Inherent Safety Features Research Articles

Preliminary Study on the Application of Heat Pipe Reactor in Unmanned Underwater Vehicles

At the beginning of the 21st century, due to the re-proposal of the United States space exploration program, heat pipe reactors once again received widespread attention from researchers. With their long lifespan, high energy density, and inherent safety features, they exhibit promising prospects for application in space nuclear reactors, land-based mobile nuclear power sources, and deep-sea unmanned equipment exploration. Consequently, they represent an ideal novel energy source for unmanned underwater vehicles (UUVs). This article selected the SUBOFF submarine model as a basis to calculate UUVs’ navigation resistance at various speeds using the 1+k method. By comparing the converted UUVs navigation power demand with that of traditional batteries and heat pipe stacks’ output power, it was concluded that heat pipe reactors offer superior power supply and endurance capabilities. Subsequently, based on the structural characteristics of the heat pipe stack, this study summarized battery advantages and characteristics while also addressing construction and development aspects of heat pipe stack energy systems. Furthermore, the key technologies that need to be overcome in the application of heat pipe stacks in UUVs are introduced. Through the formula of heat transfer limit of heat pipe, the factors that limit the heat transfer performance of heat pipe are analyzed, and the positive correlation among the number of heat pipe, heat transfer limit and heat transfer power was obtained. Finally, the challenges faced by the future application of heat pipe stacks in UUVs were pointed out.

Fundamental design activity of fast reactor with inherent safety characteristics for effective transmutation of minor actinide

Fundamental design activity of fast reactor with inherent safety characteristics for effective transmutation of minor actinide

Supercritical CO2-cooled fast reactor and cold shutdown system for ship propulsion

A neutronics study of a supercritical CO2-cooled fast reactor core for nuclear propulsion has been performed in this work. The thermal power of the reactor core is 30 MWth and a ceramic UO2 fuel can be used to achieve a 20-year lifetime without refueling. In order to make a compact core with inherent safety features, the drum-type reactivity control system and folding-type shutdown system are adopted. In addition, we suggest a cold shutdown system using gadolinium as a spectral shift absorber (SSA) against flooding. Although there is a penalty of U-235 enrichment for the core embedded with the cold shutdown system, it effectively mitigates the increment of reactivity at the flooding of seawater. In this study, the neutronics analyses have been performed by using the continuous energy Monte Carlo Serpent 2 code with the evaluated nuclear data file ENDF/B-VII.1 Library. The supercritical CO2-cooled fast reactor core is characterized in view of important safety parameters such as the reactivity worth of reactivity control systems, fuel temperature coefficient (FTC), coolant temperature coefficient (CTC), and coolant temperature-density coefficient (CTDC). We can say that the suggested core has inherent safety features and enough flexibility for load-following operation.

High-throughput droplet microfluidics screening and genome sequencing analysis for improved amylase-producing Aspergillus oryzae

BackgroundThe exceptional protein secretion capacity, intricate post-translational modification processes, and inherent safety features of A. oryzae make it a promising expression system. However, heterologous protein expression levels of existing A. oryzae species cannot meet the requirement for industrial-scale production. Therefore, establishing an efficient screening technology is significant for the development of the A. oryzae expression system.ResultsIn this work, a high-throughput screening method suitable for A. oryzae has been established by combining the microfluidic system and flow cytometry. Its screening efficiency can reach 350 droplets per minute. The diameter of the microdroplet was enlarged to 290 µm to adapt to the polar growth of A. oryzae hyphae. Through enrichment and screening from approximately 450,000 droplets within 2 weeks, a high-producing strain with α-amylase increased by 6.6 times was successfully obtained. Furthermore, 29 mutated genes were identified by genome resequencing of high-yield strains, with 15 genes subjected to editing and validation. Two genes may individually influence α-amylase expression in A. oryzae by affecting membrane-associated multicellular processes and regulating the transcription of related genes.ConclusionsThe developed high-throughput screening strategy provides a reference for other filamentous fungi and Streptomyces. Besides, the strains with different excellent characteristics obtained by efficient screening can also provide materials for the analysis of genetic and regulatory mechanisms in the A. oryzae expression system.

A Functional Janus Ag Nanowires/Bacterial Cellulose Separator for High-Performance Dendrite-Free Zinc Anode Under Harsh Conditions.

