Reactor Power System Research Articles

Numerical analysis of dynamic load following response in a natural circulation molten salt power reactor system

The Molten Salt Reactor (MSR) concept is a rapidly evolving Generation IV design that has recently attracted favorable attention due to the potential for reducing waste generation, realizing passive safety features, and seizing on the opportunity for cost effective economics. An investigation into the power transient behavior of an autonomous load-following, closed-loop, natural circulation MSR system is important to quantifying operational and safety performance under dynamic conditions. This paper presents the results of a STAR-CCM+ and a comparatively simple asymmetric, one-dimensional, numerical model to solve the compound dynamic MSR power behavior subject to flow and temperature reactivity feedback only. Results show that reactor power is affected by fuel salt flow velocity (global) and temperatures (local bulk volume averages) in a coupled, time-delayed manner that results in a unique compound dynamic, closed-loop power feedback mechanism. The 1-D simulation approach opens the possibility of performing inexpensive computations to evaluate time-dependent reactor performance relative to thermo-physical fuel salt limitations. Results also potentially motivate a convincing conclusion that natural circulation MSRs provide a leap in safety and reliability.

Improvement and Validation of the System Analysis Model and Code for Heat-Pipe-Cooled Microreactor

Heat-pipe-cooled microreactors (HPMR) use a passive high-temperature alkali metal heat pipe to directly transfer the heat of solid core to the hot end of the intermediate heat exchanger or thermoelectric conversion device, thus avoiding a single point failure. To analyze and evaluate the transient safety characteristics of an HPMR system under accident conditions, such as heat pipe failure in the core or a loss of system heat sink and other accidents, a previously developed model for transient analysis of a heat-pipe-cooled space nuclear reactor power system (HPSR) was improved and validated in this study. The models improved mainly comprise: (1) An entire 2-D solid-core heat transfer model is established to analyze the accident conditions of core heat pipe failure and system heat sink loss. In this model, radial and axial Fourier heat conduction equations are used to divide the core into r-θ direction control volumes. The physical parameters of the material in the control volume are calculated according to the volume-weighted average. (2) By coupling the heat transfer limit model and the two-dimensional thermal resistance network model, the transient model of a heat pipe for HPMR system analysis is improved. (3) Conversion system models are established to simulate the system characteristics of the advanced HPMR concept, such as thermoelectric conversion, Stirling conversion, and the open Brayton conversion analysis model. Based on the improved models, the HPMR system analysis program TAPIRSD was developed, which was verified by experimental data of the separated conversion components and the ground nuclear test device KRUSTY. The maximum deviation of the power output predicted by the energy conversion model is less than 8%. The accident conditions of the KRUSTY tests, such as load change, core heat pipe failure, and heat sink loss accident, were studied by using TAPIRSD. The results show that the simulation results of the TAPIRSD code agree well with the experimental data of the KRUSTY prototype reactor. The maximum error between the TAPIRSD code prediction and the measured value of the core temperature under accident conditions is less than 10 K, and the maximum deviation is less than 2%. The results show that the developed code can predict the transient response process of the HPMR system well. At the same time, the accuracy and reliability of the improved model are proved. The TAPIRSD is suitable for system transient analysis of different types of HPMRs and provides an optional tool for the system safety characteristics analysis of HPMR.

Development of a Performance Analysis Model for Free-Piston Stirling Power Convertor in Space Nuclear Reactor Power Systems

Space nuclear reactor power system (SNRPS) is a priority technical solution to meet the future space power requirement of high-power, low-mass, and long-life. The thermoelectric conversion subsystem is the key component of SNRPS, which greatly affects the performance, quality, and volume of SNRPS. Among all kinds of proposed thermoelectric conversion technologies, the free-piston Stirling power converter (FPSPC) has become a preferred conversion technology for small-scale advanced SNPRS due to its moderate waste heat emission temperature and high conversion efficiency, mainly composed of a linear alternator and free-piston Stirling engine (FPSE). For studying the performance of FPSPC, a quasi-steady flow thermodynamic cycle analysis model considering parasitic heat losses has been developed for FPSE. And then the performance analysis model for FPSPC has been established by coupling the thermodynamic cycle analysis model with the mechanical motion model of the piston and volt-ampere characteristic model of the linear alternator. Furthermore, the analysis model was compared and validated by the GPU-3 Stirling engine’s experimental data. The performance parameters of Component Test Power Converter (CTPT) FPSPC designed by NASA for SNRPS were also analyzed. The results show that the amplitudes position of CTPC displacer and piston are 15.1 mm and 11.2 mm, respectively. The corresponding average electric power output of CTPC is 17.316 kW. The input thermal power to the CTPT heater is 66.1 kW, leading to the converter efficiency of 26.2%. The average current and voltage of the CTPC alternator are 86.38 A and 193.15 V, respectively. Among all kinds of parasitic energy losses, the regenerator heat loss accounts for the largest proportion, with an average of about 12.7 kW. The effects of cooler and heater temperature on the performance of CTPC FPSPC were also studied.

