An exploratory study on the gaps and pathways to the Korean fusion DEMO

With the vision of being an early demonstrator of fusion energy, the strategic plans for the Fusion DEMO program of Korea (K-DEMO program) has been developed. A staged development of the K-DEMO plant was considered in the strategic plans as to verify technical feasibility in the first stage and economic feasibility in the second stage. The top-tier design requirements and assumptions of the first stage K-DEMO plant are defined and postulated. With these requirements and assumptions, the desired and current status of nuclear fusion technologies are compared to identify the gaps to be filled to design, fabricate, construct, and operate it. The pathways from KSTAR, ITER to K-DEMO plant have also been studied to identify R&D activities for K-DEMO program that are to go in parallel with KSTAR and ITER are extracted from the pathways. Cross-cutting with the fusion R&D activities of the other countries and utilizing the commonalities with the existing systems are discussed with the provision of open-innovation strategy that is one of the key strategies of K-DEMO program. The priority of the R&D activities of K-DEMO program is qualitatively determined in consideration of the gaps, cross-cutting, and risks associated with the R&D investments.     

Japanese contribution to the DEMO-R&D program under the Broader Approach activities

Several technical R&D activities mainly related to the blanket materials are newly launched as a part of the Broader Approach (BA) activities, which was initiated by the EU and Japan. According to the common interests for these parties in DEMO, R&Ds on reduced activation ferritic/martensitic (RAFM) steels as structural material, SiC_f/SiC composites as a flow channel insert material and/or alternative structural material, advanced tritium breeders and neutron multipliers, and tritium technology are carried out through the BA DEMO R&D program, in order to establish the technical bases on the blanket materials and the tritium technology required for DEMO design. This paper describes overall schedule of those R&D activities and recent progress in Japan carried out by JAEA as the domestic implementing agency on BA, collaborating with Japanese universities and other research institutes.     

Overview of EU DEMO design and R&D activities

One important objective of the EU fusion roadmap Horizon 2020 is to lay the foundation of a Demonstration Fusion Power Reactor (DEMO) to follow ITER, with the capability of generating several 100 MW of net electricity to the grid and operating with a closed fuel-cycle by 2050. This is currently viewed by many of the nations engaged in the construction of ITER as the remaining crucial step towards the exploitation of fusion power. This paper outlines the DEMO design and R&D approach that is being adopted in Europe and presents some of the preliminary design options that are under evaluation as well as the most urgent R&D work that is expected to be launched in the near-future. The R&D on materials for a near-term DEMO is discussed in detail elsewhere.     

Evolution of the ITER program and prospect for the next-step fusion DEMO reactors: status of the fusion energy R&D as ultimate source of energy

An international joint project of fusion experimental reactor, the ITER (International Thermonuclear Experimental Reactor), is reviewed in view of long-range fusion energy research and development (R&D). Its purpose, goal, evolution, and the present construction status are briefly reviewed. While the ITER is a core machine in the present stage, generation of electricity is a role of the next-step fusion demonstration power plant “DEMO.” The status of designs and technology R&D for DEMO are also reviewed.     

On the NBI system for substantial current drive in a fusion power plant: Status and R&D needs for ion source and laser neutralizer

The requirements for the heating and current drive systems of a fusion power plant will strongly depend on the DEMO scenario. The paper discusses the R&D needs for a neutral beam injection system — being a candidate due to the highest current drive efficiency — for the most demanding scenario, a steady state tokamak DEMO. Most important issues are the improvement of the wall-plug efficiency from the present ∼25% to the required 50–60% by improving the neutralization efficiency with a laser neutralizer system and the improvement of the reliability of the ion source operation. The demands on and the potential of decreasing the ion source operation pressure, as well as decreasing the amount of co-extracted electrons and backstreaming ions are discussed using the ITER requirements and solutions as basis. A further concern is the necessity of cesium for which either the cesium management must be improved or alternatives to cesium for the production of negative ions have to be identified.     