Aqueous zinc-ion batteries (AZIBs) offer promising prospects for large-scale energy storage due to their inherent abundance and safety features. However, the growth of zinc dendrites remains a primary obstacle to the practical industrialization of AZIBs, especially under harsh conditions of high current densities and elevated temperatures. To address this issue, a Janus separator with an exceptionally ultrathin thickness of 29µm is developed. This Janus separator features the bacterial cellulose (BC) layer on one side and Ag nanowires/bacterial cellulose (AgNWs/BC) layer on the other side. High zincophilic property and excellent electric/thermal conductivity of AgNWs make them ideal for serving as an ion pump to accelerate Zn2+ transport in the electrolyte, resulting in greatly improved Zn2+ conductivity, deposition of homogeneous Zn nuclei, and dendrite-free Zn. Consequently, the Zn||Zn symmetrical cells with the Janus separator exhibit a stable cycle life of over 1000 h under 80mA cm-2 and are sustained for over 600 h at 10mA cm-2 under 50 °C. Further, the Janus separator enables excellent cycling stability in AZIBs, aqueous zinc-ion capacitors (AZICs), and scaled-up flexible soft-packaged batteries. This study demonstrates the potential of functional separators in promoting the application of aqueous zinc batteries, particularly under harsh conditions.

Coordinated control of mHTGR-based nuclear steam supply systems considering cold helium temperature

For industrial energy systems, there commonly exists coupling effect amongst those under-controlled process variables. Coordinated control of energy system enhances the operation performance by fully using the coupling effect positive to stabilization. Nuclear steam supply system (NSSS) module is the central of every nuclear plant, and the coordinated control of NSSS module is important for the stable, reliable and efficient operation of nuclear plants. Currently, the process variables considered in developing NSSS coordinated control strategies (CCSs) are mainly neutron flux, steam temperature as well as hot or average primary coolant temperature. Although cold primary coolant temperature, i.e. the primary coolant temperature at reactor inlet, reflects the heat balance between primary and secondary loops, it still not considered in developing CCS for NSSS modules. The modular high temperature gas-cooled reactor (mHTGR) uses helium as primary coolant and graphite as both moderator and structural material. Due to its inherent safety feature and high temperature, the mHTGR can be applied as a heat source driving industrial processes for power generation, chemical production and etc. To further improve the operation performance of mHTGR-based NSSS module, a passivity-based CCS is newly proposed by considering cold helium temperature. The sufficient condition for asymptotical closed-loop stability is given, and numerical simulation results in the cases of power stepping of 10 % reactor full power (RFP) as well as power ramping of 50%RFP and 5%RFP/min show that the control performance can be effectively improved.

Investigation of neutronic and thermal–electric coupling phenomenon in a 100 kWe-level nuclear silent heat pipe-cooled reactor

Heat pipe-cooled reactors are attractive because of their compact structure, high power capacity and reliability, and inherent safety features. These reactors are distinguished by the use of high-temperature heat pipes that directly bridge the reactor core and energy conversion devices, thus creating a tightly knit modular system. However, comprehensive coupled analysis of these reactors remains challenging because of the intricate multiphysics interactions involved. To address this, the neutronic and thermoelectric coupled phenomena of heat pipe-cooled reactors are investigated in this study. Models are introduced for the point reactor kinetics, core heat transfer, channel heat transfer, two-phase two-dimensional heat pipes, thermoelectric coupling, and cold plates. The channel heat transfer is refined in a 100 kWe-level nuclear silent heat pipe-cooled reactor and a model for the three-dimensional heat transfer process in its thermoelectric matrix established, which effectively resolves the calculation distortions of the lumped model. The maximum deviation between the calculation results from the proposed model and the verification data for a full-system coupled computation is less than 10 K. The core temperature difference from the three-dimensional model is five times that of the lumped model with a maximum average fuel temperature difference of 225.7 K. The maximum and average thermoelectric conversion efficiencies are 15.56 % and 14.43 %, respectively. The failures of one, five, and nine heat pipes are analysed. The failure of a single heat pipe has negligible effects with a peak fuel temperature surge of only 121 K. In contrast, a failure involving nine heat pipes leads to a peak fuel temperature spike of 778 K and a 47.1 % increase in the maximum heat transfer power, marking a critical operational threshold. It is therefore imperative to consider the safety margin in the potential failure of multiple central heat pipes during the preliminary design stages.

