Fusion DEMO Plant Research Articles

Consequences of Neutron Energy Spectrum on Radiation Damage, Gas Production, and Transmutations in Fusion Materials

The inclusion of test modules in the U.S. Fusion Prototypic Neutron Source (FPNS) offers the opportunity to test a wide variety of materials in a representative radiation environment of the fusion Pilot Plant, DEMO, and power plant. The testing may include various generations of structural materials for conventional and advanced blanket and divertor concepts. Since all structural materials derived from the fission industry are inadequate for fusion applications (due to the more damaging effects of the 14-MeV fusion neutrons), radiation-resistant reduced-activation structural materials (reduced-activation ferritic-martensitic steel, vanadium alloy, W alloy, and SiC/SiC composites) were specifically developed for fusion and could be tested in the FPNS to qualify for the highly irradiated fusion components surrounding the plasma. The large atomic displacement and the helium and hydrogen generations by fusion neutrons are unique to fusion materials. The most important attribute for the FPNS would be the typical fusion-relevant He/displacements per atom (dpa) ratio of ~10 for steel in particular. By comparison, irradiation in the fission spectrum of the High Flux Irradiation Facility (HFIR) would underestimate the dpa and provide a very low He/dpa ratio of ~0.3 for steel, which is irrelevant to fusion. This paper reviews the neutron irradiation impacts and presents a few examples of dpa and transmutation products for steel, W, and SiC based on modeling in several fusion design studies. The operating conditions of advanced U.S. fusion power plants were considered along with the credible lifetime goal of 200 dpa and 20 MW·yr/m2 fluence that could be achieved with directed research and development programs coupled with the construction of the FPNS 14-MeV neutron facility.

Passive Hydrogen Recombination during a Beyond Design Basis Accident in a Fusion DEMO Plant

One of the most important environmental and safety concerns in nuclear fusion plants is the confinement of radioactive substances into the reactor buildings during both normal operations and accidental conditions. For this reason, hydrogen build-up and subsequent ignition must be avoided, since the pressure and energy generated may threaten the integrity of the confinement structures, causing the dispersion of radioactive and toxic products toward the public environment. Potentially dangerous sources of hydrogen are related to the exothermal oxidation reactions between steam and plasma-facing components or hot dust, which could occur during accidents such as the in-vessel loss of coolant or a wet bypass. The research of technical solutions to avoid the risk of a hydrogen explosion in large fusion power plants is still in progress. In the safety and environment work package of the EUROfusion consortium, activities are ongoing to study solutions to mitigate the hydrogen explosion risk. The main objective is to preclude the occurrence of flammable gas mixtures. One identified solution could deal with the installation of passive autocatalytic recombiners into the atmosphere of the vacuum vessel pressure suppression system tanks. A model to control the PARs recombination capacity as a function of thermal-hydraulic parameters of suppression tanks has been modeled in MELCOR. This paper aims to test the theoretical effectiveness of the PAR intervention during an in-vessel loss of coolant accident without the intervention of the decay heat removal system for the Water-Cooled LithiumLead concept of EU-DEMO.

Preliminary neutronics analyses of China Dual-Functional Lithium-Lead (DFLL) test blanket module for CFETR

China Fusion Engineering Test Reactor (CFETR), as one of the option fusion reactors in China, has been proposed to fill the gap between ITER and fusion DEMO plant. In order to evaluate the neutronics performance in the framework of National Magnetic Confinement Fusion Science Program of China, the neutronics analyses of the detailed model of CFETR with China dual-functional lithium lead (DFLL) test blanket module (TBM) were carried out with the Super Multi-functional Calculation Program for Nuclear Design Safety Evaluation (SuperMC) and JEFF-3.2 nuclear data library. To meet the self-sufficiency target of CFETR, the tritium breeding ratio has been improved to 1.2. Meanwhile, the nuclear heating result showed a potential for energy multiplication factor improvement. From this previous study, an optimized model was proposed to improve the shielding performance and the results showed that the DFLL-TBM with the optimized model could meet the tritium self-sufficiency target and radiation design requirements.

Historical Fission Plant Availability and the Operation Goal for a Fusion DEMO

As fusion researchers look toward the future, there have been discussions on what the plant operation goals should be for a demonstration fusion power plant (DEMO). The U.S. Research Needs Workshop (ReNeW) in 2009 stated that power producers (the companies owning power plants) could not expect an ultimate fusion power plant availability of 80% or more if a DEMO reactor cannot demonstrate a 50% or higher availability. The ReNeW panel also stated that achieving 50% availability with a DEMO plant would be a huge accomplishment. Other recent DEMO design studies have given goals for plant availability as well. This technical note presents historical plant availability values of new technology fission power plants to compare fission achievements with the suggested goals for fusion DEMO plant designs. Demonstration fission plants that met or exceeded 40% average annual availability were generally considered to be successful. These data help to support the goal values that have been put forward in various studies.