Progress and achievements of the BA activities in Rokkasho

In Rokkasho Japan, the International Fusion Energy Research Center (IFERC) project and the International Fusion Materials Irradiation Facility/Engineering Validation and Engineering Design Activities (IFMIF/EVEDA) project are on going under the Broader Approach framework. The IFERC project consists of three sub-projects; a fusion demonstration reactor (DEMO) Design and R&D Coordination Center, a Computational Simulation Center (CSC), and an ITER Remote Experimentation Center (REC). DEMO Design activity has been conducted by the IFERC project team in Rokkasho and home teams in EU and JA. In the DEMO R&D activity, five R&D tasks mainly of the blanket materials are carried out intensively. A supercomputer with 1.23 Pflops of LINPAC performance has been installed in December 2011, the operation started in January 2012. Discussion of overall plan of REC has started in 2012 between EU and Japan. In the IFMIF/EVEDA project, an IFMIF prototype accelerator system up to 9 MeV with 125 mA CW deuteron beam will be installed and tested in Rokkasho. Major components of the accelerator are under development or fabrication in EU. The first component of the accelerator, an injector with an ECR ion source, will be delivered to Rokkasho in March 2013.     

Design activities on helical DEMO reactor FFHR-d1

Based on high-density and high-temperature plasma experiments in the large helical device (LHD), conceptual design studies of the LHD-type helical DEMO reactor FFHR-d1 have been conducted by integrating wide-ranged R&D activities on core plasmas and reactor technologies through cooperative researches under the fusion engineering research project, which has been launched newly in NIFS. Current activities for the FFHR-d1 in this project are presented on design window analyses with designs on core plasma, neutronics for liquid blankets, continuous helical magnets, pellet fueling, tritium systems and plasma heating devices.     

Overview of Helium Cooled Ceramic Reflector Test Blanket Module development in Korea

Korea plans to install and test Helium Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM) in the ITER, because the HCCR blanket concept is one of options of the DEMO blanket. Currently, many design and R&D activities have been performed to develop the Korean HCCR TBM. An integrated design tool for a fusion breeder blanket has been developed based on nuclear technologies including a safety analysis for obtaining a license for testing in the ITER. A half-scale sub-module mockup of the first wall with the manifold was fabricated, and the manufacturability and thermo-hydraulic performances were evaluated. High heat load and helium cooling test facilities have been constructed. Next, the recent status of TBM material development in Korea was introduced including Reduced Activation Ferritic Martensitic (RAFM) steel, lithium ceramic pebbles and silicon carbide (SiC) coated graphite pebbles. Several fabrication methods of RAFM steel, lithium ceramic pebbles, and silicon carbide coating on graphite pebbles were investigated. Recent design and R&D progress on these areas are introduced here.     

Safety studies on Korean fusion DEMO plant using integrated safety assessment methodology

The purpose of this paper is to suggest methodology that can investigate safety issues and provides a case study for Korean fusion DEMO plant (K-DEMO) as a part of R&D program through the National Fusion Research Institute of Korea. Even though nuclear regulation and licensing framework is well setup due to the operating and design experience of Pressurized Water Reactors (PWRs) since 1970s, the regulatory authority of South Korea has concerns on the challenge of facing new nuclear facilities including K-DEMO due to the differences in systems, materials, and inherent safety feature from conventional PWRs. Even though the follow-up of the ITER license process facilitates to deal with significant safety issues of fusion facilities, a licensee as well as a licenser should identify the gaps between ITER and DEMO in terms of safety issues. First we reviewed the methods of conducting safety analysis for unprecedented nuclear facilities such as Generation IV reactors, particularly very high temperature reactor (VHTR), which is called as integrated safety assessment methodology (ISAM). Second, the analysis for the conceptual design of K-DEMO on the basis of ISAM was conducted. The ISAM consists of five analytical tools to develop the safety requirements from licensee's viewpoint in interactive manner: qualitative safety features review (QSR), phenomena identification and ranking table (PIRT), objective provision tree (OPT), probabilistic safety assessment (PSA), and deterministic and phenomenological analysis (DPA). Considering the design phase of K-DEMO, the current study focused on the PIRT process with the fusion safety advisory group in South Korea.     

The European contribution to the ITER Remote Maintenance

For a first-of-a-kind nuclear fusion reactor like ITER, remote maintainability of neutron-activated components is one of the key aspects of plant design and operations, and a fundamental ingredient for the demonstration of long-term viability of fusion as energy source.The European Domestic Agency (EU DA, i.e. Fusion for Energy, F4E) is providing important support to the ITER Organisation (IO) in specifying the functional requirements of the Remote Handling (RH) Procurement Packages (i.e. the subsystems allocated to EU DA belonging to the overall ITER Remote Maintenance Systems IRMS), and in performing design and R&D activities – with the support of national laboratories and industries – in order to define a sound concept for these packages.Furthermore, domestic industries are being involved in the subsequent detailed design, validation, manufacturing and installation activities, in order to actually fulfil our procurement-in-kind obligations.After an introduction to ITER Remote Maintenance, this paper will present status and next stages for the RH systems allocated to EU DA, and will also illustrate complementary aspects related to cross cutting technologies like radiation tolerant components and RH control systems.Finally, the way all these efforts are coordinated will be presented together with the overall implementation scenario and key milestones.     