Modeling and simulation of a micro gas-cooled nuclear reactor using Modelica

Highly compact micro nuclear reactors, which offer extensive energy benefits across ocean, land, space, and sky applications, have recently emerged as a popular research topic within the international nuclear industry. Due to its excellent inherent safety characteristics, the gas-cooled graphite-moderated reactor with TRISO fuel has attained extensive attention. Nonetheless, micro-reactors exhibit a high degree of system integration, characterized by the tight coupling and mutual constraints among various system functions. Conventional discipline-specific decoupled design patterns find it challenging to tackle the complexity arising from multi-disciplinary couplings. In response, this paper investigates the application of Modelica, a multi-domain unified modeling language, to construct models for several subsystems, encompassing the reactor, energy conversion system, and control system. This approach aims to enhance support for cross-disciplinary design. The accuracy of the reactor core model was verified by high-fidelity CFD simulation results, demonstrating a good agreement. Further investigations were then conducted on the safety and operational characteristics of the whole system. Typically, two simulations were conducted on the Gas cooled micro nuclear reactor (GCMR) design: one focused on an anticipated transients without scram accident scenario and the other on load-following operation. The simulation results demonstrated that the reactor possesses excellent inherent safety, even during extreme accidents. In such scenarios, the reactor is able to achieve shutdown solely through the negative reactivity resulting from increased core temperature. Furthermore, considering the heat accommodated in the reactor system and the constantly generated decay heat, a passive air-cooling mechanism has been investigated and successfully demonstrated with the model. The reactor also exhibits good load-following performance, which can be achieved by simply adjusting helium inventory (or pressure) and control drum position, while maintaining constant core temperature and power generation efficiency. These results can be leveraged to provide guidelines for further detailed designs of the GCMR.

Research on safety characteristics of heat pipe reactor with semiconductor thermoelectric conversion system

The heat pipe reactor exhibits notable characteristics of inherent safety and a compact configuration, rendering it a favored choice for space and ocean reactor designs. This study focuses on investigating the safety characteristics of the SAIRS-C space heat pipe reactor by employing a self-developed TAPIRS-D code under three selected typical accidents. The obtained results demonstrate that the heat pipe reactor exhibits inherent safety features, enabling it to sustain normal operation even in the event of partial heat pipe failure(2 out of 60 heat pipes fail) and partial heat sink loss(cooling rate decreasing from 100% to 10% in 10 s). The reactivity-insertion accident(0.358$ is introduced into the core) is found to exert the most significant influence on the heat pipe reactor. However, the crucial parameters of the reactor during the accident, such as fuel temperature, remain within the established safety limits. The outcomes of this research not only contribute to a comprehensive understanding of safety characteristics of heat pipe reactor, but also provide a crucial guidance for further design enhancements.

An overview of thorium as a prospective natural resource for future energy

Thorium is a naturally occurring radioactive element that has been identified as a potential alternative fuel for nuclear energy production. Additionally, thorium-based nuclear reactors have inherent safety features that reduce the risk of nuclear accidents and proliferation. As a result, there has been growing interest in the development of thorium-based nuclear energy as a viable alternative to fossil fuels. This paper looks at the present status of thorium nuclear fuel technology, providing an overview of thorium as a prospective natural resource for future energy, the global availability of mineral supplies, and discusses the technical, economic, and environmental factors that may influence its implementation. Potential advantages and challenges critical to further development associated with thorium-based nuclear energy are highlighted as well, and an outlook on its future prospects is provided. Thorium offers advantageous physical and chemical properties over uranium, has a higher energy density, and produces less waste, in addition to its greater natural abundance, making it to be considered a “future nuclear fuel”. There are concerns about the cost and scalability of thorium-based nuclear energy, with uncertainty around the cost to develop, build, and operate thorium reactors, as it has not yet been demonstrated in large-scale commercial reactors—although almost all current reactor types have been built and run using thorium—as it is the case with Uranium-based nuclear technology—the dominant form of nuclear energy for over half a century, having received much more investment and attention than thorium-based technology. Thorium has the potential to contribute towards a more sustainable nuclear industry, including lower lifecycle emissions and more efficient resource utilization, but for this, an acceleration of efforts to date is needed to ensure that this becomes an important climate change stabilizing wedge by the mid-21st century.