Recent activities on water detritiation technology in JAEA

Japan Atomic Energy Agency has developed the water detritiation technology for CECE process considering the escalation in quantity and tritium concentration of tritiated water to be processed toward realization of the Japanese fusion DEMO plant. A commercial ion exchange perfluoro-membrane for concentrated tritiated water electrolyzer is durable against irradiation up to 1500kGy in view of mechanical strength and ion exchange capacity. Concerning hydrophobic catalyst for LPCE column, a new method of manufacturing catalysts involving hydrophobic processing with an inorganic substance base has been developed. The catalyst created with this method has achieved the highest exchange efficiency, equivalent to 1.3 times the previously most powerful efficiency.

Safety studies on Korean fusion DEMO plant using Integrated Safety Assessment Methodology: Part 2

The purpose of this paper is to investigate safety issues using Integrated Safety Assessment Methodology (ISAM) proposed by Generation IV Forum Risk and Safety Working Group (RSWG) for Korean fusion DEMO plant (K-DEMO). In ongoing nuclear energy research such as Generation IV fission power plant (GEN-IV), new methodology based on Technology-Neutral Framework (TNF) has been applied for safety assessment. In this methodology, design and regulatory requirements for safety of nuclear power plants are considered simultaneously. The design based on regulatory requirements can save resource, time, and manpower while maintaining high level safety. ISAM is one of the options to apply TNF in K-DEMO. We have performed safety studies for K-DEMO using Phenomena Identification and Ranking Table (PIRT) and Objective Provision Tree (OPT) which are constitutive part of ISAM. Considering the design phase of K-DEMO, the current study focused on PIRT and OPT for K-DEMO. Results have been reviewed and updated by Korean fusion advisory group after considering the views of specialists from domestic universities, industries, and national institutes in South Korea.

Preliminary Shielding Design and Analysis of Liquid Lead–Lithium Blanket for Chinese Fusion Engineering Testing Reactor

To fill the gap between ITER and fusion DEMO plant, the conceptual design of Chinese Fusion Engineering Testing Reactor (CFETR) has been proposed. In this study, the shielding blanket of liquid lead–lithium blanket for CFETR was designed with newly-developed three dimensional (3D) models. The detailed shielding scheme was obtained with the 3D CFETR model to optimize the shielding parameters such as the shielding material and size, by using the Super Monte Carlo Simulation Program for Nuclear and Radiation Process. The radiation limits of fast neutron flux and nuclear heating density in the Toroidal Field Coils were considered and with the nuclear analysis for the 3D model, the obtained shielding scheme could meet the requirement of radiation limits.

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.

Fusion Licensing and Safety Studies in Korea

Korean activities on fusion licensing and safety studies have been going through three phases. The first phase started with the licensing of the Korean Superconductor Tokamak Advanced Research (KSTAR) classified as a radiation-generating device on the basis of the existing nuclear regulations. To accommodate the specific characteristics of a tokamak device that is to use deuterium in fusion plasma physics studies, a set of instructions to examine the safety of KSTAR and to approve an operating permit were developed and issued by the regulatory body, Korean Institute of Nuclear Safety. The second phase is being investigated for the International Thermonuclear Experimental Reactor (ITER). To verify the design of the test blanket module that will be installed and tested in the ITER, safety studies were carried out using the experimental results and computational codes. To minimize occupational irradiation damages, a study on the tritium behaviors is being carried out as Korea will provide a part of the tritium systems of ITER. The last phase has initiated to recognize significant safety issues and to support future licensing process for a fusion DEMO plant. Relevant studies are being carried out in collaboration with the academies funded by the National Fusion Research Institute of Korea. These front-end studies include the conceptual analysis on the safety and radiological consequences, preliminary studies on the layout of research, and on the future regulatory requirements of the fusion DEMO plant. This paper summarizes these activities on fusion safety with the findings of the front-end studies to date. It also proposes a pathway for the novel licensing process of a fusion DEMO plant with the tokamak concept.

A Study on the Site Plot Plan and Building Schematics of a Fusion DEMO Plant

To expedite realization of magnetic fusion energy with the tokamak concept, a parallel process of developing engineering technologies required to design, fabricate, construct, start up, and operate the fusion DEMO plant of Korea (K-DEMO Plant) at the same time with researching fusion science and technologies in KSTAR (Korean Superconductor Tokamak Advanced Research) and ITER (International Thermonuclear Experimental Reactor) was adopted in the fusion DEMO program of Korea (K-DEMO Program). As a part of these engineering studies, an exploratory study on the layout and building schematics of K-DEMO Plant have carried out in consideration of economic and safety aspects. The buildings of K-DEMO Plant are named and their volumes are estimated with an order of magnitude analysis based on the sizes of the buildings of ITER and nuclear power plants. This exploratory study on the layout and building schematics is also required to estimate the costs of K-DEMO Program and analyze its economic feasibility.

