Demonstration Fusion Reactor Research Articles

Compression properties of beryllide pebbles at high temperatures

DEMO (demonstration) fusion reactors require advanced neutron multipliers with low swelling and high stability at high temperatures. Research on beryllium intermetallic compounds (beryllides (Be12V)) has been conducted, and the authors have successfully fabricated Be12V and Be12Ti0.3V0.7 ternary pebbles, which are considerably stable at high temperatures. In this study, to evaluate the mechanical properties of these pebbles, compression tests for a single pebble were performed at room temperature and high temperatures of 600 and 800 K. Be12Ti0.3V0.7 pebbles exhibited a higher crush load than Be12V pebbles; Be12Ti0.3V0.7 pebbles comprised remarkably fine grains due to dual phases. Scanning electron microscopy images of the fractured area demonstrated that the Be12V pebbles had a relatively large fracture with sharp edges, whereas the Be12Ti0.3V0.7 pebbles included considerably small grains. There was no noticeable temperature dependence of the crush load. The fracture and tensile strengths of Be12V and Be12Ti0.3 V0.7 pebbles were evaluated.

Development of a thermal-hydraulic model of the EU-DEMO Water Cooled Lithium Lead Breeding Blanket Primary Heat Transport System

The EUROfusion consortium is developing the project of a DEMOnstration Fusion Reactor (EU-DEMO) which would follow ITER in the pathway towards the quest for the exploitation of fusion energy. EU-DEMO has been conceived to deliver net electric power to the grid. Therefore, proper critical evaluations of the tokamak cooling and power conversion systems are needed because they play a pivotal role in the design and licencing of the overall plant. The EU-DEMO reactor will be based on the tokamak concept and, as such, it is supposed to undergo a pulsed duty cycle under normal conditions, which might challenge the qualified lifetime of the main equipment inducing undue thermal and mechanical cycling. Moreover, the EU-DEMO plasma control strategy postulates the possible occurrence of planned and off-normal plasma overpower transients that might jeopardise the structural integrity of the plasma facing components. It is, therefore, of paramount importance to have appropriate tools to reproduce the thermal-hydraulic behaviour of tokamak cooling systems during major operational and accidental scenarios in a realistic and reliable way. In this context, University of Palermo in cooperation with EUROfusion has developed a finite volume model of the Primary Heat Transport System (PHTS) feeding the EU-DEMO Water Cooled Lithium Lead Breeding Blanket (WCLL BB). The activity has been led following a theoretical–computational approach based on the adoption of the TRACE thermal-hydraulic system code. Particular attention has been paid to capturing all the main geometrical, hydraulic and heat transfer features characterising both in-vessel and ex-vessel components. Preliminary analyses have also been carried out to check the code’s predictive potential in fusion relevant applications. Models, assumptions, and outcomes of the analyses are herewith reported and critically discussed.

Dose rate analysis for a pre-conceptual design of the high energy beam transport section in the IBTF

A pre-conceptual design study to evaluate the Breeding Blanket (BB) performance has been carried out because the BB, producing heat energy and tritium, is a key component for the fusion demonstration reactor. The Integrated Breeding Test Facility (IBTF) is based on an accelerator-driven system using a 40 MeV deuteron beam with a maximum current of 10 mA. Accelerated deuteron reacts with a solid beryllium target to generate neutrons. The total yield of the produced neutrons is estimated to be 6.68E+15 n/sec. The forward direction neutrons irradiate the tritium breeding unit, while the backscattered neutrons through the beam pipes activate the Deuteron Accelerator Unit (DAU) components that require temporary human access maintenance.Therefore, this study did a dose rate analysis for the pre-conceptual design of the High Energy Beam Transport (HEBT) section in the DAU during maintenance. The HEBT section is divided into regions A and B. For the reference HEBT model, the dose rate in only region B satisfies the human access requirement (dose rate ≤ 100 uSv/h) during short-term maintenance (STM) and long-term maintenance (LTM). To reduce the dose rate, the beam pipe material was changed to Al6063, which has a low weight fraction of Mn and Ni compared to SS304. A lead shutter was also used to shield the dose rate from the test cell. As a result, the dose rate in region A is lower than 100 uSv/h during the STM with a cooling time of 1 week. The dose rate in region B satisfies the hands-on maintenance requirement (dose rate ≤ 10 uSv/h). The dose rate in region A is lower than 100 uSv/h except for near the beam pipe during the LTM. The dose rate in region B is lower than 10 uSv/h. The dose rate in the HEBT section is increased as the LTM is repeated. However, the increasing rate of the dose is reduced as the LTM repeats. Neutron transport calculation was performed by SRC-UNED-v3.1.4 with the TENDL2019 cross-section library, and the gamma source was estimated using FISPACT-Ⅱ with the TENDL 2017 cross-section library.