Requirements for helium cooled pebble bed blanket and R&D activities

This work aims to give an outline of the design requirements of the helium cooled pebble bed (HCPB) blanket and its associated R&D activities. In DEMO fusion reactor the plasma facing components have to fulfill several requirements dictated by safety and process sustainability criteria. In particular the blanket of a fusion reactor shall transfer the heat load coming from the plasma to the cooling system and also provide tritium breeding for the fuel cycle of the machine. KIT has been investigating and developed a helium-cooled blanket for more than three decades: the concept is based on the adoption of separated small lithium orthosilicate (tritium breeder) and beryllium (neutron multiplier) pebble beds, i.e. the HCPB blanket. One of the test blanket modules of ITER will be a HCPB type, aiming to demonstrate the soundness of the concept for the exploitation in future fusion power plants. A discussion is reported also on the development of the design criteria for the blanket to meet the requirements, such as tritium environmental release, also with reference to the TBM.The selection of materials and components to be used in a unique environment as the Tokamak of a fusion reactor requires dedicated several R&D activities. For instance, the performance of the coolant and the tritium self-sufficiency are key elements for the realization of the HCPB concept. Experimental campaigns have been conducted to select the materials to be used inside the solid breeder blanket and R&D activities have been carried out to support the design. The paper discusses also the program of future developments for the realization of the HCPB concept, also focusing to the specific campaigns necessary to qualify the TBM for its implementation in the ITER machine.     

Overview of R&D activities on tritium processing and handling technology in JAEA

In JAEA, the tritium processing and handling technologies have been studied at TPL (Tritium Process Laboratory). The main R&D activities are: the tritium processing technology for the blanket recovery systems; the basic tritium behavior in confinement materials; and detritiation and decontamination. The R&D activities on tritium processing and handling technologies for a demonstration reactor (DEMO) are also planned to be carried out in the broader approach (BA) program by JAEA with Japanese universities. The ceramic proton conductor has been studied as a possible tritium processing method for the blanket system. The BIXS method has also been studied as a monitoring of tritium in the blanket system. The hydrogen transfer behavior from water to metal has been studied as a function of temperature. As for the behavior of high concentration tritium water, it was observed that the formation of the oxidized layer was prevented by the presence of tritium in water (0.23 GBq/cc). A new hydrophobic catalyst has been developed for the conversion of tritium to water. The catalyst could convert tritium to water at room temperature. A new Nafion membrane has also been developed by gamma ray irradiation to get the strong durability for tritium.     

Overview of TBM R&D activities in India

In India, development of Lead–Lithium Ceramic Breeder (LLCB) blanket is being performed as the primary candidate of Test Blanket Module (TBM) towards DEMO reactor. The LLCB TBM will be tested from the first phase of ITER operation (H-H phase) in one-half of an ITER port no. 2. The Indian TBM R&D program is focused on the development of blanket materials and critical technologies: structural material (IN-RAFMS), breeding materials (Pb–Li, Li₂TiO₃), development of technologies for Lead–Lithium cooling system (LLCS), helium cooling system (HCS), tritium extraction system (TES) and TBM related fabrication technologies. This paper will provide an overview of LLCB TBM R&D activities under progress in India.     