Study on containment safety systems for Korean fusion DEMO plant

The purpose of this study is to review the strategy for radiation barriers in the fusion power plants and to produce simulation data for the conceptual design of safety features to maintain the integrity of such barriers as a part of R&D program through the National Fusion Research Institute of Korea. Even though the amount of radioactive source term in fusion power plants should be much less than that of fission power plants, internal as well as external events can result in damage to facilities such that public can be critically exposed by radiation. In the first part of this study, we reviewed and compared the multiple defenses to protect radioactive hazard in fission and fusion power plants. Containment was characterized as an indispensable physical barrier and the integrity of containment particularly enveloping a fuel cycle which is a major radioactive source term, tritium, should be secured. Since water is assumed as one of the coolant options in the Korean fusion DEMO plant, the thermo-hydraulic analysis was carried out using computer simulations to produce key parameters related with the integrity of containment in the second part. The performance of both of active and passive safety features to control the key parameters was compared to take recent fission technologies into account.

Safety Classifications for the Fusion DEMO Plant of Korea

Regulatory requirements and industrial codes and standards (C&S) are to be compulsory top-tier prerequisites for the design and construction of the fusion DEMO plant of Korea (DEMO Plant) as it will have safety-related functions to mitigate radiation hazards to the environment. However, these prerequisites will be developed in parallel with the design and construction of DEMO Plant due to the first-of-a-kind nature of the DEMO Plant. To overcome this dilemma, a tentative set of general design criteria and classification criteria that are to be applicable for the design studies of DEMO Plant is proposed, making use of the commonalities between the DEMO Plant and existing nuclear power plants (NPPs). To determine the tentative set, the radiation release to the environment and consequences of the postulated events in the DEMO Plant are analyzed with rough approximation assuming the inherent safety of a fusion DEMO reactor. The classification criteria of the tentative set include safety class, IEEE electrical classes, quality groups, and seismic categories for the structures, systems, and components of the DEMO Plant. As it will take considerable time and substantial amounts of investment to develop the regulatory requirements and C&S, the use of ASME Section III for the design study of DEMO Plant is also proposed to make the best use of the commonalities between the DEMO Plant and existing NPPs.

Cross-Cutting Ideas for a Fusion DEMO Plant with Current and Generation IV Nuclear Power Plants

The fusion DEMO program of Korea has been conceptualized to realize magnetic fusion energy with the tokamak concept at the end of 2030s or early 2040s. In this program, to expedite the development of a fusion DEMO plant, cross-cutting based on the commonalities between the fusion DEMO plant and existing systems. Among the existing systems, the current and generation IV nuclear power plants will have many areas of commonalities with the fusion DEMO plant including regulatory requirements and licensing processes, codes and standards, design methods and computational codes for thermo hydraulic analysis, and safety analysis methods. Theses commonalities will be used for discovering a pathway of resolving the nested logic dilemma incurred by the inherent first-of-a-kind nature of the fusion DEMO plant. This paper presents the result of an exploratory study on the subject cross-cutting.

Design failure mode and effect analysis for Korean fusion DEMO plant

The construction of Korean Fusion DEMO Plant (KFDP) for a demonstration of the plasma analysis and engineering feasibilities is planned in 2030s based on the Korean fusion technology roadmap. The radiation safety should be assured for nuclear facilities, so that, the KFDP is required to research for the regulatory requirements and industrial codes and standards. The final design guidance of the engineered safety features should be served in future. As the first step for this research, the failure modes and effects analysis in a design stage was performed. This leads to find the list of potential hazard elements and to obtain the list of initiating events for the future probabilistic risk assessment. The hazard elements expected to seriously threaten the integrity of the KFDP were investigated and determined to quantify the effect of the initiating events in the effect analysis: (1) a total loss of active cooling water to occur during the burn with decay heat calculation and (2) coolant ingress from cooling circuits into the vacuum vessel, cryostat and containment building. For those initiating events, the quantitative simulations using transient mass and energy calculation and computational fluid dynamics were performed.

