Neutron Wall Loading Research Articles

An innovative coil fabrication and assembly process for the VNS tokamak based on in-situ winding

The paper introduces a new approach to assemble tokamak machines applied to the plasma-based Volumetric Neutron Source (VNS) device, addressing challenges related to machine topology. The VNS, designed as a compact tokamak, serves as a qualification testbed for in-vessel fusion components, especially the breeding blanket. The project aims to bridge the technology gap between ITER and DEMO. The machine size has been minimized to maximize the Neutron Wall Load (NWL). This results in a small, dense plasma and, due to the massive neutron shielding structure, a comparably large distance between the plasma and the equilibrium magnets, challenging the primary stabilization and shaping of the plasma. To overcome this, a magnet system architecture is proposed for VNS in which the poloidal field (PF) coil system is placed inside the TF coils. An assembly procedure is proposed, where the superconducting PF coils are first completed and positioned around the vacuum vessel, followed by the in-situ winding and assembly of the toroidal field coil system. The main magnetic cage components of the machine are based on High-Temperature Superconductors (HTS) to avoid the need for heat treatment. The paper presents the advantages of the magnet system architecture, and the preliminary feasibility study of the machine assembly process.

Study on hot isostatic pressing conditions of ARAA for fabrication of the breeding blanket first wall

The helium-cooled concept with solid breeding materials is one of the candidates for demonstration fusion power reactor (DEMO) breeding blanket. The first wall (FW) is a core component of the breeding blanket. The FW should keep the structural integrity during the normal plasma operation which is subject to high surface heat flux, neutron wall load, and high coolant pressure simultaneously. Various fabrication methods have been studied for the development of the FW fabrication technologies. This study focuses on FW fabrication technologies with hot isostatic pressing (HIP) bonding which is considered as the most promising method for achieving such complex shape like FW having internal cooling channels. We investigated the HIP bonding conditions using Advance Reduced Activation Alloy, ARAA, which is a reduced activation ferritic-martensitic steel under development in Korea. Temperature, holding time and pressure were considered as the parameters to optimize the HIP bonding conditions. The microstructures around the HIP joints changed with respect to the HIP bonding temperature. The amount of oxide particles around the HIP joints varied depending on the degassing conditions. The tensile strength of the HIP joint was not greatly affected by the presence of the HIP bonding line and degassing conditions. On the other hand, it has been confirmed that the Charpy absorbed energy of the HIP joint was significantly affected by degassing conditions. An increase in oxide particles around the HIP joints occurred when degassing temperature and vacuum level were low, and as the quantity of oxide particles increased, the Charpy absorbed energy of the HIP joint decreased substantially.

Massive parametric study for prospective design space of a compact tokamak fusion reactor

Using a novel computational algorithm that integrates tokamak systems analysis with neutron transport calculations, the minimum major radius (R0) and resulting maximum magnetic field at the plasma center (BT) can be determined uniquely for a given plasma performance. Plasma performance was varied extensively using a supercomputer (KAIROS) and scanning over a wide range of physics, technology, and system parameters to derive the optimal design space for a compact tokamak fusion reactor. Given the design goals of fusion gain Q > 20.0, net electric power > 100 MW, neutron wall loading < 2.0 MW/m2, indicator of divertor power handling PSOL/R0 < 25 MW/m, direct capital cost < $4.0 billion, and steady-state operation, a prospective design space was determined depending on the levels of physics and technology. Advanced engineering features, such as maximum allowable magnetic field at the toroidal field coil (Bmax = 23 T), were implemented by adopting high-temperature superconducting magnet technology, using an advanced shield material such as tungsten carbide (WC), and a plug-bucked magnet support structure. It was found that by exceeding the energy confinement scaling laws of IPB98y2, a design space of R0 < 4.0 m could be accessed under the following conditions for a conventional tokamak: confinement enhancement factor H > 1.7, bootstrap current fraction fBS > 0.55, magnetic field at the magnetic axis BT > 4.0 T, fusion power Pfusion > 600 MW, and thermodynamic efficiency ηth > 0.33. For a spherical tokamak, these conditions were H > 1.3, fBS > 0.6, BT 〈 6.0 T, and Pfusion 〉 500 MW. Sensitivity studies on the energy confinement scaling laws showed that stricter conditions on the physics parameters were required with the ITPA20 scaling law, whereas the conditions were mitigated with the β-independent scaling law.

