Fusion Reactor Blanket Designs Research Articles

Evaluation of the impact of plasma operation scenarios on the fusion reactor blanket design using an integrated numerical plasma and neutronics analysis suite

A very limited operation scenario has been used for fusion neutronics for the design of tritium breeding blankets. However, many advanced operation scenarios are under development and the impact of choosing a plasma operation scenario for fusion neutronics, which requires integration of neutronics analysis with plasma analysis, needs to be evaluated. An automated process from plasma analysis to neutronics analysis to create a viable fusion plasma neutron source was considered using a few different codes. In this work, we develop an integrated suite combining plasma modeling codes and neutronics codes. McCARD, a Monte-Carlo neutron-photon transport simulation code, is integrated into TRIASSIC, a tokamak reactor integrated automation suite for simulation and calculation, as the module for neutronics analysis of various plasma operation scenarios. McCARD integration allows TRIASSIC to perform neutronics analysis along with analytical and predictive plasma simulations. To evaluate this integrated suite, KSTAR experimental data were used to calculate the neutron flux at the first wall via TRIASSIC with the McCARD module in three different cases. First, the effect of plasma shape parameters, elongation and triangularity, on neutron flux is analyzed. Second, plasma scenario with the changes to the shape and the performance of plasma over time is studied to evaluate the time varying simulation. Third, two different plasma scenarios are compared to analyze the impact of plasma performance on neutron flux.

Deuterium permeation and lithium‑lead corrosion behaviors of ceramic‑iron joint coating

In fusion reactor blanket design, the main challenges are to control the permeation of tritium and prevent the corrosion of the structural materials. This study focuses on investigating the behavior of ceramic‑iron joint coatings in terms of deuterium permeation and Li-Pb corrosion aiming to develop effective barriers against permeation while providing corrosion protection. The research involved fabricating ZrO2 and Fe2O3 ceramic coatings on F82H steel substrates using metal-organic decomposition followed by joining the coated samples with Fe foils using a hot press machine. Deuterium permeation measurements were conducted to assess the effectiveness of the coating, and liquid Li-Pb exposure tests in a flow environment were performed to evaluate its resistance against corrosion. The results showed that the coating reduced deuterium permeation flux by a factor of 2000 at 400 °C and 7000 at 550 °C compared to uncoated samples suggesting an improved crystallinity or grain growth without degradation due to the joining process. Furthermore, the corrosion layers formed on the surface of the Fe layer after the flowing Li-Pb exposure test. Although this layer gradually decreased in thickness over time while not significantly affecting the underlying Fe layer's thickness. This observation indicates that the corrosion layer performed mitigating the corrosion rate of the Fe layer. This paper aims to demonstrate how ceramic‑iron joint coatings can be compatible with the advanced fusion reactors.

Qualification and Commissioning of Helium Flow Loop Experiment for Blanket Design Measurements

Sufficient cooling of plasma-facing materials remains an outstanding challenge in the design of fusion reactor blankets in commercial power demonstration plants. Due to its chemical inertness and low neutron interaction cross section, pressurized helium is a candidate coolant fluid for such systems; however, helium has a small thermal mass compared to liquid coolants, potentially reducing heat removal performance. To address this need, a number of heat transfer enhancements have been proposed to improve the cooling efficiency of such components, thereby decreasing pumping power needs and improving overall plant efficiency. Toward this end, a helium flow loop experiment (HFLE) has been designed and commissioned to test advanced passive heat transfer enhancements in unit-cell test sections, providing necessary data for model validation and subsequent system design. The HFLE is designed to provide flow of pressurized (up to 4 MPa) helium at flow rates up to 80 g/s, enabling heat transfer and pressure drop measurements in test pieces at Reynolds numbers in excess of 180 000. To explore the effects of novel and complex heat transfer enhancements, test sections are produced via additive manufacturing, providing geometries not typically obtainable by conventional machining. In this work, we present results from HFLE commissioning and the initial thermal-hydraulic tests of an additively manufactured rifled-rib test section. Results are compared to smooth pipe correlations, and plans are described for future HFLE measurements. These preliminary experiments indicate the utility of the HFLE for heat transfer enhancement testing and simulation validation activities.