Nuclear fusion waste classification of K-DEMO under the Korean radioactive waste regulations

The nuclear fusion waste generated by the activation of the constitutent components of the Korean fusion DEMOnstration reactor (K-DEMO) is assessed after the end of life (EOL). Based on the results, an attempt is made to classify the radioactive waste (RW) according to the Korean regulations on RW classification and self-disposal standards. One hundred years after EOL, most (∼90%) of the reduced activation ferritic/martensitic steel structures in the tritium-breeding blanket can be classified as intermediate-level waste (ILW). One-third of the total weight of borated stainless steel inside the shielding blanket can also be classified as ILW 100 years after EOL. The remaining components of the K-DEMO are assessed as low-level waste. Based on this study, fusion waste disposal strategies will be planned in parallel with the advance of the K-DEMO design to minimize the fraction of ILW.

Parameter study of L–H transition for plasma operation scenario development in JA DEMO

We have investigated the L–H transition in JA DEMO, a design concept of the steady-state tokamak fusion demonstration reactor. We develop an L–H transition model, which determines the minimum pedestal growth time such that a power ratio, the net plasma loss power across the separatrix over the L–H threshold power, is maintained at a constant value during pedestal growth for a given external heating power. We conduct parameter studies using an integrated modeling code, considering impurity transport. The dependence of the heating power and time required for the L–H transition on the plasma density, temperature, threshold power, impurity density fraction, and impurity transport coefficients is clarified. We evaluate the appropriate impurity density fraction for the L–H transition and the divertor heat load reduction. The fraction depends on the impurity transport coefficients and external heating power. The obtained results are necessary to confirm the L–H transition feasibility and to develop the plasma operation scenario consistent with the various engineering constraints in a design-stage reactor. The heating power required for the L–H transition turns out to be lower than the power supposed to be installable in the current design activities of JA DEMO.

Updates on CEA Design and Experimental Activities on EU DEMO TF System

In the framework of EU design activities for dimensioning the future fusion DEMOnstration reactor (DEMO), in-depth analyses were conducted in EUROfusion context, aiming to define the design of the DEMO magnets system. For the last DEMO baseline, CEA has proposed for Toroidal Field (TF) coils a concept with radial plates (named WP#4) which is ITER-like, with a round conductor embedded in steel plates. In order to consolidate this design, CEA conducted fine analyses that assess thermal hydraulic and mechanical aspects to allow ensuring compliance with design criteria. The outcomes of theses analyses were used to improve the TF design in order to avoid relying on conservative approaches at pre-design stage, which often end up in material over dimensioning, that penalize cost and space occupation. In this regard, two improving approaches will be exposed, that deal with mechanical consideration and temperature operation, together with future perspectives. On the other hand, CEA designed and manufactured in collaboration with ASIPP a full-size conductor sample derived from a previous TF CEA concept (square-in-square conductor) which is expected to operate at 88 kA and 12T. The two conductor legs of this sample were designed by CEA and manufactured by ASIPP following an extensive quality assurance (QA) preparation process. The sample is presently in course of assembly in CEA Cadarache and is expected to be tested in SULTAN facility (Villigen, CH) to assess its behaviour in both DC and AC regimes. An overview of the status of activities and future perspectives is given in this paper.