Multiphysics simulations of Self-Regulating performance of an optimized molten metal fuel microreactor design

Advanced microreactors are expected to play an indispensable role in reliable energy solutions for civilian/military applications in off-grid regions and space applications such as power supplies for Lunar and Mars bases. An innovative fast neutron spectrum heat pipe microreactor (HP-MR) concept was recently proposed with two unique features: molten U-Mn fuel and heat pipes as the heat removal mechanism. The microreactor was designed to be self-regulated, solely relying on reactivity feedbacks from fuel temperature. The reactor design has been further optimized to provide sufficient engineering safety margin and effective heat removal. In this study, the self-regulating performance of the optimized molten metal fuel microreactor was investigated by multiphysics simulations (neutronics, heat-transfer and heat pipe modeling) based on the MOOSE MultiApps system. The responses of the reactor to a series of transient scenarios were simulated to demonstrate the exceptional and inherent safety features of the novel microreactor concept.

Multi-physics coupled simulation on steady-state and transients of heat pipe cooled reactor system

With superior inherent safety and low maintenance demand, heat pipe cooled micro-reactor is a potential solution for decentralized remote electricity supply, as well as space power technology. Multi-physics coupled analysis is important to evaluate the inherent safety feature of micro-reactor. However, in the existing multi-physics coupling simulations, lumped parameter core heat transfer model and low-fidelity heat pipe model cannot accurately predict steady-state and transient safety margin of the reactor. In order to improve the simulation accuracy, a standalone two-dimensional transient high-temperature heat pipe analysis code was developed and validated. The heat pipe analysis code was coupled with the three-dimensional thermal conduction calculation of reactor core, as well as the point reactor kinetics, the alkali metal thermal-to-electric converter model and the heat pipe radiator model, to realize fully coupled analysis of reactor system. With the power distribution in the reactor core, the steady state, startup transient and single heat pipe failure accident were simulated based on the multi-physics coupling. The steady-state simulation results show that the heat flux of heat pipes shows significant non-uniformity in both circumferential and axial directions. The peak heat flux (120 kW/m2), with the twice value of averaged one (60 kW/m2), occurs on the heat pipes near the reactor edge rather than in the center. During reactor startup, the reactivity-insertion rate of 1.67e-3 $/s increases the transient peak heat flux to 260 kW/m2. The isothermal feature of heat pipes results in inverse heat transfer, i.e. from heat pipes to reactor structure, which improves the axial and radial reactor heat transfer and flattens the reactor temperature distribution. In the central heat pipe failure accident, more severe local temperature rise was observed in the core structure than that in the fuel pins. The peak heat flux location is shifted to the heat pipes adjacent to the failed heat pipe. Benefitted from the developed high-fidelity multi-physics coupling approach, the realistic transient phenomena in heat pipe cooled micro-reactor could be revealed.

Exploring the benefits of molten salt reactors: An analysis of flexibility and safety features using dynamic simulation

There has been a growing interest in Molten Salt Reactors (MSRs) in recent years due to the significant potential for increasing flexibility, security, and reliability of the grid, as well as the inherent passive safety features when compared to traditional pressurized water reactors (PWRs). MSRs can help meet many future nuclear energy goals, such as improved sustainability, high security, high efficiency, and high safety passive features, and help reduce nuclear waste. In this study, to investigate MSRs’ passive safety features, a dynamic model of 9 graphite nodes and 18 fuel salt nodes are simulated in 7 safety scenarios. These simulation results are compared with a traditional PWR dynamic simulation. The simulation shows the stability of MSR operations during these 7 safety scenarios, showing that the coolant and graphite temperature within the system stay within the safety limits of operation. The negative feedback coefficient of the fuel salt within MSR cores plays a significant role in stabilizing the power response inside the core, keeping the power from significant excursions. A one-year simulation is also conducted to test the load-following capabilities of MSRs in comparison with traditional PWRs. It is found that MSRs increase the flexibility, reliability, and security of the grid by operating in load-following mode without the need to change the position of the control rods. MSR's increased efficiency also leads to a reduction in backup fossil-fuel based electricity generation by 82% when compared to traditional PWRs operating in load-following mode.

Kirigami-Inspired 3D Printable Soft Pneumatic Actuators with Multiple Deformation Modes for Soft Robotic Applications.