Three-dimensional MHD flow in moderate change ratio orifice

In fusion reactor blanket design, liquid metals are attractive working fluids since it is possible to combine in a single fluid the functions of coolant, tritium carrier and breeder. These electrically conductive fluids flow in the presence of a strong magnetic field, inducing the appearance of Lorentz forces and magnetohydrodynamic MHD effects. Increased pressure loss, particularly in complex geometry elements, is a critical point for blanket design. The MHD flow through an orifice plate made by electroconductive walls (c = 0.01 ÷ 0.1) has been analysed in this paper using ANSYS CFX in the range Re = 108, and Ha = 0 ÷ 300. A wide recirculation region is detected after the flow exits the orifice, with potentially harmful consequences for efficient tritium removal. Large pressure loss occurs in the orifice due to conductive wall and non-negligible axial length. The 3D pressure drop term is characterized through a local resistance coefficient (k) that is found to be k ≈ 0.205 for well conducting walls (c = 0.1) and k ≈ 0.063 for poorly conducting ones (c = 0.01).

Numerical Investigation of the MHD Flow Under the Influence of Heat Transfer in Rectangular Ducts

Liquid metal (LM) breeder blanket is a topic of great interest in the fusion reactor blanket design because of many advantages. In the breeding blanket of the fusion reactor, the dynamic viscosity of LM is impacted by heat transfer, resulting in the change of velocity distribution. The influence of heat transfer on magnetohydrodynamic flow in a rectangular duct at high Hartmann number is investigated by a coupling method. By this method, the velocity field is calculated through a second-order projection method, coupled with the temperature distribution computed by a finite-volume method. The numerical result without temperature influence is validated by Hunt’s and Shercliff’s analytical solutions, which shows very good accuracy. On the basis of this numerical code, the velocity distribution of a Hunt’s case with temperature influence is simulated. The simulation result indicates that the velocity field is different from the benchmark solution as the result of the heat transfer.

Magnetohydrodynamic flow and heat transfer around a heated cylinder of arbitrary conductivity

The interaction of the liquid metal with the plasma confinement magnetic field constitutes a challenge for the design of fusion reactor blankets, due to the arise of MHD effects: increased pressure drops, heat transfer suppression, etc. To overcome these issues, a dielectric fluid can be employed as coolant for the breeding zone. A typical configuration involves pipes transverse to the liquid metal flow direction. This numerical study is conducted to assess the influence of pipe conductivity on the MHD flow and heat transfer. The CFD code ANSYS CFX was employed for this purpose. The fluid is assumed to be bounded by rectangular walls with non-uniform thickness and subject to a skewed magnetic field with the main component aligned with the cylinder axis. The simulations were restricted to Re = (20; 40) and M = (10; 50). Three different scenarios for the obstacle were considered: perfectly insulating, finite conductivity and perfectly conducting. The electrical conductivity was found to affect the channel pressure penalty due to the obstacle insertion only for M = 10 and just for the two limiting cases. A general increment of the heat transfer with M was found due to the tendency of the magnetic field to equalize the flow rate between the sub-channels individuated by the pipe. The best results were obtained with the insulating pipe, due to the reduced electromagnetic drag. The generation of counter-rotating vortices close to the lateral duct walls was observed for M = 50 and perfectly conducting pipe as a result of the modified currents distribution.

Direct numerical simulation of turbulent liquid metal flow entering a magnetic field

This paper presents direct numerical simulations (DNS) of fully developed turbulent liquid-metal flow in a circular duct entering a magnetic field. The case of a magnetohydrodynamic flow leaving a strong magnetic field has been extensively studied experimentally and numerically owing to its similarity to typical flow configurations appearing in liquid metal blankets of nuclear fusion reactors. Although also relevant to the design of fusion reactor blankets, the flow entering the fringing field of a magnet remains unexplored because its high intricacy precludes any simplification of the governing equations. Indeed, the complexity of the magnetohydrodynamic–turbulence interaction can only be analysed by direct numerical simulations or experiments. With that purpose, this paper addresses the case of a fully developed turbulent flow (Reτ≈520) entering low, intermediate and strong magnetic fields under electrically insulating and poorly conducting walls by means of three-dimensional direct numerical simulations. Purely hydrodynamic computations (without the effect of the magnetic field) reveal an excellent agreement against previous experimental and numerical results. Current MHD results provide a very detailed information of the turbulence decay and reveal new three-dimensional features related to liquid-metal flow entering strong increasing magnetic fields, such as flow instabilities due to the effect of the Lorentz forces within the fringing region at high Ha numbers.