Activation analysis of a tritium breeding blanket in the Korean fusion demonstration reactor using the SuperMC code

Activation analysis of a tritium breeding blanket in the Korean fusion demonstration reactor using the SuperMC code

The effect of dissolved oxygen respectively dissolved hydrogen on corrosion behavior of CuCrZr alloy in high temperature water

• Corrosion of CuCrZr in high temperature water with dissolved O 2 and H 2 was studied. • Sample in water with dissolved O 2 showed highest weight loss. • Sample in water with dissolved H 2 showed smallest weight loss. • Pure copper precipitates found on sample surface tested in water with dissolved H 2. • H 2 might reduce dissolved copper ions back to atoms, atoms then return to surface. In Japan Demonstration fusion reactor, the divertor plays a critical role for it helps sustaining plasma shape and hence demanding high requirements on its structural materials for the cooling system. CuCrZr alloy is considered as one of candidates and would be used in one independent cooling system. Proposals of operating conditions would be similar with fission light water reactor, where experiences have shown that water chemistry significantly influences corrosion behavior. Available references on the corrosion of CuCrZr also address the importance of water chemistries. Because the design proposal could be one of proposals towards reactor, it is worthwhile an investigation to understand the corrosion behavior of CuCrZr alloy under high temperature water environments. This work aims at the effect of dissolved oxygen, and dissolved hydrogen on corrosion behavior. Tests were conducted in an autoclave with relatively static water flow. Three different water chemistries: water with dissolved oxygen 100 ppb, water with dissolved hydrogen 1600 ppb and degassed water were chosen. The exposure time of CuCrZr coupons was 100 hr. Results showed that specimens in water with oxygen 100 ppb had highest weight loss with light-reflecting, grain-oriented surface while in water with hydrogen 1600 ppb had lowest weight loss with tarnished surface and nanoscale pure copper precipitates attached. It is believed that hydrogen reduced dissolved copper ions back to copper atoms; atoms returned to surface in precipitate forms hence reducing weight loss.

Preliminary assessment of the safety factors in K-DEMO for fusion compatible regulatory framework

We open an avenue for discussing how we can pave the way for compliance with existing regulations is a far-reaching factor for settling nuclear fusion technology. Based on a model of the Korean Fusion Demonstration Reactor (K-DEMO) with a target fusion power of 2.2 GW, we assess the intrinsic safety determinants of internal energy sources, the expected radioactive waste, and the tritium management. Regarding these safety factors, we scrutinize the compatibility of the current legislative environment in Korea with K-DEMO and envisage foreseeable obstacles, such as licensing of the nuclear facilities and acceptability of the radioactive waste. Based on precedent licenses for the Korean Superconducting Tokamak Advanced Research (KSTAR) and lessons learned from the International Thermonuclear Experimental Reactor (ITER), we examine hazardous factors that would threaten regulatory compliance of K-DEMO. This approach can help shape a fusion-compatible framework for consolidating the necessary technical provisions and regulatory baselines reflecting social acceptance with a sense of safety. Fusion-compatible aspects in the regulatory environment are discussed, from fusion philosophy to subordinate administrative and technical guidelines, facility classification, and detailed methods guaranteeing integrity and safety. This paper will contribute to the timely settlement of fusion demonstration facilities and subsequent commercial plants.

Development and Application of SONIC Divertor Simulation Code to Power Exhaust Design of Japanese DEMO Divertor

An integrated divertor simulation code, SONIC, has been developed in order to predict a self-consistent transport solution of the plasma, neutral and impurities in the scrape-off layer (SOL) and divertor. SONIC code has contributed to determining the divertor design and power handling scenarios for the Japanese (JA) fusion demonstration (DEMO) reactor. Radiative cooling scenario of Ar impurity seeding and the divertor performance have been demonstrated to evaluate the power exhaust scenarios with Psep = 230–290 MW. The simulation identified the decay length of the total parallel heat flux profile as being broader than the electron one, because of the ion convective transport from the outer divertor to the upstream SOL, produced by the plasma flow reversal. The flow reversal also reduced the impurity retention in the outer divertor, which may produce the partial detachment. Divertor operation margin of key power exhaust parameters to satisfy the peak qtarget ≤ 10 MWm−2 was determined in the low nesep of 2 − 3 × 1019 m−3 under severe conditions such as reducing radiation loss fraction, i.e., f*raddiv = (Pradsol + Praddiv)/Psep and diffusion coefficients (χ and D). The divertor geometry and reference parameters (f*raddiv ~ 0.8, χ = 1 m2s−1, D = 0.3 m2s−1) were consistent with the low nesep operation of the JA DEMO concepts. For either severe assumption of f*raddiv ~ 0.7 or χ and D to their half values, higher nesep operation was required. In addition, recent investigations of physics models (temperature-gradient force on impurity, photon transport, neutral–neutral collision) under the DEMO relevant SOL and divertor condition are presented.