Overview of R&D at TLK for process and analytical issues on tritium management in breeder blankets of ITER and DEMO

Safe, reliable, and efficient tritium management in the breeder blanket will have to face unprecedented technological challenges. Beside the efficiency for tritium recovery from the breeder blanket (Tritium Extraction (TES) and Coolant Purification Systems (CPS)), the accuracy for tritium tracking between the inner and the outer fuel cycle must also be demonstrated. This paper focuses on the recent R&D carried out at the Tritium Laboratory Karlsruhe to tackle these issues. For ITER, the recently consolidated TES and CPS designs comprise adsorption columns and getter beds operated in semi-continuous mode. Different approaches for the tritium accountancy stage (TAS) have been evaluated. Balancing static (batch-wise gas collection at the TBM outlets and the tritium plant) or dynamic (in/on-line) approaches with respect to the expected analytical performances and integration issues, the first conceptual design of the TAS for EU TBMs is presented. For DEMO, the overall strategy for tritium recovery and tracking has been revisited. The necessity for on-line real-time tritium accountancy and improved process efficiency suggest the use of continuous processes such as permeator and catalytic membrane reactor. The main benefits combining the PERMCAT process with advanced membranes is discussed with respect to process improvements and facilitated accountancy using spectroscopic methods.     

Lessons learnt from ITER safety & licensing for DEMO and future nuclear fusion facilities

One of the strong motivations for pursuing the development of fusion energy is its potentially low environmental impact and very good safety performance. But this safety and environmental potential can only be fully realized by careful design choices. For DEMO and other fusion facilities that will require nuclear licensing, S&E objectives and criteria should be set at an early stage and taken into account when choosing basic design options and throughout the design process.Studies in recent decades of the safety of fusion power plant concepts give a useful basis on which to build the S&E approach and to assess the impact of design choices. The experience of licensing ITER is of particular value, even though there are some important differences between ITER and DEMO. The ITER project has developed a safety case, produced a preliminary safety report and had it examined by the French nuclear safety authorities, leading to the licence to construct the facility. The key technical issues that arose during this process are recalled, particularly those that may also have an impact on DEMO safety. These include issues related to postulated accident scenarios, environmental releases during operation, occupational radiation exposure, and radioactive waste.     

A low wall-loading DEMO reactor design with high priority for early and reliable realization of a tokamak fusion reactor over the cost performance

Based on scientific databases adopted for designing ITER plasmas and on the advancement of fusion nuclear technology from the recent R&D program, a low wall-loading DEMO fusion reactor has been designed, where high priority has been given to the early and reliable realization of a tokamak fusion plasma over the cost performance. Since the major radius of this DEMO reactor is chosen to be 10 m, plasma ignition is achievable with a low fusion power of 0.8 GW and an operation period of 4–5 hours is available only with inductive current drive. The low ignition power makes it possible to adopt a first wall with an austenitic stainless steel, for which significant databases and operating experience exists, due to its use in the presence of neutron irradiation in fission reactors. In step with development of advanced materials, a step-wise increase of the fusion power seems to be feasible and realistic, because this DEMO reactor has the potential to produce a fusion power of 5 GW.     

Comparative studies for two different orientations of pebble bed in an HCCB blanket

The Indian Test Blanket Module(TBM) program in ITER is one of the major steps in its fusion reactor program towards DEMO and the future fusion power reactor vision. Research and development(RD) is focused on two types of breeding blanket concepts: lead–lithium ceramic breeder(LLCB) and helium-cooled ceramic breeder(HCCB) blanket systems for the DEMO reactor. As part of the ITER-TBM program, the LLCB concept will be tested in one-half of ITER port no. 2, whose materials and technologies will be tested during ITER operation. The HCCB concept is a variant of the solid breeder blanket, which is presently part of our domestic RD program for DEMO relevant technology development. In the HCCB concept Li_2TiO_3 and beryllium are used as the tritium breeder and neutron multiplier, respectively, in the form of a packed bed having edge-on configuration with reduced activation ferritic martensitic steel as the structural material. In this paper two design schemes, mainly two different orientations of pebble beds, are discussed. In the current concept(case-1), the ceramic breeder beds are kept horizontal in the toroidal–radial direction. Due to gravity, the pebbles may settle down at the bottom and create a finite gap between the pebbles and the top cooling plate, which will affect the heat transfer between them. In the alternate design concept(case-2), the pebble bed is vertically(poloidal–radial) orientated where the side plates act as cooling plates instead of top and bottom plates. These two design variants are analyzed analytically and 2 D thermal-hydraulic simulation studies are carried out with ANSYS, using the heat loads obtained from neutronic calculations.Based on the analysis the performance is compared and details of the thermal and radiative heat transfer studies are also discussed in this paper.     