Magnetohydrodynamic Flows in Arbitrary Geometries in Strong, Nonuniform Magnetic Fields — A Numerical Code for the Design of Fusion Reactor Blankets

Magnetohydrodynamic flows play an important role in the design of liquid-metal fusion reactor blankets. The interaction of the plasma-confining strong magnetic field and the electrically conducting coolant and breeding material may cause high pressure drop and unusual flow structures compared with hydrodynamic flows. In strong magnetic fields, duct flows exhibit a core where viscous effects are unimportant, while all flow variables are matched to the boundary conditions within extremely thin layers. In the inertialess inductionless limit, the governing equations can be reduced to a set of coupled two-dimensional equations for pressure and potential through analytical integration in the core and the layers. The use of curvilinear boundary-fitted coordinates leads to a unique numerical procedure for flow calculations in arbitrary geometries. The wide range of possible applications is demonstrated by some examples.

The evolution of US helium-cooled blankets

This paper reviews and compares four helium-cooled fusion reactor blanket designs. These designs represent generic configurations of using helium to cool fusion reactor blankets that were studied over the past 20 years in the United States of America (US). These configurations are the pressurized module design, the pressurized tube design, the solid particulate and gas mixture design, and the nested shell design. Among these four designs, the nested shell design, which was invented for the ARIES study, is the simplest in configuration and has the least number of critical issues. Both metallic and ceramic-composite structural materials can be used for this design. It is believed that the nested shell design can be the most suitable blanket configuration for helium-cooled fusion power and experimental reactors.

Inverse energy cascades in three-dimensional turbulence

Fully three-dimensional magnetohydrodynamic (MHD) turbulence at large kinetic and low magnetic Reynolds numbers is considered in the presence of a strong uniform magnetic field. It is shown by numerical simulation of a model of MHD that the energy inverse cascades to longer length scales when the interaction parameter is large. While the steady-state dynamics of the driven problem is three-dimensional in character, the behavior has resemblance to two-dimensional hydrodynamics. These results have implications in turbulence theory, MHD power generator, planetary dynamos, and fusion reactor blanket design.

Neutron Energy Spectra in Lithium: Measurements and Calculations

This paper reports a 15-MeV clean neutron field used to investigate various problems encountered in fusion reactor blanket designs. In particular, the adequacy of the {sup 7}Li tritium production neutron cross section was analyzed by comparing measured neutron energy spectra in a lithium tank with detailed Monte Carlo calculations. New evaluations are needed that would consider the various mechanisms that contribute to tritium production in{sup 7}Li and determine the shape of the secondary neutron energy spectrum. It is also found that uncertainties in the alignment of the source-target-detector configuration in integral experiments may result in considerable discrepancies in the analyses of such experiments.

Authors

Authors

Mhd enhancement of heat transfer and its relevance to fusion reactor blanket design

The effect of a uniform magnetic field on the heat transfer of liquid metal forced flows in straight rectangular channels subjected to a uniform heat flux on one of the walls parallel to the field is experimentally studied. The experiments covered the domain 3.8 ⩾ U ⩾ 18 cm/s; 0 ⩾ B ⩾ 0.9 T; 0 ⩾ M ⩾ 300 and 4 × 103 ⩾ Re ⩾ 2 × 104. It is found that the application of the magnetic field improves the heat transfer in channels made of both non-conducting walls and of conducting walls. Whereas in the former case this improvement results from the creation of enhanced anisotropic turbulence across he entire channel, in the conducting channel the improvement results from an increase in the velocity near the heated wall as well as enhancement of turbulence in the vicinity of this wall.

Tritium breeding performance of a self-cooled water based blanket

A major issue in the design of fusion reactor blankets is the trade-off between tritium breeding and other blanket design requirements. While net breeding is required, the blanket design should also ensure adequate heat removal, efficient power production and sufficient shielding. A novel aqueous self-cooled blanket concept (ASCB) based on lithium compounds dissolved in water has been proposed and analyzed using one-dimensional neutronics calculations. This concept utilizes zircaloy for the structural material and a vanadium alloy for the first wall. Light water as well as heavy water systems lead to an acceptable design with respect to tritium breeding. One-dimensional tritium breeding ratios in the range 1.1–1.2 seem feasible for the proposed concept. Contrary to conventional blanket designs, the 3-D tritium breeding ratio is expected to be comparable to the 1-D performance because of the additional breeding in water cooled duct shields and high heat-flux components. The resulting design is simple, utilizes materials with a large data base, does not require additional neutron multiplying materials, and satisfies the commonly proposed criteria for a fusion blanket.