Concept of a fast breeder reactor to transmute MAs and LLFPs

The long-term issues of nuclear power systems are the effective use of uranium resources and the reduction of radioactive waste. Important radioactive wastes are minor actinides (MAs: 237Np, 241Am, 243Am, etc.) and long-lived fission products (LLFPs: 129I, 99Tc, 79Se, etc.). The purpose of this study was to show a concept that can simultaneously achieve the breeding of fissile materials and the transmutation of MAs and LLFPs in one fast reactor. Transmutation was carried out by loading innovative Duplex-type MA fuel in the core region and LLFP-containing moderator in the first layer of the radial blanket. Breeding was achieved in the core and axial blanket. As a result, it was clarified that in this fast breeder reactor, a breeding ratio of approximately 1.1 was obtained, and MAs and LLFPs achieved a support ratio of 1 or more. The transmutation rate was 10.3%/y for 237Np, 14.1%/y for 241Am, 9.9%/y for 243Am, 1.6%/y for 129I, 0.75%/y for 99Tc, and 4%/y for 79Se. By simultaneously breeding fissile materials and transmuting MAs and LLFPs in one fast reactor, it will be possible to solve the long-term issues of the nuclear power reactor system, such as securing nuclear fuel resources and reducing radioactive waste.

Robust controller design for small pressurized water reactors using non-smooth optimization

The flexible operating conditions of small pressurized water reactors (SPWRs) under parameter uncertainties require highly robust control systems. This paper presents a robust controller design for an SPWR using non-smooth optimization. First, nonlinear mathematical models of the SPWR nuclear steam supply system were adopted and linearized to obtain the corresponding transfer function models. Then, sensitivity analysis of the SPWR physical and thermodynamic parameters was conducted to determine the critical uncertainty parameters for the robust controller design. Subsequently, the reactor power and steam generator feedwater robust control systems were designed under large uncertainties in the identified critical parameters. Finally, simulation studies of typical transients for the SPWR under high- and low-speed operating conditions of the main pump were performed. The results indicate that the designed robust control systems for the SPWR exhibit good performance and excellent robustness under large parameter uncertainties during various load change transients.

Parametric sensitivity analysis of liquid metal helical coil once-through tube steam generator

Lead-bismuth Fast Reactor (LFR) has attracted attentions due to its intrinsic characteristics, such as sustainability, safety, and economics. Helical coil once-through tube steam generator (HCOTSG) is a proposed form of steam generator. Due to its unique advantages, HCOTSG is widely used in various reactor power systems, including LMFR. In this paper, FLUENT CFD commercial code is adopted to simulate heat transfer from the primary side (liquid metal) and secondary side (water/steam). The robustness of the method is validated by comparing it with relevant experimental research, and the maximum error is less than 25%. Based on this method, 9 models with different geometrical parameters were established to analyze the effects of geometrical parameters on the performance of liquid metal HCOTSG. The wall heat flux and comprehensive performance factor (En) of different models is compared. The optimal geometric arrangement scheme for the working conditions is recommended. The geometrical parameters of the tube side and shell side are analyzed as sensitivity parameters. Among the established geometric models, the HCOTSG has the best comprehensive performance when Pr /D is 0.167 and Pa /D is 0.137 under the working conditions in this paper. This study provides a numerical simulation method on structural design optimization of liquid metal HCOTSG.

Reactor core design of UPR-s: A nuclear reactor for silence thermoelectric system NUSTER

The heat pipe reactor is featured with the passive heat discharge, anti-single point failure and silent operation. Therefore, it could be applied for the power system of underwater vehicles. In this paper, the concept of Unattended Uranium-fueled Portable Reactor (UPR) for the NUclear Silence ThermoElectric Reactor (NUSTER) power system is introduced and the reactor core with a thermal power of 1 MW is neutronics designed. In the core, highly enriched UO2 fuel, sodium heat pipe, molybdenum alloy matrix, beryllium oxide and beryllium reflector are adopted and the reactivity is controlled by four sliding reflectors and four safety rods. In design process, the NECP-MCX code is used to carry out the neutronics calculation. At the same time, the heating unit, core loading, control system and power distribution of the core are described. Furthermore, criticality safety, burnup characteristics, power distribution of the core are specifically analyzed. The results show that the proposed core design meets the requirements of criticality safety and operating life.