Plant life management and modernisation: Research challenges in the EU

The European network of excellence NULIFE (nuclear plant life prediction) has been launched with a clear focus on integrating safety-oriented research on materials, structures and systems and exploiting the results of this integration through the production of harmonised lifetime assessment methods. NULIFE will help provide a better common understanding of the factors affecting the lifetime of nuclear power plants which, together with associated management methods, will help facilitate safe and economic long-term operation of existing nuclear power plants. In addition, NULIFE will help in the development of design criteria for future generations of nuclear power plant.NULIFE was kicked-off in October 2006 and will work over a 5-year period to create a single organization structure, capable of providing harmonised research and development (R&D) at European level to the nuclear power industry and the related safety authorities. Led by VTT (Technical Research Centre of Finland), the project has a total budget in excess of 8 million euros, with over 40 partners drawn from leading research institutions, technical support organizations, electric power utilities and manufacturers throughout Europe. NULIFE also involves many industrial organizations and, in addition to their R&D contributions, these take part in a dedicated End User Group.Over the last 15 years the European Commission has sponsored a significant number of R&D projects under the Euratom Framework Programme and its Joint Research Centre has developed co-operative European Networks for mutual benefits on specific topics related to plant life management. However, their overall impact has been reduced due to fragmentation. These networks are considered forerunners to NULIFE. The importance of the long-term operation of the plants has been recognized at European level, in the strategic research agenda of SNETP (Sustainable Nuclear Energy Technology Platform). In NULIFE, the joint EU-wide coordinated research strategy for plant life integrity management and long-term operation has been defined.Mapping exercise of expertises performed under NULIFE confirmed that NULIFE R&D resources are versatile and high quality. In addition to the wide range of technical expertise available, these are widely spread at geographical and organizational level. Four expert groups, with identified members and links to national programmes, have now produced state-of-the-art type reports related to their expertises. Stress corrosion cracking and thermal fatigue pilot projects have finished concluding reports. Several project proposals have been introduced and optimised for new NULIFE pilot projects or other R&D projects.Based on NULIFE business plan, the discussion of long-term business plan, operational model and statute of the future NULIFE institute has been started. NULIFE maintains the sustainability of nuclear power by focusing on the continued, 60+ years of safe operation of nuclear power plants.     

Design development and analytical assessment of LLCB TBM in Russian Federation during 2012–2013

The activity on the design, analysis, and R&D for the test blanket module (TBM) with lead–lithium (LL) eutectic coolant and ceramic breeder (CB) was performed in the Russian Federation (RF) according to the technical program of cooperation between the leading research institutes of India (“leader” of the LLCB TBM concept) and RF (“partner”). During the period of 2012–2013, the joint efforts of the RF and Indian specialists were focused on the development of the TBM's basic design with an optimal set of parameters (in particular for testing on both H-H and H-D operation phases of International Thermonuclear Experimental Reactor (ITER) machine). This article briefly describes the results of the TBM design and analysis that have been obtained by the RF specialists (“NIKIET” and D.V. Efremov Institute) in support of the LLCB concept (both DEMO blanket and TBM itself). The main directions of this activity in RF institutes were as follows:

• –development of the TBM design taking into account the ability to manufacture the TBM elements (load-bearing casing, tritium-breeding zone, and attachment system);
• –thermal analysis (in both stationary and transient approaches) of TBM design options (four variations of helium and eutectic flowing directions);
• –structural analysis of TBM design elements for Inductive I operation mode; and
• –recommendations (based upon the results of comparative analysis) on the reference design to be used on further stages of concept development.

The critical issues and further plans on the development of LLCB TBM and corresponding DEMO blanket in the RF are also presented in this article.     

Safety Approach and R&D Program for Future French Sodium-Cooled Fast Reactors

This paper presents briefly the safety approach as well as the R&D program that is underway to support the deployment of future French Sodium-Cooled fast Reactors (SFRs): A) Safety objectives and principles for future reactors. The content of the first section reflects the works of AREVA, CEA, and EDF concerning the safety orientations for the future reactors. The availability of such orientations and requirements for the SFRs has to allow introducing and managing the process that will lead to the detailed definition of the safety approach, to the selection of the corresponding safety options, and to the identification and motivation of the supporting R&D. B) Strategy and roadmap in support of the R&D for future SFRs. This section describes the R&D program led jointly by CEA, EDF, and AREVA, which has been developed with the objectives to be able to preliminarily define, by 2012, the safety orientations for the future SFRs, and to deduce from them the characteristics of the ASTRID prototype.