Tritium breeding performance of a self-cooled water based blanket

A major issue in the design of fusion reactor blankets is the trade-off between tritium breeding and other blanket design requirements. While net breeding is required, the blanket design should also ensure adequate heat removal, efficient power production and sufficient shielding. A novel aqueous self-cooled blanket concept (ASCB) based on lithium compounds dissolved in water has been proposed and analyzed using one-dimensional neutronics calculations. This concept utilizes zircaloy for the structural material and a vanadium alloy for the first wall. Light water as well as heavy water systems lead to an acceptable design with respect to tritium breeding. One-dimensional tritium breeding ratios in the range 1.1–1.2 seem feasible for the proposed concept. Contrary to conventional blanket designs, the 3-D tritium breeding ratio is expected to be comparable to the 1-D performance because of the additional breeding in water cooled duct shields and high heat-flux components. The resulting design is simple, utilizes materials with a large data base, does not require additional neutron multiplying materials, and satisfies the commonly proposed criteria for a fusion blanket.

Evaluation of surface modifications for oxidation protection of vanadium-base alloys in helium-cooled blanket designs

Vanadium-base alloys are currently being considered for application to high-temperature helium-cooled fusion reactor blanket designs. A major drawback is their known sensitivity to low levels of oxidizing impurities that would normally be present as water vapor in the helium coolant gas. Although baseline data on the oxidation resistance of vanadium alloys are lacking, protection of vanadium alloy surfaces exposed to the helium coolant will probably be necessary for contiuous blanket operation at temperatures above 875 K. Surface alloying of vanadium-base alloys appears to be a feasible method of creating an oxidation-resistant surface. To examine this approach, a series of V-15Cr-5Ti alloy sheet specimens was surface alloyed with chromium. Oxidation tests were then conducted in flowing helium containing up to 100 vppm H 2O. Temperatures up to 925 K and times up to 1000 h were utilized. Control samples of unmodified V-15Cr-5Ti and a Type 316 stainless steel were also included. Significant reductions in oxygen pickup were observed with surface-alloyed specimens. These results suggest that surface alloying techniques offer great promise for protecting vanadium alloy surfaces against oxidation at blanket operating temperatures.

Status of Fusion Reactor Blanket Design

This paper provides a brief review of the Blanket Comparison and Selection Study (BCSS)1 and an overview of more recent fusion reactor blanket design efforts. Specific areas covered include improve...

Experimental study of neutron multiplication in BeO by the shift-register coincidence technique

The multiplication of 14 MeV neutrons by Be, which is used as a neutron multiplier in the design of fusion reactor blankets, is of topical interest as T-breeding depends on this. Earlier investigations have shown that the experimental multiplication of 14 MeV neutrons in Be and BeO is 20–25% lower than the calculated values. Recently, a different method has been proposed to measure the leakage multiplication based on a statistical correlation technique. This note explores the applicability of such a technique to determine experimentally the leakage multiplication of Am-Be source neutrons of average energy 4.5 MeV in a BeO assembly. The experimental method uses a high-level neutron coincidence counter (HLNCC) equipped with a shift-register coincidence logic unit. The deduced value of neutron leakage multiplication for 5 and 10 cm thick BeO in near rectangular geometry is 1.12 ± 0.03 and 1.15 ± 0.03, respectively.

A Comparison of Different Concepts of Integral Experiments for Fusion Reactor Blanket Design

Different concepts of integral experiments for fusion blanket neutronics are investigated. The first is with the neutron source (tritium target of a neutron generator) located inside of or in immediate proximity to the stack of blanket materials under consideration. The second is based on irradiation of the stack by means of a collimated and, therefore, monoenergetic T(d,n)UHe neutron beam with a tritium target placed outside the stack. The comparison between the different concepts is carried out by means of the Monte Carlo transport code MCNP with continuous energy treatment. The comparison between the two approaches reveals that the integral experiments with a collimated monoenergetic T(d,n)UHe neutron beam result in a neutron spectrum that is better correlated with the details of elastic and inelastic scattering to the first level of the material's nuclei than the one with a neutron source inside a stack. In the case of a collimated neutron beam, there is a clearer separation between energy regions of different neutron interactions and, therefore, the source of discrepancies between measurement and calculation can be identified more easily and corrected by a proper treatment of the cross sections of the specified nuclear reactions.

Authors

Authors