DEMO Fusion Reactor Research Articles

Feasibility study on mass-production of oxide dispersion strengthened Cu alloys for the divertor of DEMO fusion reactor

Oxide dispersion strengthened Cu (ODS Cu) alloys possess a high potential to be used as promising heat sink materials for the divertor system in DEMO fusion reactors. Considering their future application in fusion reactor divertor systems, mass-production of ODS Cu alloys is a must. The purpose of this study is to investigate the feasibility of large-scale production of ODS Cu alloys by improving the production process based on the already established mass-production process of ODS Fe alloys in cooperation with the steel manufacturer Kobelco Research Institute. After optimization of the mechanical alloying (MA) parameters, a large-scale production of ODS Cu powders was successfully achieved with a recovery rate of over 90% and around 1Kg powder in one batch production using an industry-level attritor ball mill cooled by circulating water. The mass-produced ODS Cu alloys were comprehensively characterized, including mechanical properties, microstructure and thermal diffusivity, and a new concept of low-energy ball milling was proposed based on the results to realize the mass-production of ODS Cu alloys.

Neutron irradiation-induced shutdown dose rate estimation in EU DEMO divertor

The EU DEMO fusion reactor, which is part of the EUROfusion PPPT program and the successor to ITER, aims to facilitate the transition of fusion from the science laboratory to industrial use. Its key objective is net energy production. Its critical component, the divertor, manages power exhaust, ash, and helium removal. The key elements of the divertor include Plasma Facing Components (PFCs), shielding liners, reflector plates, divertor cassettes, and a vacuum vessel locking system. It is imperative that the shutdown dose rates are assessed to ensure workers' radiation protection during maintenance, plan the disposal of radioactive material, and ensure the overall safety of the plant. The shutdown dose rate tool employed in the present study is R2Smesh, which couples MCNP for radiation transport and FISPACT for neutron activation to calculate neutron-induced shutdown dose rates, utilizing an advanced, semi-heterogeneous EU DEMO divertor model, presenting a challenging task in neutron transport calculations.

ACABLoop simulation tool: Improving the activation prediction of flowing PbLi alloy in support of DEMO fusion reactor design

ACABLoop simulation tool: Improving the activation prediction of flowing PbLi alloy in support of DEMO fusion reactor design

Study of the production of radioisotopes at IFMIF-DONES: 177Lu with deuterons

The production of radioisotopes for nuclear medicine is boosted by different international agencies. The International Fusion Materials Irradiation Facility - Demo Oriented NEutron Source (IFMIF-DONES) will be a infrastructure for testing under high neutron flux the materials to be used in future DEMO fusion reactor. IFMIF-DONES will generate neutrons by means of the impact of a high-power deuteron beam onto a lithium-jet target. There is an important effort to take advantage of the outstanding characteristics of the facility in terms of neutrons and deuterons. Thus, a complementary program is being developed, where the production of radioisotopes for medicine is one of the main applications. Here, we discuss the production of 177Lu with deuterons at IFMIF-DONES.

Overview of progress on water cooled ceramic breeder blanket in Japan

National institutes for Quantum Science and Technology (QST) are the implementing body of ITER project in Japan and takes the lead in fusion DEMO reactor development in Japan. The primary candidate of the breeder blanket for JA DEMO is Water Cooled Ceramic Breeder (WCCB) blanket. ITER-Test Blanket Module (TBM) program is an opportunity for the demonstration of DEMO breeding blanket concept under actual fusion environment. QST has led the development of the WCCB TBM Concept in the program since the signing of the TBM Arrangement in 2014. The preliminary design of test blanket system (TBS) is on-going. Recently, the status of the preliminary design of TBS was assessed between QST and the ITER Organization (IO). So, the status of the latest TBS design and the DEMO relevancy is presented.

Water activation products generation and transport in DEMO divertor

In water-cooled nuclear reactors, the issue of neutron-activated products transport along the primary heat transfer system (PHTS) is very demanding, as it is a coupled neutronic/fluid-dynamic problem requiring a challenging balance between accuracy and reasonable computational time. This work addresses the transport of water activation products in large hydraulic circuits. Regarding the nuclear calculations, the assessment of the production rates of the radioisotope concentrations has been performed by Monte Carlo analyses adopting the MCNP5.1.6 code, while for the transportation calculations, an innovative method has been expressly developed. It foresees a one-dimensional nodalization, in a MATLAB-Simulink environment, of the hydraulic circuit considered with a computational fluid-dynamic (CFD) characterization (by ANSYS CFX code) of the nodes under neutron flux, that is the components where radioisotopes are formed, and the highest gradients of concentration are present. The method was compared with one-dimensional models not supported by fluid-dynamic analysis. The results of this comparison showed that in cases involving fairly complicated geometries and radioisotopes with a small half-life, CFD analyses are necessary to achieve adequate accuracy. The procedure was applied to very large and rather complex hydraulic circuits like the divertor PHTSs of DEMO fusion reactor to obtain the concentrations of the activation products of the water constituents (16N, 17N, 19O, 14C, 41Ar) along such systems.