Nuclear scoping analysis of ITER bioshield top lid toward its preliminary design review

During ITER operations, electronics located in the crane hall, which is above the tokamak, will be exposed to neutron and photon fields from both the plasma and the activated water. To protect the electronics, the implementation of dedicated shielding on the crane hall platform and the bioshield top lid is required. The design demands optimisation attending to constructability, weight limits, and radiation shielding requirements. This work evaluates eight shielding configurations by assessment of the neutron flux and dose accumulated over 4700 h of operation at 500 MW for electronics protection. This corresponds to a neutron wall load of 0.3 MW a/m² as specified in the ITER Project Specification. An intermediate-source approach has been followed with SRC-UNED, considering all relevant radiation sources while minimising the computational time required. Results were presented at the top lid Conceptual Design Review aiming to support decision-making. Further optimisation has since been performed to reach a top lid proposal for its Preliminary Design Review. All outcomes show that radiation levels above the north and south crane hall platforms are compatible with the critical electronics requirements.

Effects of Neutron Source Models on Neutron Wall Loading in the EAST Tokamak

As a significant input for neutronics analysis of tokamak devices, the plasma neutron source is a bridge connecting fusion plasma physics and engineering design. A plasma neutron source model is established based on the EAST plasma configuration. The maximum neutron wall loading (NWL) is near the outboard midplane and is more than ~20% that of the inboard midplane. The investigations demonstrate that special care should be taken for radiation shielding and protection of the key components located on the outboard midplane and on the inboard midplane. The effect of the tritium accumulation on neutronics analysis should be carefully considered for EAST tokamak D-D plasma operations. The effect of density peaking (DP) on the neutronics analysis is also investigated. It is evident that the peak NWLs are all near the outboard midplane and that the poloidal distributions of the NWL are slightly different for these cases. With increasing DP, both the outboard and inboard peak NWLs decrease. However, the decrease in NWL is very small; NWL decreases only 11% when the neutron source peak increases about 1.5 times.

Three-dimensional thermal-hydraulics/neutronics coupling analysis on the full-scale module of helium-cooled tritium-breeding blanket

Blanket is of vital importance for engineering application of the fusion reactor. Nuclear heat deposition in materials is the main heat source in blanket structure. In this paper, the three-dimensional method for thermal-hydraulics/neutronics coupling analysis is developed and applied for the full-scale module of the helium-cooled ceramic breeder tritium breeding blanket (HCCB TBB) designed for China Fusion Engineering Test Reactor (CFETR). The explicit coupling scheme is used to support data transfer for coupling analysis based on cell-to-cell mapping method. The coupling algorithm is realized by the user-defined function compiled in Fluent. The three-dimensional model is established, and then the coupling analysis is performed using the paralleled Coupling Analysis of Thermal-hydraulics and Neutronics Interface Code (CATNIC). The results reveal the relatively small influence of the coupling analysis compared to the traditional method using the radial fitting function of internal heat source. However, the coupling analysis method is quite important considering the nonuniform distribution of the neutron wall loading (NWL) along the poloidal direction. Finally, the structure optimization of the blanket is carried out using the coupling method to satisfy the thermal requirement of all materials. The nonlinear effect between thermal-hydraulics and neutronics is found during the blanket structure optimization, and the tritium production performance is slightly reduced after optimization. Such an adverse effect should be thoroughly evaluated in the future work.