Dynamic simulation of a space gas-cooled reactor power system with a closed Brayton cycle

Dynamic simulation of a space gas-cooled reactor power system with a closed Brayton cycle

Impact of axial power distribution on thermal-hydraulic characteristics for thermionic reactor

Reactor fuel's power distribution plays a vital role in designing the new generation thermionic Space Reactor Power Systems (SRPS). In this paper, the 1/12th SPACE-R's full reactor core was numerically analyzed with two kinds of different axial power distribution, to identify their impacts on thermal-hydraulic and thermoelectric characteristics. In the benchmark study, the maximum error between numerical results and existing data or design values ranged from 0.2 to 2.2%. Four main conclusions were obtained in the numerical analysis: a) The axial power distribution has less impact on coolant temperature. b) Axial power distribution influenced the emitter temperature distribution a lot, when the core power was cosine distributed, the maximum temperature of the emitter was 194 K higher than that of the uniform power distribution. c) Comparing to the cosine axial power distribution, the uniform axial power distribution would make the maximum temperature in each component of the reactor core much lower, reducing the requirements for core fuel material. d) Voltage and current distribution were similar to the axial electrode temperature distribution, and the axial power distribution has little effect on the output power.

Research on power flattening method and neutron characteristic analysis of a megawatt-class space gas-cooled fast reactor

According to the versatility and durability requirements of the deep space missions, the space nuclear reactor (SNR) power system is becoming a more suitable choice compared with traditional solar and chemical power systems in the fields with the larger-scale and longer-life applications. For a megawatt-class space gas-cooled reactor, the power peak factor is a major factor that influences the reactor power and safety. In the present paper, two methods are proposed to flatten the power profile of SNR and the corresponding neutron characteristic of the space gas-cooled fast reactor is analyzed. One is to change the material of the reflector, and the other is to add neutron burnable poisons in the middle of the core. The result reveals that when the reactor is with the same initial reactivity, the power peak factor and the reflector mass of the reactor using Be as the reflector material is smaller than that using the BeO. Adding 5 wt% Gd2O3 in the range of the four layers fuel rods in the center of the reactor, the power profile of the reactor can be flattened greatly, and it can also meet the 10-year lifetime requirement. Finally, through calculation and analysis, one reasonable arrangement scheme of the control rod is proposed. The results can lay basis for the design of the large power space nuclear reactor.

A solar thermal storage power generation system based on lunar in-situ resources utilization: modeling and analysis

Continuous energy supply is crucial to the crew and assets of lunar outposts during the darkness lunar night of 350 h in the long term lunar exploration. A solar energy storage power generation system based on in-situ resource utilization (ISRU) is established and analyzed. An efficient linear Fresnel collector is configured for solar concentration. The thermal energy reservoir (TER) coupling with Stirling power generator is designed using the fuel tanks of descent module and lunar regolith. A comprehensively theoretical model based on finite time thermodynamics is developed to analyze the energy flow and efficiency of thermal storage power generation system, and the major irreversibilities are taken into account. The results show that the designed system can produce an average power of 6.5 kW during the lunar night with 19.6% utilization efficiency of collected solar energy in the daytime. The evaluated launch mass of designed power system has a competitive advantage than those of nuclear reactor power and photovoltaic-battery power systems. The influences of major heat transfer processes including heat leakage of TER, heat exchange capability of Stirling engine and radiator are also discussed, which provides physical insight for optimal design of future power generation system based on ISRU.

Design and heat transfer optimization of a 1 kW free-piston stirling engine for space reactor power system

The Free-Piston Stirling engine (FPSE) is of interest for many research in aerospace due to its advantages of long operating life, higher efficiency, and zero maintenance. In this study, a 1-kW FPSE was proposed by analyzing the requirements of Space Reactor Power Systems (SRPS), of which performance was evaluated by developing a code through the Simple Analysis Method. The results of SAM showed that the critical parameters of FPSE could satisfy the designed requirements. The heater of the FPSE was designed with the copper rectangular fins to enhance heat transfer, and the parametric study of the heater was performed with Computational Fluid Dynamics (CFD) software STAR-CCM+. The Performance Evaluation Criteria (PEC) was used to evaluate the heat transfer enhancement of the fins in the heater. The numerical results of the CFD program showed that pressure drop and Nusselt number ratio had a linear growth with the height of fins, and PEC number decreased as the height of fins increased, and the optimum height of the fin was set as 4 mm according to the minimum heat exchange surface area. This paper can provide theoretical supports for the design and numerical analysis of an FPSE for SRPSs.