Conceptual design of poloidal horseshoe limiter layout for JA DEMO

In the development of the JA DEMO fusion reactor, the proposed poloidal horseshoe limiter aims to mitigate the surface heat load localized on the breeding blanket. The limiter protrudes from the first wall toward the plasma to shade the blanket from the charged particles, which account for the majority of the surface heat load. To concretize the limiter design, this study presents a limiter layout that implements a concept that allows for sufficient cooling and remote maintenance. The limiter shape was determined by evaluating the surface heat load using 3-D magnetic field line tracing. A fully poloidal limiter with a width of 300 mm, which corresponds to 2.4 % of the FW area, was selected. The limiter heating, consisting of surface heat load and nuclear heating, was evaluated to discuss the required cooling, and the concept of a limiter cassette and modules for cooling was proposed. For the limiter replacement method, the outboard limiter is designed to be extracted from the upper port with the mounting segment. The inboard limiter is then grasped and transported to reach the upper port after the outboard limiter extraction. Both the inboard and outboard limiter are expected to be replaced without affecting other in-vessel components in this study.

Due consideration of the breakup and direct reaction mechanisms within (d, p), (d, 2p), (d, xn2p), and (d, xn) reactions

Suitable account of available excitation-function of deuterons interaction with target nuclei components of candidates materials for the ITER fusion reactor, the European DEMO fusion reactor, and the IFMIF-DONES Irradiation Facility has been proved by consistent analysis of the reaction mechanisms involved in the complex deuteron-nucleus interaction. In this work the attention has been focused on the analysis of the deuteron activation cross sections related to the gas accumulation, (d, p), (d, 2p), (d, xn2p), and to the strong neutron emission, (d, xn), interaction processes of interest for the radiation damage and shielding design studies devoted to the structural materials selection. The key role of direct interactions, i.e., breakup, stripping and pick-up processes is stressed out by the comparison of data with theoretical and evaluation predictions.

Thermal and structural analysis of the European water-cooled lithium lead breeding blanket concept in the L-H mode transition

The design of a breeding blanket system (BB) which reliably fulfils the prescribed requirements is pivotal for the construction of the EU DEMO fusion reactor. For this reason, the BB conceptual design activities require a tighter interaction with plasma physics modelling, so to employ relevant boundary conditions to compare, and then down-select, the BB design alternatives. The preliminary thermo-mechanical study herein reported highlights the importance to assess in deep the BB performance during plasma operational transients. In particular, analyses of the BB thermal and mechanical transients during plasma ramp-up have been carried out, with focus on the L-H transition. Attention has been paid to the Water-Cooled Lithium Lead (WCLL) BB concept, one of the two driver blanket candidates for the EU DEMO project. Adopting DEMO-relevant plasma configurations, the fusion power and the radiative power time evolutions during the plasma ramp-up phase have been modelled under different assumptions using the code ASTRA. The obtained power transients have been used to calculate the time-dependent nuclear power density and radiative heat flux the WCLL BB is subjected to, in order to perform the corresponding transient thermal analysis. Then, the structural analysis has been carried out in order to verify the fulfilment of the RCC-MRx structural design criteria in the most critical time steps. The performed thermo-mechanical analysis has been carried out adopting the Abaqus FEM code. Results obtained are presented and critically discussed, highlighting their importance for the EU DEMO integrated design.