Strategic Plans for the Fusion DEMO Program of Korea

The strategic plans for the Korean fusion DEMO program of which the total investment and duration are estimated to be 5 to 11 billion US dollars and around 27 years respectively were developed by adopting various practical skills of the business management and project management disciplines including gap study, analytic hierarchy process and work breakdown structure.The vision, mission and strategic objectives of the program were set in line with the national policies on the promotion of the fusion energy development. The key strategies and strategic initiatives of the program were derived from SWOT (Strengths, Weaknesses, Opportunities and Threats) and matched pair analysis. The implementing measures for the strategic initiatives were studied and the critical success factors for the program were also defined.To facilitate the R&D portfolio management, the Program was divided into three sub-programs in accordance with the three phases of the Program.The R&D activities and facilities of the 2nd phase program were pulled from the final product of the program that is the fusion DEMO Plant by carrying out a gap study. The implementation plans including the estimated costs, human resource plans and timelines are also prepared and introduced.

Recent activities related to the development of the plasma facing components for the ITER and fusion DEMO plant

In the ITER procurement, the Japanese Domestic Agency (JADA) will procure the divertor outer vertical target (OVT). The qualification of the manufacture of the components has been started by validating JADA's technical capability. JADA has developed vertical target qualification prototypes that cover most of the critical technical issues in the series production. The prototypes have been high heat flux tested under effective coordination by the ITER organization (IO) and showed sufficient durability. JADA has successfully obtained the certification to conduct the procurement of the OVT by the IO. Development of a breeding blanket is one of the most important issues to realize the DEMO. Test blanket module (TBM) testing is a key milestone toward the DEMO. In JAEA, R&D on the water-cooled blanket has been performed. As a result, a full-scale TBM first wall mock-up has successfully been developed. This mock-up showed sound thermal performance in preliminary testing.

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/m 2 is performed recently. Three sets parameters of core plasma for different DEMO design objectives are proposed. A helium-cooled blanket system with ceramic breeder (Li 4SiO 4), 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.

Torus configuration and materials selection on a fusion DEMO reactor, SlimCS

SlimCS is the conceptual design of a compact fusion DEMO plant assuming technologies foreseeable in 2020–2030s. For continuity of blanket technology from the Japanese ITER-TBM, the prime option of blanket is water-cooled solid breeder with Li 2TiO 3 (or Li 4SiO 4) and Be. A reduced-activation ferritic–martensitic steel (RAFM) and subcritical water are chosen as the structural material and coolant, respectively. The reactor has somewhat complex torus configuration with a sector-wide conducting plate slipped in between the replaceable (front) and permanent (back) blanket. In order to allow flexibility of maintenance in such a complex configuration, sector transport hot cell maintenance scheme is adopted. This paper describes characteristics of SlimCS with a focus on materials selection.

Evaluation of operation scenario for fusion DEMO plant at JAEA — Constraint of neutral beam injection system

The ramp-up scenario of the plasma current ( I p) without the center solenoid (CS) coil in the case of a DEMO plant at JAEA is evaluated using the zero-dimensional system code. The scenario is estimated from low I p (2 MA) to the final fusion reaction during the period of neutral beam injection (NBI). The beam energy ( E NBI) of NBI is required to be changed using new constant current variable voltage (CCVV) technology more than two times during the plasma current ramp-up. The constraints of NBI power ( P NBI) are evaluated from the plasma requirements and use of CCVV technology. Time evolutions of plasma parameters are evaluated using the circuit equation with zero-dimensional particle and power balance equations. The final state with a fusion output of 3 GW from low I p (2 MA) can be reached by changing E NBI using CCVV technology. During the plasma current ramp-up phase, NBI power is needed to be larger (90 MW) than that in a flattop. It takes 2.5 h to reach the final state.

Consideration on blanket structure for fusion DEMO plant at JAERI

Our idea on the DEMO plant is that it must demonstrate (1) an electric power generation of one GW level, (2) self-sufficiency of tritium fuel (TBR is more than 1.05), (3) year-long continuous operation, etc. At the same time, DEMO is expected to use technologies to be proven by 2020 and present an economical prospect of fusion energy in the operational time of the reactor. The design guidelines for the blanket are defined in order to meet the mission of the DEMO plant as mentioned above. Major design conditions are surface heat flux of 0.5 MW/m 2 with peaking factor of 2, a neutron wall load of 3.5 MW/m 2 with peaking factor of 1.5 and a neutron fluence of about 10 MW/m 2. To moderate the technological extrapolation, reduced activation ferritic steel (F82H) structural material, Li 2TiO 3 and Be neutron multiplier are considered. To improve the economical aspect, supercritical water with inlet/outlet temperatures of 280/510 °C is chosen as coolant material, with coolant pressure of 25 MPa. As a result, a thermal efficiency of 41% is achieved. To obtain higher plasma performance, MHD instabilities suppressing shell structure is adopted with structural compatibility to the blanket structure. To meet higher plant availability requirements (more than 75%), the hot cell maintenance approach is selected for the replaceable power core components.