Soft robots have received much attention due to their impressive capabilities including high flexibility and inherent safety features for humans or unstructured environments compared with hard-bodied robots. Soft actuators are the crucial components of soft robotic systems. Soft robots require dexterous soft actuators to provide the desired deformation for different soft robotic applications. Most of the existing soft actuators have only one or two deformation modes. In this article, a new soft pneumatic actuator (SPA) is proposed taking inspiration from Kirigami. Kirigami-inspired cuts are applied to the actuator design, which enables the SPA to be equipped with multiple deformation modes. The proposed Kirigami-inspired soft pneumatic actuator (KiriSPA) is capable of producing bending motion, stretching motion, contraction motion, combined motion of bending and stretching, and combined motion of bending and contraction. The KiriSPA can be directly manufactured using 3D printers based on the fused deposition modeling technology. Finite element method is used to analyze and predict the deformation modes of the KiriSPA. We also investigated the step response, creep, hysteresis, actuation speed, stroke, workspace, stiffness, power density, and blocked force of the KiriSPA. Moreover, we demonstrated that KiriSPAs can be combined to expand the capabilities of various soft robotic systems including the soft robotic gripper for delicate object manipulation, the soft planar robotic manipulator for picking objects in the confined environment, the quadrupedal soft crawling robot, and the soft robot with the flipping locomotion.

Developing PCC and DCC integral effects test experiments at the High Temperature Test Facility

Among the Next-Generation Nuclear Plant (NGNP) designs, the High-Temperature Gas-Cooled Reactors (HTGRs) are very attractive, due to their inherent safety features, high power conversion efficiency, and potential of providing high-temperature process heat. To perform a thorough safety study and to license these types of reactors, sufficient information needs to be provided about the phenomena that occur during accident scenarios. While several experimental research efforts have been dedicated in the past to investigate accident scenarios, knowledge gaps still exist in the phenomena characteristic of pressurized and depressurized conduction cooldown (PCC/DCC) transients as well as for normal operation scenarios. This paper summarizes the Oregon State University High Temperature Test Facility (HTTF) test matrix, experimental campaign, and selected tests results. High Temperature Test Facility is a scaled Integral Test Facility (IET) that is capable of mimicking scaled dimensions and operational conditions of the Modular High-Temperature Gas Cooled Reactor (MHTGR). The goal of the High Temperature Test Facility is to provide experimental data on the DCC, PCC and normal operating scenarios of the reference Modular High-Temperature Gas Cooled Reactor design. The DCC, PCC, mixing, heat up and cooldown tests described in this paper were performed at prototypical Modular High-Temperature Gas Cooled Reactor temperatures, scaled initial pressure conditions (∼200 kPa), and thermal power input of less than 70 kW. Presented test data show temperature distributions in the High Temperature Test Facility core, upper plenum, cross duct, or lower plenum. Based on these temperature profiles attempts to investigate stratified flow, natural convection flow, heat up, cooldown and mixing phenomena are made. Furthermore, this paper evaluates the performed test campaign in the light of the Very High Temperature Gas-cooled Reactor Phenomena Identification and Ranking Table (PIRT) and proposes experiments to complement the existing PCC/DCC testing database for the validation of the thermal-hydraulic codes.

Reconfigurable Large-Current and High-Voltage Test Bench for HVDC Circuit Breaker Verification

This article presents a novel circuit configuration of a high-voltage direct current circuit breaker (HVdcCB) test bench that is based on a modified H-bridge modular multilevel cascaded converter (MMCC). The modified MMCC is composed of fewer H-bridge cells, and it can be reconfigured during operation to allow the proposed test bench to output large current or high voltage for the current breaking and dielectric withstand tests. Although simultaneous output of large current and high voltage is not possible, the maximum transient interrupt voltage (TIV) withstand test can be performed with reduced ratings. The controllable output allows generation of complex waveforms to simulate a wide range of fault conditions. Furthermore, the modified MMCC has some inherent safety features that can reduce the need for additional protective equipment in case of operational failure of the HVdcCB. In contrast, the conventional charged capacitor and inductor-based designs cannot generate arbitrary waveforms, are only suitable for current breaking tests, and require additional circuits to generate initial conditions for the HVdcCB. AC short-circuit generator-based designs offer one degree of freedom for control of the output waveforms and can sustain the maximum TIV withstand test. However, the ac output is unsuitable for the dielectric withstand test, and additional circuits are required to provide initial conditions for the HVdcCB. The proposed test bench circuit configuration is verified using a downscaled experimental test bench that consists of a total of nine H-bridge cells with an equivalent switching frequency of 92.5 kHz.