Numerical and experimental development of cupronickel filler brazed joints for divertor and first wall components in DEMO fusion reactor

The brazeability of a cupronickel commercial alloy (Cu10Ni) was evaluated for its use as a filler alloy for high-temperature joining of tungsten to the reduced activation ferritic/martensitic steel EUROFER 97 (W-E) and between tungsten base materials (W-W) for its application at the first wall and divertor of future fusion reactors. In addition, given the importance of the residual stresses in these heterogenous joints, a study of the brazing conditions and the impact of the selected filler has been conducted using numerical software to understand its impact on the quality of the joint.Two thermal cycles were evaluated (1165 °C and 1190 °C) and selected based on the thermal characterization of the filler alloy. The microstructural examination revealed that, in W-E joints, nickel acts as an activator element, reacting and forming interfacial layers at the EUROFER 97 - Cu10Ni interface. In the case of the W-W joints, a lower level of diffusion phenomenon and metallurgical interaction between Cu10Ni and base materials were observed. The hardness profile indicated that the hardening process of EUROFER 97 was associated with the formation of untempered martensite. On the other hand, tungsten kept the received hardness. The mechanical characterization by shear test reported similar values between both types of joints carried out at 1190 °C but different when the temperature was increased (1165 °C), associated with the brittle character of tungsten and its lower metallurgical interaction.The numerical analysis of the brazing process carried out with ANSYS software shows that residual stresses are accumulated mainly at the interfaces. The information provided by the simulation shows, for a 50 µm filler thickness, the importance of mitigating the residual stress by selecting a filler with an intermediate Coefficient of Thermal Expansion (CTE) that alleviates mechanical stresses relative to the base materials.

T production using a high-temperature gas-cooled reactor for the DEMO fusion reactor: Li rod structure for the initial irradiation test

This work proposes tritium (T) production using a high-temperature gas-cooled reactor (HTGR) and studies it for the initial T inventory in a demonstration fusion reactor and a prior engineering test with T handling. At this stage, the aim is to investigate the compatibility between electricity and T production, which makes the stable confinement of T in the Li-loading rods during the HTGR operation period a crucial issue. The total T outflow into the helium gas during the reactor operation period was attempted to be reduced using Zr spheres with Ni coating as a T absorption material to avoid an increase in the inner T pressure. The basic hydrogen absorption properties of the Zr spheres with Ni coating, which coexist in an environment with Al2O3 and/or LiAlO2 oxides, were measured. The structures of the Li-loading rod for a typical commercial HTGR and irradiation test were studied using the obtained data. In addition, an outline of the irradiation test is reported.

Ab initio study of hydrogen in titanium beryllides

Titanium beryllide Be12Ti is a candidate material for the neutron multiplier for the demonstration fusion reactor DEMO. Experimental studies show that under certain conditions, Be12Ti may contain inclusions of other phases such as Be2Ti, Be17Ti2. In this regard, it is important to study the behavior of tritium and its isotopes in the crystal lattices of these phases. All calculations are performed using ab initio methods. Solution energies of a hydrogen atom in all non-equivalent interstitial sites of the three studied titanium beryllides were found to be lower than that in pure beryllium. The formation energy of all types of vacancies in all studied beryllides is found to be higher than that in beryllium. The binding energies of a single hydrogen atom located both inside and outside the vacancies are calculated. Hydrogen inside monovacancy is more strongly bound as compared to that outside this vacancy. It turned out that in some cases hydrogen can be captured by vacancy being outside of it. The results obtained are essential for further study of interstitial diffusion of hydrogen and analysis of tritium retention in titanium beryllides.

Procedure qualification and integration of the continuous external Rogowski diagnostics to the superconducting magnets of ITER tokamak

Continuous External Rogowski (CER) are the primary plasma current diagnostics for the ITER tokamak, as well as the only diagnostics built into the ITER Toroidal Field Coils (TF coils). The integration of CER diagnostics to the TF coils faced numerous technical challenges, such as: the integrated components have to withstand the maximum pressure foreseen during ITER TF coil resin impregnation (350 kPa), comply with the high vacuum requirements that the helium leak rate shall be <2.7 × 10−9 Pa-m3/s, and survive at cryogenic temperature. The CER diagnostics integration campaign turned out to be a marked success, thanks to a comprehensive pre-qualification program and well-coordinated collaborations among the participating parties. All four sets of CER diagnostics have so far been integrated to the respective TF coils. As a pioneering attempt of building plasma diagnostics into reactor-scale superconducting magnets, the potential implications of this venture to the diagnostics development for DEMO and future fusion reactors are discussed. The gained experiences and lessons learned from this venture are summarized at last.

