor (DEMO) features challenging operational conditions such as high neutron fluxes, high temperatures, and significant thermo-mechanical stresses. These conditions do not require only a selection of advanced structural materials, but also the development of reliable means to assemble the in-vessel components together; allowing thermal expansions, disassembly, and maintenance in attractive scenarios. Over the course of DEMO lifetime, the materials are subjected to embrittlement by neutron irradiation, swelling, considerable thermo-mechanical fatigue and creep. Traditional joining methods may be rarely used in the harsh fusion environment to assemble different components. In addition any proposed layout should cope with the limited space available inside the vacuum vessel (VV).The objective of this study is to review the proposed attachment systems (developed within the latest European DEMO Conceptual Study) for the vertical segmentation concept called “multi module segments” (MMS). In order to find some place to house the attachments the blanket is cut respecting the Tritium Breeding Ratio limit for tritium self sufficiency. The conditions, neutronic and thermal, in which the attachments are supposed to operate, are calculated. The effects of pulsed operations have also been taken into account. The design of the attachments with the available structural materials with and without an active cooling system is analysed and a new concept for plug/unplug attachments is also suggested.     

The technological features of manufacturing process for RF CHC TBM

The development of manufacturing technology for the ceramic helium-cooled test blanket module (CHC TBM) is performed in the framework of the concept for RF Federal government program to master the fusion nuclear energy and as a part of the development of DEMO blanket technology.The main technical approach to the development of CHC TBM manufacturing technology is to provide the “combined” analogy with design decisions of DEMO blanket structural elements.The manufacturing technology of CHC TBM structural elements (first wall cramp, load-bearing back cramp, tritium-breeding element and attachment system) has been proposed during the period of 2004-2007. The design details of TBM structural elements and critical issues of manufacturing technology development are also presented in this paper.     

Normal operation and maintenance safety lessons from the ITER US PbLi test blanket module program for a US FNSF and DEMO

A leading power reactor breeding blanket candidate for a fusion demonstration power plant (DEMO) being pursued by the US Fusion Community is the Dual Coolant Lead Lithium (DCLL) concept. The safety hazards associated with the DCLL concept as a reactor blanket have been examined in several US design studies. These studies identify the largest radiological hazards as those associated with the dust generation by plasma erosion of plasma blanket module first walls, oxidation of blanket structures at high temperature in air or steam, inventories of tritium bred in or permeating through the ferritic steel structures of the blanket module and blanket support systems, and the ²¹⁰Po and ²⁰³Hg produced in the PbLi breeder/coolant. What these studies lack is the scrutiny associated with a licensing review of the DCLL concept. An insight into this process was gained during the US participation in the ITER Test Blanket Module (TBM) Program. In this paper we discuss the lessons learned during this activity and make safety proposals for the design of a Fusion Nuclear Science Facility (FNSF) or a DEMO that employs a lead lithium breeding blanket.     

Evaluation of remote maintenance schemes by plasma equilibrium analysis in Tokamak DEMO reactor

The remote maintenance schemes in a DEMO reactor are categorized by insertion direction, blanket segmentation, and divertor maintenance scheme, and are quantitatively evaluated by analysing the plasma equilibrium. The positions of the poloidal field (PF) coil are limited by the size of the toroidal field (TF) coil and the maintenance port layout of each remote maintenance scheme. Because the PF coils are located near the larger TF coil and far from the plasma surface, the horizontal sector transport maintenance scheme requires the largest part of total PF coil current, 25% larger than that required for separated sector transport using vertical maintenance ports with segmented divertor maintenance (SDM). In the unsegmented divertor maintenance (UDM) scheme, the total magnetic stored energy in the PF coils at plasma equilibrium is about 30% larger than that stored in the SDM scheme, but the time required for removal and installation of all the divertor cassettes in the UDM scheme is roughly a third of that required in the SDM scheme because the number of divertor cassettes in the UDM scheme is a third of that in the SDM scheme. From the viewpoint of simple maintenance operations, the merit of the UDM scheme has more merit than the SDM scheme.     