Interaction of transmutation products with precipitates, dislocations and grain boundaries in neutron irradiated W

Tungsten is the primary candidate materials for the high neutron flux, high temperature components of a future demonstration fusion reactor. Despite this, there is a lack of data on W under fusion relevant neutron doses and irradiation temperatures. In this study, single crystal and polycrystalline W samples irradiated at the High Flux Reactor (HFR) at 900 ∘C were characterised using Atom Probe Tomography (APT) and Scanning Transmission Electron Microscopy (STEM). Bulk chemical and isotopic concentration predictions were validated by analysing the mass spectrum from APT experiments. A post irradiation composition of W-1.26 ± 0.15 at.%Re - 0.08 ± 0.02 at.%Os - 0.01 ± 0.01 at.%Ta was measured in the single crystal sample, whereas W-1.09 ± 0.07 at.%Re - 0.08 ± 0.02 at.%Os - 0.01 ± 0.01 at.%Ta was measured for the polycrystalline. APT and STEM showed that a high number density of Re and Os rich precipitates had formed under neutron irradiation. These typically consisted of a core rich in Re and Os, surrounded by a less dense Re rich cloud. Multiple analysis methods were applied to investigate the composition of these clusters. APT showed that the centres of some of the precipitates had a rod shaped core which were rich in both Re and Os. Line profile analysis suggests that in the centre of the precipitates, the threshold composition for σ phase formation may have been reached, as has been observed in higher transmutation rate experiments. In addition, dislocations, sub grain boundaries and dislocation loops were all shown to be decorated with both Re and Os, in agreement with predictions from DFT simulations.

Divertor of the European DEMO: Engineering and technologies for power exhaust

In a power plant scale fusion reactor, a huge amount of thermal power produced by the fusion reaction and external heating must be exhausted through the narrow area of the divertor targets. The targets must withstand the intense bombardment of the diverted particles where high heat fluxes are generated and erosion takes place on the surface. A considerable amount of volumetric nuclear heating power must also be exhausted. To cope with such an unprecedented power exhaust challenge, a highly efficient cooling capacity is required. Furthermore, the divertor must fulfill other critical functions such as nuclear shielding and channeling (and compression) of exhaust gas for pumping. Assuring the structural integrity of the neutron-irradiated (thus embrittled) components is a crucial prerequisite for a reliable operation over the lifetime. Safety, maintainability, availability, waste and costs are another points of consideration.In late 2020, the Pre-Conceptual Design activities to develop the divertor of the European demonstration fusion reactor were officially concluded. On this occasion, the baseline design and the key technology options were identified and verified by the project team (EUROfusion Work Package Divertor) based on seven years of R&D efforts and endorsed by Gate Review Panel.In this paper, an overview of the load specifications, brief descriptions of the design and the highlights of the technology R&D work are presented together with the further work still needed.

Occupational Radiation Exposure Estimated for Fusion Demonstration Reactors and Beyond

The fusion reactor fueled by deuterium and tritium will generate many neutron activation products, causing occupational exposure and radiation risk. The minimization of occupational radiation exposure (ORE) is one of the safety goals for fusion reactors. However, detailed designs and management schemes are still lacking for fusion reactors, and the ORE evaluations are still well simplified. In this paper, an integrated assessment approach is proposed for fusion reactors at the conceptual or detailed design stage. The core idea is to estimate the ORE by referring to the dose rates and work efforts of mature fission reactors and ITER and modifying the data of these similar systems by a proportional coefficient according to the differences of component scale, operating environment, etc. The results showed that water cooling fusion reactors will generate the highest collective dose of 2635 p-mSv/year, while the PbLi cooling ones come next with about 1684 p-mSv/year and the helium cooling ones are the least. This method will contribute to fusion reactor design, operation, and maintenance optimization at the earlier stages and provide guidance to reduce the overall potential ORE to workers.