Exploratory Thermo-Mechanical Assessment of the Bottom Cap Region of the EU DEMO Water-Cooled Lead Lithium Central Outboard Blanket Segment

The Water-Cooled Lead Lithium (WCLL) Breeding Blanket (BB) is one of the two BB concept candidates to be selected as the driver blanket for the EU DEMO fusion reactor. In this regard, the development of a sound architecture of the WCLL Central Outboard Blanket (COB) Segment, ensuring the fulfilment of the thermal and structural design requirements, is one of the main goals of the EUROfusion consortium. To this purpose, an exploratory research campaign has been launched to preliminarily investigate the thermo-mechanical performances of the Bottom Cap (BC) region of the WCLL COB segment because of its peculiarities making its design different from the other regions. The assessment has been carried out considering the nominal BB operating conditions, the Normal Operation (NO) scenario, as well as a steady-state scenario derived from the in-box LOCA accident, the Over-Pressurization (OP) scenario. Starting from the reference geometric layout of the WCLL COB BC region, a first set of analyses has been launched in order to evaluate its structural performances under a previously calculated thermal field and to select potential geometric improvements. Then, the analysis of a complete BC region was conducted from both the thermal and structural standpoints, evaluating its structural behaviour in compliance with the RCC-MRx code. Finally, after some iterations and geometric updates, a promising geometric layout of the BC region has been obtained even though some criticalities still persist in the internal Stiffening Plates and First Wall. However, the obtained results clearly showed that the proposed layout is worthy to be further assessed to achieve a robust enough configuration. The work has been performed following a theoretical-numerical approach based on the Finite Element Method (FEM) and adopting the quoted Ansys commercial FEM code.

European DEMO Fusion Reactor: Design and Integration of the Breeding Blanket Feeding Pipes

This article describes the design and configuration of the DEMO Breeding Blanket (BB) feeding pipes inside the upper port. As large BB segments require periodic replacement via the upper vertical ports, the space inside the upper port needs to be maximized. At the same time, the size of the upper port is constrained by the available space in between the toroidal field coils and the required space to integrate a thermal shield between the vacuum vessel (VV) port and the coils. The BB feeding pipes inside the vertical port need to be removed prior to BB maintenance, as they obstruct the removal kinematics. Since they are connected to the BB segments on the top and far from their vertical support on the bottom, the pipes need to be sufficiently flexible to allow for the thermal expansion of the BB segments and the pipes themselves. The optimization and verification of these BB pipes inside the upper port design are critical aspects in the development of DEMO. This article presents the chosen pipe configuration for both BB concepts considered for DEMO (helium- and water-cooled) and their structural verification for some of the most relevant thermal conditions. A 3D model of the pipes forest, both for the Helium-Cooled Pebble Bed (HCPB) and Water-Cooled Lithium Lead (WCLL) concepts, has been developed and integrated inside the DEMO Upper Port (UP), Upper Port Ring Channel, and Upper Port Annex (UPA). A preliminary structural analysis of the pipeline was carried out to check the structural integrity of the pipes, their flexibility against the thermal load, their internal pressure, and the deflection induced by the thermal expansion of the BB segments. The results showed that the secondary stress on the hot leg of the HCPB pipeline was above the limit, suggesting future improvements in its shape to increase the flexibility. Moreover, the WCLL concept did not have a critical point in terms of the secondary stress on the pipeline, since the thicknesses and the diameters of these pipes were smaller than the HCPB ones.

Neutronic Activity for Development of the Promising Alternative Water-Cooled DEMO Concepts

An emerging breeding blanket that fulfills performance criteria, meets the safety requirements, and is reliable enough to meet the plant availability is a challenging issue that assumes complex studies involving numerous neutronic analyses based on the Monte Carlo simulations with MCNP code. Two different concepts are now candidates to be implemented as a driver blanket for DEMO fusion reactor: WCLL (Water-Cooled Lithium Lead) and HCPB (Helium-Cooled Pebble Bed). The current R&amp;D work within the EUROfusion DEMO project is concentrated on a search for the new water-cooled blanket layouts: a deep upgrade of the WCLL blanket to ensure a sufficient tritium breeding capability and an elaboration of the hybrid concept coupling technological advantages of water coolant, lead neutron multiplier, and ceramic breeder. To this end, very detailed, fully heterogeneous MCNP geometry models were developed for the newest designs of the WCLL-db (WCLL-double bundle) and WLCB (Water-cooled liquid Lead Ceramic Breeder) DEMO blankets to verify the new engineering solutions. This makes rigorous calculations possible to find an optimal breeder blanket layout. The basic response, tritium breeding ratio (TBR), was assessed for both concepts, and it appeared to be TBR = 1.16 for the WCLL-db and TBR ≤ 1.13 for the WLCB DEMOs, respectively. Several geometry layouts of the WLCB breeder blanket were investigated to reach the TBR sufficient for a sustainable tritium fuel cycle. Two promising novel solutions were suggested to enhance the tritium breeding performance of the WLCB blanket and to achieve TBR ≥ 1.16: heavy water coolant and an advanced breeder ceramic. Various nuclear safety aspects of the technologies utilized in both blanket concepts are addressed in this work to facilitate engineering decisions aimed at the consolidated blanket design for the upcoming DEMO reactor.