Study on the effect of neutron shielding design on tritium breeding based on water-cooled ceramic blanket for CFETR

The Chinese Fusion Engineering Test Reactor (CFETR) is a magnetic confinement fusion reactor independently designed and developed by China, which is based on the ITER design experience. As one of the candidates for the CFETR, the water cooled ceramic blanket (WCCB) mainly carries out many important tasks, such as tritium breeding and radiation shielding. The high tritium breeding ability is one of the most significant goals of the blanket. For the design of the CFETR, in addition to meeting the requirements of tritium breeding, the design of the blanket must also consider meeting relevant shielding limits. Due to the limited space of the blanket, the improvement in tritium breeding space will inevitably lead to the reduction in neutron shielding space behind the breeding area, and vice versa. Therefore, under the premise of meeting the requirements of neutron shielding, increasing the tritium breeding space as much as possible is the focus of research. In this work, a three-dimensional neutronics model containing the WCCB blanket is developed, and the neutronics performance is calculated based on 1.5 GW fusion power. A set of nuclear analyses are carried out by the MCNP code, including analysis of the neutron wall load, tritium breeding ratio (TBR), and fast neutron fluence of the TF coil. It is found that the shielding space of certain blanket modules could be optimized. After the shielding optimization, the global TBR increased from 1.168 to 1.186, an increase of 0.018 TBR. The current research has important guiding significance for the future design and optimization of the WCCB for the CFETR.

The investigation of the neutronic responses of components in the K-DEMO

The neutronic response of consisting materials of the Korean fusion DEMOnstration reactor (K-DEMO) in its fusion neutron environment is a crucial consideration factor from the conceptual design stage of the K-DEMO. Especially in the design of in-vessel components (IVCs) of the K-DEMO that are placed in the most extreme neutron irradiation field, neutronic damage of constituent components is a major limiting factor that determines the lifetime of IVCs. From the analysis of the neutronic response of IVCs of the K-DEMO, the neutron wall loading (NWL) related to the tritium breeding ratio and nuclear heating of IVCs can be quantified to assess the self-sufficient supply of tritium, and thermal energy transferred from fusion neutrons, respectively. The calculated NWL shows that the harmonizing design of the cooling configuration of each blanket segment with the corresponding NWL is critical thermos-hydraulic design issue for the efficient utilization of thermal energy in the blanket. Another finding is that the double null magnetic field configuration and related blanket configuration with a water-cooled pebble bed of the K-DEMO make a self-sufficient tritium supply challenging. The implicated lifetime of the first plasma-facing tungsten wall of the K-DEMO is around 2 full power years (FPY) in the severely neutron-irradiated region. For the reduced activation ferritic/martensitic steel layers in the blanket, the lifetime of it is estimated around 4 FPY in the inboard region. Based on the response analysis results of this study, optimization of the design of the K-DEMO will continue iteratively in the future.

Study on neutronics modeling with 22.5° model using ANSYS for CFETR

The detailed China Fusion Engineering Test Reactor (CFETR) 22.5° computer aided design (CAD) model is very difficult to convert into Monte Carlo N Particle Transport Code (MCNP). Manually writing MCNP input data is complicated, which is not only time-consuming but also cannot guarantee accuracy. Therefore, in order to improve the efficiency and accuracy of model transformation, modeling with CAD using CATIA is introduced, and MCNP files are converted by ANSYS. This is because ANSYS has a function that converts CAD “stp” format to MCNP input in the geometry section. Meanwhile, ANSYS can also reverse the converted MCNP input file to inspect which module has the problem. Compared with the software platform that can automatically cut, although the CATIA-to-ANSYS method is inferior in terms of automatic operation, it has advantages in accuracy and quickly dealing with error modules. Moreover, it can also perform parametric modeling in CATIA, which facilitates the optimization of the blanket structure. In this paper, the detailed CFETR 22.5° model was developed, and then parametric modeling of the blanket based on CATIA was performed. Finally, a detailed neutronics model is obtained by ANSYS transformation and inspection. Some representative models were initially validated by comparing volume changes before and after conversion. Then, the final neutronics model was used to calculate the nuclear analyses, including the neutron wall loading, fast neutron flux, and nuclear heating on the inboard side. The results show that the volume of the transformed model is basically consistent with the original model, and the error of results is small.