Research and development status on fusion DEMO reactor design under the Broader Approach

The main objective of DEMO design activity under the Broader Approach is to develop pre-conceptual design of DEMO options by addressing key design issues on physics, technology and system engineering. This paper describes the latest results of the design activity, including DEMO parameter study, divertor and remote maintenance. DEMO parameter study has recently started with “pulsed” DEMO having a major radius (R_p) of 9 m, and “steady state” DEMO of R_p = 8.2 m or more. Divertor design study has focused on a computer simulation of fully detached plasma under DEMO divertor conditions and the assessment of advanced divertor configuration such as super-X. Comparative study of various maintenance schemes for DEMO and narrowing down the schemes is in progress.     

Maintenance duration estimate for a DEMO fusion power plant,based on the EFDA WP12 pre-conceptual studies

The erosion and high neutron flux in a fusion power plant results in the need for frequent remote replacement of the plasma facing components. This is a complex and time consuming remote handling operation and its duration directly affects the availability and therefore the commercial viability of the power plant.A tool is needed to allow the maintenance duration to be determined so that developments in component design can be assessed in terms of their effect on the maintenance duration. This allows the correct balance to be drawn between component cost and performance on the one hand and the remote handling cost and plant availability on the other.The work to develop this tool has begun with an estimate of the maintenance duration for a fusion power plant based on the EFDA DEMO WP12 pre-conceptual design studies [1]. The estimate can be readily adjusted for changes to the remote maintenance process resulting from design changes. The estimate uses data extrapolated from recorded times and operational experience from remote maintenance activities on the JET tokamak and other nuclear facilities.The Power Plant Conceptual Study from 2005 [2] proposes that commercial viability of a power plant would require an availability of 75% or above. Results from the maintenance estimate described in this paper suggest that this level of availability could be achieved for the planned maintenance using a highly developed and tested remote maintenance system, with a large element of parallel working and challenging but feasible operation times.     

Prolegomena to the RAMI analysis of a conceptual LHCD system for DEMO

Future fusion reactors are asked to deliver large amount of energy in a continuative way to electrical power grids. In this frame it is very important to care the availability of the reactors since the preliminary phases of the design. Techniques to assure high levels of reliability must be extensively used and accurate programmes of maintenance must be studied, caring in the same time the training of maintenance personnel. Finally the layout of the reactor and of all its auxiliary systems must be carefully examined in order to assure an easy inspectability and a fast recovery of the power station after unavoidable failures. All the previously listed activities are summarized by the acronym RAMI (reliability, availability, maintainability and inspectability). The basics of the RAMI analysis for a conceptual LHCD system for DEMO are given in the paper.     

Contribution to DEMO reactor RH maintenance assessment

The scope of this paper is a preliminary assessment of the maintenance scheme in support of the European study for the next generation of fusion reactor: DEMO. Despite other fusion machine requiring remote handling maintenance operations, DEMO is supposed to work under steady state operational conditions. Therefore, requirement on the maintenance scheme is stronger. To target a good availability of the machine along machine operation plan, it is necessary to draw an adequate maintenance scheme. Indeed, due to the high fluxes generated by the plasma in the vacuum vessel, the first wall lifetime is limited, so the frequent replacement is necessary. On current fusion experimental machine, as first wall load conditions are less severe, DEMO condition implies high level of requirement on maintenance time. During DEMO lifetime, several full first wall replacements are expected. To provide access to the vacuum vessel machine for first wall removal, preparatory work is required to set the machine to adequate maintenance conditions and to open the machine properly, the same situation at the end of the maintenance period. Shutdown duration for first wall replacement should be as short as possible to reach the availability target of the machine. From this statement, the maintenance duration should not exceed 20% of the total lifetime of the reactor operation. First wall segmentation (i.e. total number of component to replace) has a high impact onto the replacement time. Considering the number of feasible designs for the first wall segmentation, we concentrate remote handling concept assessments one type of segmentation, the one minimizing the numbers of modules to replace [4], [5], [6]. Assumption on Divertor segmentation for these DEMO studies have similarities with Divertor ITER design; therefore ITER design output is relevant [1], [2]. We assume divertor removal performed in shadow time, while removing the other first wall modules.     