Overview of Thermal Hydraulic Optimization and Verification for the EU-DEMO HCPB BOP ICD Variant

When progressing from the International Thermonuclear Experimental Reactor (ITER) to the Demonstration Fusion Reactor (DEMO), a system for transferring plasma heat exhaust to a power conversion system is necessary for the so-called Balance of Plant (BOP). During the preconceptual phase of the EU-DEMO project, different BOP concepts were investigated in order to identify the main requirements and feasible architectures to achieve that goal in the most efficient way. This paper comprises the investigations performed during the DEMO preconceptual design phase (p-CDP) and compares the different variants. The main aspect was focused on the helium-cooled pebble bed (HCPB) breeding blanket (BB) concept. After all assessments were performed, the indirect coupled design (ICD) was chosen as the reference configuration for the DEMO HCPB BOP for further development and optimization. The ICD provides decoupling using a molten salt storage loop, which accumulates thermal power during plasma pulses that are released during dwell periods. The work is supported by simulations using design codes EBSILON and MATLAB/SIMULINK, providing the basis for the next design phase.

Stress Analysis of HCPB BB under Ex-Vessel LOCA Condition

Helium Cooled Pebble Bed Breeding Blanket (HCPB BB) is a kind of concept for the European demonstration fusion reactor (DEMO). The blanket attachment system plays an important role in the mechanical connection of the BB and vacuum vessel. Typically, the mechanical and thermal loads should meet the requirement to avoid collapse of the system with off-normal conditions, e.g., under ex-vessel Loss of Coolant Accident (LOCA. This paper investigates the loading requirement corresponding to the maximum stress that can sustain to avoid the LOCA condition. Firstly, a model of the BB is constructed using SolidWorks. Then, stress analysis is carried out based on the cross section of the blanket. Through simulation, the critical condition for the LOCA case and the maximum stress value for the model are obtained. According to the relevant size dimension from the reference, the blanket’s cross section is drawn, and one can get the stress field under the ex-vessel LOCA through stress analysis. The stress distribution under the ex-vessel LOCA condition is simulated to find out the maximum stress field that the blanket can sustain through this paper. The significance is to predict the possible conditions leading to an accident and find possible methods to avoid them.

Thermal-hydraulic analysis of the CFETR TF coils when subject to nuclear heat load

China Fusion Engineering Test Reactor (CFETR) has received much attention over the past several years, aiming at bridging the gap between the International Thermonuclear Experimental Reactor (ITER) and the Demonstration Fusion Reactor (DEMO). The toroidal field (TF) coils play an important role in the tokamak, which provide the main magnetic field to confine the plasma. In order to evaluate the feasibility of superconducting magnets used in CFETR, it is important to predict the magnet performance in terms of temperature margin during normal operation conditions. The simulations confirm the need to increase the mass flow rate, or decrease the hydraulic length of high field windings. The results show that the proposed reduction of hydraulic length is more effective to increase the minimum temperature margin.

A study on ohmic plasma initiation for JA DEMO

We have investigated ohmic plasma initiation for JA DEMO, a design concept of the tokamak demonstration fusion reactor examined in Japan. A calculation model describing the tokamak plasma initiation phase consisting of the two-dimensional analysis of the coil and eddy currents and the zero-dimensional plasma analysis is developed. It is revealed that the conducting shell does not necessarily hinder plasma initiation, while it provides the stabilizing effect of vertical displacement events. The available loop voltage and prefill gas pressure for initiation in JA DEMO are clarified for developing the initiation scenario. We demonstrate a possible scenario of ohmic plasma initiation, which is consistent with the design specifications of superconducting poloidal field coils.