Radiation Damage Analysis of FNSF Components Using McCad and MCNP

The Fusion Energy System Studies Fusion Nuclear Science Facility (FESS-FNSF) concept represents a transitional step between ITER and a commercial fusion power plant. The FNSF is a conceptualized D-T fueled tokamak with 518 MW of fusion power that has been extensively used to explore and optimize design features. The energetic 14.1-MeV neutrons can produce significant localized heating and activations, and can cause damage to plasma-facing components, which can determine maintenance/outage scheduling needs and also impact the lifetime of the device as a whole. This study illustrates a neutronics analysis that was conducted on a 22.5-degree symmetric sector of the FNSF with the goal of understanding the neutron heating and radiation damage that can be characterized by quantifying the displacements per atom (dpa). Concurrently, this study also focused on the development of analysis capabilities by converting a three-dimensional computer-aided design model of the FNSF into MCNP6.2 input using the McCad code. Accordingly, some confirmatory results on tritium production and the tritium breeding ratio (TBR) are provided to support model validation. The results produced by MCNP6.2 simulations showed that the highest heating and damage occurred in the outboard region, which concentrated approximately 290 MW of the total nuclear heating, in contrast to 97 MW within the inboard region. These results are consistent with previous studies that employed earlier versions of the FNSF concept and different modeling approaches. This study also provides additional details on neutron wall loading, as well as total heating from neutrons and gammas, results which show the total heating of the device (16 sectors) is approximately 477.83 ± 0.80% MW, indicating a neutron energy multiplication factor of 1.15. Additionally, the capability to calculate hydrogen and helium production, as well as dpa, is illustrated. Finally, the neutronics effects of using alternative materials to tungsten carbide were evaluated for the vacuum vessel, low-temperature shield, and structural ring components, which showed that compounds like YH2, Mg(BH4)2, and ZrH2 could reduce the total heating on the magnet and also reduce the TBR.

Inelastic collision effects of high-energy neutrons in tungsten materials

Because the inelastic impact of high flux, high-energy neutrons on tungsten material in a commercial fusion reactor would drastically decrease the service life and steady-state functioning of the core plasma in the future, it is vital to investigate the damage process. In this paper, a particle sputtering theory approach is used to derive an equation for calculating the sputtering yield of high-energy neutrons by simulating the transport of primary knocked-out atoms (PKA) resulting from inelastic collisions of 14.1 MeV neutrons with tungsten materials at different depth levels, namely Y =∫E0Emaxf(E)*Y(E)dE, where f (E) is the energy probability distribution function of PKA, Y (E) is the sputtering yield at a given energy of PKA particles generated by inelastic scattering collisions, Emax is the maximum kinetic energy of PKA, and E0 is the minimum kinetic energy of PKA (generally taken as 0). According to the result, when the neutron wall load is 1 MW/m2, and the annual average equivalent material wastage is less than 1 nm, sputtering damage to tungsten material can be ignored. However, it must be kept in mind that the reflection of the PKA into the plasma will have an impact on the performance of the plasma confinement system and the damage caused by the PKA's internal energy.

A deterministic method for the fast evaluation and optimisation of the 3D neutron wall load for generic stellarator configurations

The neutronic assessment of a fusion power plant design is usually a challenging and time-consuming task involving experts from several disciplines in order to assemble the geometry, source, as well as carry out computationally heavy Monte Carlo transport simulations. In order to overcome this challenge, we present in this work a deterministic method, which combines all these aspects in a single framework that can directly calculate a key neutronics performance indicator, the neutron wall load (NWL), at minimal computational cost for arbitrary stellarator configurations. As our method is based on simple vector and matrix manipulations, a speed on the order of a few CPU-seconds is achieved, which makes it suitable for optimisation frameworks. We demonstrate with a simple optimisation algorithm that it is possible to use our method to generate a first wall with reduced ‘heterogeneity’ in the NWL distribution.