Conceptual integrated approach for the magnet system of a tokamak reactor

In the framework of the reflexion about DEMO, a conceptual integrated approach for the magnet system of a tokamak reactor is presented. This objective is reached using analytical formulas which are presented in this paper, coupled to a Fortran code ESCORT (Electromagnetic Superconducting System for the Computation of Research Tokamaks), to be integrated into SYCOMORE, a code for reactor modelling presently in development at CEA/IRFM in Cadarache, using the tools of the EFDA Integrated Tokamak Modelling task force. The analytical formulas deal with all aspects of the magnet system, starting from the derivation of the TF system general geometry, from the plasma main characteristics. The design criteria for the cable current density and the structural design of the toroidal field and central solenoid systems are presented, enabling to deliver the radial thicknesses of the magnets and enabling also to estimate the plasma duration of the plateau. As a matter of fact, a pulsed version DEMO is presently actively considered in the European programmes. Considerations regarding the cryogenics and the protection are given, affecting the general design. An application of the conceptual approach is presented, allowing a comparison between ESCORT output data and actual ITER parameters and giving the main characteristics of a possible version for DEMO.     

An Analysis of Ripple and Error Fields Induced by a Blanket in the CFETR

The Chinese Fusion Engineering Tokamak Reactor(CFETR) is an important intermediate device between ITER and DEMO. The Water Cooled Ceramic Breeder(WCCB)blanket whose structural material is mainly made of Reduced Activation Ferritic/Martensitic(RAFM) steel, is one of the candidate conceptual blanket design. An analysis of ripple and error field induced by RAFM steel in WCCB is evaluated with the method of static magnetic analysis in the ANSYS code. Significant additional magnetic field is produced by blanket and it leads to an increased ripple field. Maximum ripple along the separatrix line reaches 0.53% which is higher than 0.5% of the acceptable design value. Simultaneously, one blanket module is taken out for heating purpose and the resulting error field is calculated to be seriously against the requirement.     

Conceptual design study of fusion DEMO plant at SWIP

The basic definition and development strategy of the DEMO plant based on the Chinese fusion power plant (FPP) program are presented briefly. A conceptual design study of fusion HCSB-DEMO reactor with a fusion power of 2550 MW and a neutron wall loading of 2.3 MW/m² is performed recently. Three sets parameters of core plasma for different DEMO design objectives are proposed. A helium-cooled blanket system with ceramic breeder (Li₄SiO₄), the structure material of low-activation ferritic steel (LAF/M) and Be neutron multiplier based on Chinese ITER HCSB-TBM design foundation are considered. The design parameters, preliminary analyses and the basic structure as well as development strategy of HCSB-DEMO reactor are introduced.     

Nuclear heating and helium production at the inboard divertor rail for DEMO

Monte Carlo simulations were carried out for the DEMO model. Distributions of both the nuclear heating and the helium production in the area between the blanket and the divertor were calculated with the MCNP5 code for the reference case, when the DEMO geometry was not changed. Next a segment of the divertor and the lower part of the manifold were modified. Two new arrangements were studied. The simulations show that for one of the examined cases the helium production and the nuclear heating can be reduced roughly three or even four times in the investigated area. Besides the nuclear heating and the He production were estimated at the fastener (bolt head). The use of the modified divertor and a rail protecting the blanket is essential in the DEMO design.     

Efficiency and availability driven R&D issues for DEMO

DEMO is the main step foreseen after ITER to demonstrate the technological and commercial viability of a fusion power plant. DEMO R&D requirements are usually identified on the basis of the functions expected from each individual system. An approach based on the analysis of overall plant functional requirements sheds new light on R&D needs. The analysis presented here focuses on two overall functional requirements, efficiency and availability. The results of this analysis are presented here putting emphasis on systems not sufficiently considered up to now, e.g. the heating and current drive systems, while more commonly addressed systems such as tritium breeding blankets are not discussed in detail. It is also concluded that an overall functional analysis should be adopted very early in the DEMO conceptual design studies in order to provide a fully integrated approach, which is an absolute requirement to ensure that the ambitious goals of this device will be ultimately met.