Validation of Feedback Reactivity Evaluation Models for Plant Dynamics Analysis Code During Unprotected Loss of Heat Sink Event in Sodium-Cooled Fast Reactors

Abstract Feedback reactivity caused by core deformation is one of the inherent safety features in a sodium-cooled fast reactor (SFR). To validate the evaluation models of the reactivity feedback equipped in the in-house plant dynamics analysis code, named Super-COPD, we have been conducting the benchmark analyses for the unprotected loss of heat sink (ULOHS) tests of balance of plant (BOP)-302R and BOP-301 as one of the representative issues in Experimental Breeder Reactor-II (EBR-II), a pool-type experimental SFR in the United States. During the transient of the ULOHS tests, the reactor power decreased to the decay heat level due to the negative reactivity caused by the radial expansion of the core. Developing the evaluation models of the reactivity feedback is essential for predicting the plant behavior in unprotected events, such as ULOHS events, and the numerical analyses modeling the feedback reactivity were conducted. By comparing the numerical results and the experimental data, the increment temperature profiles of the core inlet and the decreasing reactor power profile calculated by Super-COPD were comparable with experimental data. The applicability of the evaluation models for the reactivity feedback was indicated during the ULOHS event. In addition, through the sensitivity analyses on the cold pool model, it was found that the modification of the plenum model of the cold pool to take into account the thermal stratification was required to improve the core inlet temperature profile.

A novel approach on designing ultrahigh burnup metallic TWR fuels: Upsetting the current technological limits

The grand challenge of “net-zero carbon” emission calls for technological breakthroughs in energy production. The traveling wave reactor (TWR) is designed to provide economical and safe nuclear power and solve imminent problems, including limited uranium resources and radiotoxicity burdens from back-end fuel reprocessing/disposal. However, qualification of fuels and materials for TWR remains challenging and it sets an “end of the road” mark on the route of R&D of this technology. In this article, a novel approach is proposed to maneuver reactor operations and utilize high-temperature transients to mitigate the challenges raised by envisioned TWR service environment. Annular U-50Zr fuel and oxidation dispersion strengthened (ODS) steels are proposed to be used instead of the current U-10Zr and HT-9 ferritic/martensitic steels. In addition, irradiation-accelerated transport of Mn and Cr to the cladding surface to form a protective oxide layer as a self-repairing mechanism was discovered and is believed capable of mitigating long-term corrosion. This work represents an attempt to disruptively overcome current technological limits in the TWR fuels.Impact statementAfter the Fukushima accident in 2011, the entire nuclear industry calls for a major technological breakthrough that addresses the following three fundamental issues: (1) Reducing spent nuclear fuel reprocessing demands, (2) reducing the probability of a severe accident, and (3) reducing the energy production cost per kilowatt-hour. An inherently safe and ultralong life fast neutron reactor fuel form can be such one stone that kills the three birds. In light of the recent development findings on U-50Zr fuels, we hereby propose a disruptive, conceptual metallic fuel design that can serve the following purposes at the same time: (1) Reaching ultrahigh burnup of above 40% FIMA, (2) possessing strong inherent safety features, and (3) extending current limits on fast neutron irradiation dose to be far beyond 200 dpa. We believe that this technology will be able to bring about revolutionary changes to the nuclear industry by significantly lowering the operational costs as well as improving the reactor system safety to a large extent.Graphical abstract

Molten Salt Reactor thermal-fluid dynamics evaluation using a CFD code for a theoretical power density distribution

The molten salt reactors are a kind of nuclear reactor that had received additional attention worldwide due to their inherent safety features and the possibility of use the Thorium, a more abundant resource than Uranium, as a fuel. Despite of this promise concept and works being developed or under course, some aspects related to their thermal hydraulics should be better understand. In this work, the initial thermal-hydraulics results obtained using the ANSYS-CFXÒ code are presented. The simulations were conducted in 3D for 1/16 of a cylinder volume that represents the molten reactor core. The results show a good agreement for this simplified geometry and for a theoretical power density distribution between the benchmark work and those obtained within this for the temperature. An evaluation for the other properties and for different turbulence models, among others, are planned for future.

PO113 / #853 BRAIN INTERCHANGE BCI NEURAL INTERFACING SYSTEM: NEURAL RECORDING AND STIMULATION USING MICRO-ECOG AND DBS ELECTRODES IN SHEEP: E-POSTER VIEWING

The rising use of active neural implants and bi-directional brain-computer interfaces (BCIs) in translational and investigational therapeutical research and medical applications is emerging in today's neuroscientific landscape. Fully implantable neural interfaces and BCI devices must provide a broad range of technical flexibility, long-term implantation potential, and medical safety to the user and to patients to meet the complex requirements for potential medical applications. The system employed here meets the aforementioned requirements: recording configuration choice, open- and closed-loop stimulation functionality, flexible electrode type choice, inherent safety features and biocompatible materials.