Neutronics analyses of COOL blanket for CFETR

In this study, an advanced liquid blanket based on a liquid PbLi technology is proposed, and it is called a supercritical CO2 (S-CO2) cooled Lithium-Lead (COOL) blanket. As a candidate for the Chinese Fusion Engineering Testing Reactor (CFETR), it aims to achieve high tritium breeding and shielding performance. It uses S-CO2 as the coolant for the structural components and PbLi as the tritium breeder and neutron multiplier. To evaluate the neutronics performances of this blanket, a detailed 22.5° three-dimensional neutronics model of CFETR was built using the cosVMPT conversion tool. To enhance the shielding capability of the blanket, a shielding block was designed and added at the back of the manifold to moderate neutrons. Evaluation results of the neutron damage of toroidal field coils (TFC) components confirmed that the blanket meets shielding requirements. In addition, the main neutronics parameters including neutron wall loading (NWL), tritium breeding ratio (TBR), nuclear heat, and irradiation damage, were calculated using MCNP5. The results indicated that TBR was 1.205, and the total nuclear heat was 1191.64 MW under the fusion power of 1.5 GW. The peak NWL was 1.97 MW/m2, and the maximum DPA and gas production were 14.80 DPA/FPY, 132.65 He appm/FPY, and 631.30 H appm/FPY.

Physics design point of high-field stellarator reactors

The ongoing development of electromagnets based on high temperature superconductors has led to the conceptual exploration of high-magnetic-field fusion reactors of the tokamak type, operating at on-axis fields above 10 T. In this work we explore the consequences of the potential future availability of high-field three-dimensional electromagnets on the physics design point of a stellarator reactor. We find that, when an increase in the magnetic field strength B is used to maximally reduce the device linear size R ∼ B −4/3 (with otherwise fixed magnetic geometry), the physics design point is largely independent of the chosen field strength/device size. A similar degree of optimization is to be imposed on the magnetohydrodynamic, transport and fast ion confinement properties of the magnetic configuration of that family of reactor design points. Additionally, we show that the family shares an invariant operation map of fusion power output as a function of the auxiliary power and relative density variation. The effects of magnetic field over-engineering and the R(B) scaling of design points with constant neutron wall loading are also inspected. In this study we use geometric parameters characteristic of the helical axis advanced stellarator reactor, but most results apply to other stellarator configurations.

Effect of the Fusion Fuels’ Polarization on Neutron Wall Loading Distribution in CFETR

The China Fusion Engineering Test Reactor (CFETR) is a new superconducting tokamak device being designed in China, aiming to bridge the gaps between ITER and future fusion power plants. In addition to the temperature dependence, the cross section also depends on the spin states of the reactant nuclei. In this paper, we calculate the neutron source and neutron wall loading (NWL) distributions and investigate the effect of spin polarization on them. For the two unpolarized scenarios at the CFETR, the neutron source distributions have obvious differences, but the poloidal distributions of the NWL have a similar tendency and are just a little different except near the outboard midplane. For the hybrid mode scenario, the maximum of the NWL is near the outboard midplane. However, for the full parallel or antiparallel polarization, the NWL distributions have a big difference in the poloidal direction, and the maximum of the NWL occurs in the upper region of the first wall. The calculation results show that it is possible to optimize blanket design by using polarized fuels at the CFETR, and then increase the working life of the first wall.