A comparative study of internal kink stability in EU DEMO designs with negative and positive triangularity

Internal kink (IK) instability is investigated for European demonstration fusion reactor (EU DEMO) plasmas in both negative triangularity (NT) and positive triangularity (PT) configurations. For NT plasmas, the IK becomes more unstable as an ideal conformal wall moves away from the plasma boundary, with the mode growth rate saturating at the wall radial location of about , where a is the plasma minor radius and b the wall radial location. The plasma resistivity destabilizes the IK mode. The effect of sub-sonic toroidal plasma flow is sufficiently weak and can thus be ignored for these EU DEMO equilibria. These results are consistent with those for PT plasmas, albeit with larger mode growth rate in the NT configuration. Both perturbative and self-consistent magneto-hydrodynamic (MHD)-kinetic hybrid calculations predict (partial) stabilization of the IK modes in both NT and PT configurations, with inclusion of various kinetic contributions. Precessional drift motion of trapped fusion-born alphas in EU DEMO produces weak stabilization to the IK mode. Stronger stabilization occurs with the toroidal precession of trapped thermal particles (ions and electrons) and the bounce-transit motion of thermal ions. The stabilization is similar between the NT and PT configurations, due to the similarity of the mode eigenfunction (occupying a nearly circular region in the plasma core) despite the sign difference in the triangularity. The non-perturbative MHD-kinetic hybrid model predicts much less stabilization of the mode than the perturbative model, primarily due to the self-consistent determination of the mode eigenvalue in the former. Generally, no significant difference in the IK mode stability is found between the NT and PT plasmas in EU DEMO.

Feasibility of D-D start-up under realistic technological assumptions for EU-DEMO

One of the main issues in view of the realization of a DEMOnstration fusion reactor is the availability of a sufficient external supply of tritium (T) to start operation. T is an unstable nuclide, which is almost absent in nature and is currently available as by-product in e.g. CANDU, whose operation in the next decades (both in terms of life extension of existing reactors and construction of new ones) is at the moment under debate. During DEMO operation, T will be generated on-site by breeding blanket, employing the neutrons originating from D-T reaction. However, it is considered that a certain initial amount of T is needed to start operation, the so-called start-up inventory. An alternative approach consists of obtaining the start-up inventory exploiting reactions occurring in a D-D plasma, which generate T both directly in the plasma and via breeding in the breeding blanket. In the present paper, the conditions under which D-D start-up becomes a favorable option for a power plant are discussed. The analysis mainly focuses on the EU-DEMO reactor concept, for which design data are sufficient for a fairly quantitative evaluation of the relevant parameters. It is found that the unavoidable presence of elements requiring saturation before they are able to release significant amounts of T clamps the T production rate to the same order of magnitude as D-D reaction rate. Thus, under very optimistic assumptions, several hundreds of full-current D-D discharges are necessary for T to be available for plasma fueling, but more realistic estimates let this number raise up to several thousands.

Updates on Magnet Design For EU-DEMO Reactor: Optimization Studies on TF and CS Systems

In the framework of EU design activities for dimensioning the future fusion DEMOnstration reactor (DEMO), extensive analyses were conducted in EUROfusion context, aiming at ultimately defining the design of the DEMO magnets system. In this objective CEA (Commissariat à l'Énergie Atomique et aux Énergies Alternatives) proposes design for both Central Solenoid (CS) and Toroidal Field (TF) cryomagnetic systems. Considering the latest DEMO reactor baseline both systems were investigated using pre-dimensioning CEA tool MADMACS, with an outcome of several options for each of them. In a second step those designs were investigated with detailed analyses (electromagnetic, thermohydraulic, mechanic). In the paper a broad analysis of the CS design considering fatigue constraints is shown, together with recommendations considering flux target. The TF system considered is an ITER-like option with radial plates and the detailed electromagnetic analysis is exposed and discussed in comparison with former methods. The outcome is used for a thermohydraulic detailed static analysis and a mechanical analysis of TF winding pack is provided and discussed in terms of results but of method as well. Global considerations and recommendations will also be provided to give a first view on cryomagnetic system optimization with respect to cost merits related to both magnet systems.