Compatibility of tritium breeders and neutron multipliers as a mixed-pebble packing concept for JA DEMO fusion reactor

To verify the feasibility of a mixed-pebbles packing concept, the high-temperature compatibility of beryllide (Be12V) and Li2TiO3 pebbles with 1mm diameter at high temperatures, 973 and 1173 K was investigated. It was evident from the x-ray diffraction and scanning electron microscopy observations that the Li2TiO3 pebbles exhibited no remarkable change after the compatibility tests at 973 and 1173 K. However, inductively coupled plasma mass spectrometry analysis revealed that Be12V pebbles contained a small amount of Li as an impurity, revealing that as time increased, the Li amount in the Be12V pebbles increased with parabolic reaction after the compatibility tests. Moreover, the contact number and area were evaluated to be 24 points and 38.61 μm2, respectively by assumption that Be12V and Li2TiO3 with approximately 2.4 and 0.6 mm in diameter loaded.

Structural material nuclear data basic research

The nuclear data evaluation for deuteron-induced reactions andα-particle emission by neutron interactions is addressed within “Nuclear data for fusion technology, from basic research to full-scale applications.” The status and open questions related to these subjects in the area of nuclear data for fusion technology, specifically for the nuclear design of the ITER fusion device, the European DEMO fusion reactor, and the IFMIF-DONES Irradiation Facility, are briefly reviewed. A firm demand for accurate cross-sections of reactions induced by neutrons and deuterons exists, in this respect, within a more enlarged energy range up to 50 MeV than for fission applications. The current requirements are closely met by the TENDL Evaluated Nuclear Data Library, settled using the TALYS nuclear model code, which is one of the most widely used codes in basic research and applications including nuclear fusion technology. However, further improvement of this data library has recently been suggested, while, with respect to fission applications, not only the aforementioned energy range but also the diversity of nuclear data for fusion technologies is plainly stretched. Consequently, the progress of nuclear data activities conducted more recently on deuteron-induced reactions andα-emission by neutron interactions, throughout the European Fusion Program and subsequent to previous achievements within F4E and EUROfusion programs, is wholly summarized.

Development of a Steam Generator Mock-Up for EU DEMO Fusion Reactor: Conceptual Design and Code Assessment

Recent R&amp;D activities in nuclear fusion have identified the DEMO reactor as the ITER successor, aiming at demonstrating the technical feasibility of fusion plants, along with their commercial exploitation. However, the pulsed operation of the machine causes an “unconventional” operation of the system, posing unique challenges to the functional feasibility of the steam generator, for which it is necessary to define and qualify a reference configuration for DEMO. In order to facilitate the transitions between different operational regimes, the Once Through Steam Generator (OTSG) is considered to be a suitable choice for the DEMO primary heat transfer systems, being characterized by lower thermal inertia with respect to the most common U-tube steam generators. In this framework, the ENEA has undertaken construction of the STEAM facility at Brasimone R.C., aiming at characterizing the behavior of the DEMO OTSG and related water coolant systems in steady-state and transient conditions. A dedicated OTSG mock-up has been conceived and designed, adopting a scaling procedure, keeping the height 1:1 of the DEMO OTSGs. The conceptual design has been supported by RELAP5/Mod3.3 thermal-hydraulic calculations. CFD and FEM codes have been used for fluid-dynamic analyses and mechanical stress analyses, respectively, in specific parts of the component.

The applicability of remote field eddy current testing to outer flaws on cooling tubes in the blanket of a fusion DEMO reactor

This study investigated the applicability of remote field eddy current testing (RFECT) to the non-destructive inspection of cooling tubes in the blanket of a fusion DEMO reactor. Pipes made of F82H steel with inner and outer diameters of 9.0 and 11.0 mm, respectively, were prepared, and artificial grooves were machined on the outer surface of the pipes. The pipes were inspected using an RFECT probe consisting of one exciter and one detector. The results confirmed that RFECT was capable of detecting a groove with a depth of 0.5 mm if the axial length of the groove was 5 mm or more. Subsequently, stress analysis was performed using the three-dimensional finite element method to evaluate the stress caused by the presence of a groove. The calculated stress was compared with the strength of the F82H steel. The stress caused by undetectable flaws was smaller than the strength.