Investigation on the Tritium Self-Sufficiency and Neutron Shielding Based on a Water-Cooled Blanket for CFETR at 1.5 GW

China Fusion Engineering Test Reactor (CFETR) is a new test tokamak device in China, which is to bridge the scientific and technical gaps between ITER and DEMO. One of the main missions of CFETR is to achieve tritium self-sufficiency with a tritium breeding ratio (TBR) of no less than 1.05. In addition, the neutron shielding evaluation is also a significant issue, which could provide a significant reference for the water-cooled ceramic breeder (WCCB) blanket modification and improvements. In this work, a 3-D simplified neutronics model containing the WCCB blanket is developed and the neutronics performance is calculated based on the 1.5 GW fusion power. The nuclear analyses are carried out by the Monte Carlo N-particle (MCNP) transport code, including the TBR, neutron wall loading (NWL), fast neutron flux on inboard side, and nuclear heating deposition on main in-vessel components. It is found that the net TBR reaches 1.23, which meets the condition of tritium self-sufficiency. Moreover, it is found that the peak NWL of inboard and outboard occurs in the equatorial plane. The inboard peak NWL is 1.89 MW/m <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">2</sup> and the outboard peak NWL is 1.97 MW/m <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">2</sup> . In addition, the fast neutron flux and nuclear heating deposition at the inboard midplane in toroidal field coil (TFC) are obtained. According to the radiation limits of the CFETR TFC referring to ITER, the results show the nuclear responses of the TFC meet the limit criteria.

Neurtonics/Thermal-hydraulic analyses of the CFETR HCCB blanket for multiple operation modes under the poloidal nonuniform neutron wall loading condition

Helium Cooled Ceramic Breeder (HCCB) blanket is one significant component of China Fusion Engineering Test Reactor (CFETR). The neutronics and thermal-hydraulic performances of the blanket are the bases of the safe operation for CFETR, and both of them are influenced by the neutron wall loading (NWL) distribution on the blanket. In our previous work, the influences of the poloidal nonuniform NWL on the neutronics performances of the CFETR HCCB blanket have been preliminarily investigated. However, as the nuclear heating rate in each blanket component is adopted as the internal heat source for the thermal-hydraulic analyses, it's necessary to further investigate the influences of the poloidal nonuniform NWL on the thermal-hydraulic performances of the blanket. Besides, CFETR aims to cover multiple fusion powers (from 200 MW to 1.5 GW) at the present stage. Therefore, it's also essential to compare the different influences of the poloidal nonuniform NWL on blanket performances for multiple operation modes.

The advanced tokamak path to a compact net electric fusion pilot plant

Physics-based simulations project a compact net electric fusion pilot plant with a nuclear testing mission is possible at modest scale based on the advanced tokamak concept, and identify key parameters for its optimization. These utilize a new integrated 1.5D core-edge approach for whole device modeling to predict performance by self-consistently applying transport, pedestal and current drive models to converge fully non-inductive stationary solutions, predicting profiles and energy confinement for a given density. This physics-based approach leads to new insights and understanding of reactor optimization. In particular, the levering role of high plasma density is identified, which raises fusion performance and self-driven ‘bootstrap currents’, to reduce current drive demands and enable high pressure with net electricity at a compact scale. Solutions at 6–7 T, ∼4 m radius and 200 MW net electricity are identified with margins and trade-offs possible between parameters. Current drive comes from neutral beam and ultra-high harmonic (helicon) fast wave, though other advanced approaches are not ruled out. The resulting low recirculating power in a double null configuration leads to a divertor heat flux challenge that is comparable to ITER, though reactor solutions may require more dissipation. Strong H-mode access (x2 margin over L–H transition scalings) and ITER-like heat fluxes are maintained with ∼20%–60% core radiation, though effects on confinement need further analysis. Neutron wall loadings appear tolerable. The approach would benefit from high temperature superconductors, as higher fields would increase performance margins while potential for demountability may facilitate nuclear testing. However, solutions are possible with conventional superconductors. An advanced load sharing and reactive bucking approach in the device centerpost region provides improved mechanical stress handling. The prospect of an affordable test device which could close the loop on net-electric production and conduct essential nuclear materials and breeding research is compelling, motivating research to validate the techniques and models employed here.

Preliminary thermal-hydraulic influence of the poloidal nonuniform neutron wall loading on the CFETR HCCB blanket

Abstract Helium Cooled Ceramic Breeder (HCCB) blanket, which has been determined as one of the blanket candidates for China Fusion Engineering Test Reactor (CFETR), is the critical component of the fusion reactor. The neutronics and thermal-hydraulic performances of the blanket are the bases of the safe operation, and they are also the initial boundary conditions for the other performance analyses. Both performances are influenced by the neutron wall loading (NWL) distribution on the blanket and the radial structural arrangement of the internal functional zones. In our previous work, the influences of the poloidal nonuniform NWL on the neutronics performances of the blanket have been investigated preliminarily. However, as the nuclear power deposition in each blanket component is the internal heat source for the thermal-hydraulic analyses, it’s essential to further investigate the influence of the poloidal nonuniform NWL on the thermal-hydraulic performances of the blanket. In this paper, taking the typical CFETR HCCB blanket module as the research object, firstly, both the neutronics and the thermal-hydraulic models are established. For investigating the influence of the poloidal nonuniform NWL on blanket performances, both models are divided into five identical parts along the poloidal direction. Afterwards, by referencing the actual poloidal NWL distribution in CFETR, we have aggregately proposed three kinds of poloidal nonuniform NWL distribution forms, including two symmetrical and one stepped NWL. On these bases, both the neutronics and the thermal-hydraulic analyses under the poloidal nonuniform NWL distributions are conducted, and the calculation results are compared with those corresponding to the uniform NWL distribution. The comparison results show that the maximum temperature of every functional zone calculated under the three nonuniform NWL distributions is generally higher than that corresponding to the uniform NWL distribution, which means that the thermal-hydraulic results calculated under the simplified uniform NWL distribution are not conservative. Under the condition of nonuniform NWL distributions, there is obvious poloidal temperature nonuniformity in every functional zone. Besides, the temperature deviations between the stepped and the uniform NWL distributions are generally larger than those between the symmetrical and the uniform ones. This paper can provide some valuable references for the detailed engineering design, analyses, and optimizations of the CFETR HCCB blanket.

Cost Drivers for a Tokamak-Based Compact Pilot Plant

A comprehensive systems code that includes a range of physics and engineering considerations along with a simplified costing model has been utilized to evaluate the primary cost drivers for a compact tokamak pilot plant. The systems code has been benchmarked against several tokamak reactor designs and is utilized with sophisticated optimization algorithms to develop optimal solutions for a set of user-specified assumptions and design constraints. In contrast to previous models that have focused on the cost of electricity as the key cost metric, this study uses the estimated capital cost of the facility. The analysis suggests that a pilot plant with the following features may offer potential for a cost-attractive pilot plant: A ~ 3, H98y2 > 1.5, Pnet = 200 MW, ~ 1 to 2 h utilizing rare-earth barium copper oxide (REBCO) magnet technology and a plug-bucked central solenoid/toroidal field (CS/TF) magnet support structure. While REBCO magnets offer some advantages relative to Nb3Sn magnets in all cases, the most gain is obtained when combined with the plug-bucked CS/TF bucking solution. Pulsed operation reduces capital cost requirements relative to steady-state operation, especially at low confinement. Cost sensitivity studies indicate that there are significant cost uncertainties associated with the achievable confinement quality, tritium breeding capability, attainable thermal efficiency, and achievable neutron wall loading, suggesting that these areas are the most critical areas in reducing the cost risk for a compact tokamak pilot plant. Further cost sensitivity studies indicate that the estimated cost is most sensitive to the underlying cost of the magnetic coils, providing further impetus to better establish cost-effective means for producing fusion magnets.