Influence of non-ideal gas characteristics on working fluid properties and thermal cycle of space nuclear power generation system

Space nuclear Closed Brayton Cycle (CBC) power generation system with Helium-xenon gas (He–Xe) as a work fluid has attracted much attention with the development of space technology. The appropriate thermophysical property model of He–Xe is the key to achieve high-precision simulation of CBC power generation system. The thermophysical property model based on virial coefficient was developed and proposed. It was compared with the ideal gas model to clarify the differences and non-ideal gas characteristics of He–Xe. The deviations of He–Xe thermophysical parameters, main cycle parameters under different temperatures and pressures were compared numerically with these models. Based on non-ideal gas properties, a thermodynamic model was developed to analyze the influence of non-ideal gas characteristics on the efficiency of a 3.0 MWth lithium-cold fast reactor power system. The results showed that He–Xe physical parameters deviated from ideal gas characteristics significantly under the molar mass of more than 40 g/mol. Lower temperature (<500K) or higher pressure (>3.0 MPa) tends to make non-ideal gas characteristics on thermophysical parameters more effective. The thermophysical property model with non-ideal gas characteristics improves the system simulation accuracy about 4.91 %. This work can be used as reference for the development, accurate simulation and assessment of space nuclear CBC power generation system.

Heat Pipe as a Passive Cooling System Driving New Generation of Nuclear Power Plants

The technology of the Heat Pipe (HP) system is very well known for scientists and engineers working in the field of thermal-hydraulic since its invention at Las Alamos Nation Laboratory around the 1960s time frame. It is a passive heat transfer/heat exchanger system that comes in the form of either a constant or variable system without any mechanical built-in moving part. This passive heat transfer system and its augmentation within the core of nuclear power reactors have been proposed in the past few decades. The sodium, potassium, or mercury type heat pipe system using any of these three elements for the cooling system has been considered by many manufacturers of fission reactors and recently fusion reactors particularly Magnetic Confinement Fusion (MCF). Integration of the heat pipes as passive cooling can be seen in a new generation of a nuclear power reactor system that is designed for unconventional application field such as a space-based vehicle for deep space or galaxy exploration, planetary surface-based power plants as well as operation in remote areas on Earth. With the new generation of Small Modular Reactor (SMR) in form of Nuclear Micro Reactors (NMR), this type of fission reactor has integrated Alkali metal heat pipes to a series of Stirling convertors or thermoelectric converters for power generation that would generate anywhere from 13kwt to 3Mwt thermal of power for the energy conversion system.

DESIGN OF A TRISO PARTICLE FUEL BASED INTEGRATED GAS-COOLED SPACE NUCLEAR REACTOR

According to versatile and long-lasting requirements of deep space missions, space nuclear reactor (SNR) power system is becoming a more suitable choice compared to traditional solar and chemical power systems in large-scale and long-life applications. From NASA’s previous research, the gas-cooled reactor along with closed Brayton cycle (CBC) could achieve optimized weight-power ratio and be more applicable for large power system (100 kWe or MWe level). In this paper, a concept of integrated gas-cooled space nuclear reactor named IGCR-200 is introduced, which is designed based on the TRISO particle fuel and could achieve 200 kWe output combined with highly efficient He/Xe CBC generator. The design requirements include an operation lifetime of at least 10 years in full power mode, maximum fuel temperature &lt; 1600K, negative temperature reactivity feedback, passive decay heat removal, redundancy in reactor control, and sub-criticality during water flooding accidents. It has an outer diameter of 70.0 cm, a height of 66.0 cm (reactor part), a total mass around 1000 kg, total Uranium inventory of 226.8 kg (235U enrichment as 93%), and 1 MW thermal power output. The reactor physics, thermal hydraulics and other required analysis are taken out to show the feasibility and performances of the design.

Photoneutron production and characterization in fluoride-salt cooled high-temperature reactors

Due to the presence of beryllium-bearing salts, fluoride-salt cooled high-temperature reactors (FHRs) are expected to have higher production of delayed photoneutrons than other common power reactor systems which may impact transient behavior. New methods for computationally quantifying the characteristics of delayed photoneutrons using Monte Carlo particle transport, including absolute fraction, effective fraction, and group decay constants, are outlined. These methods are validated where possible against existing data from CANDU reactors, and then used to make a best-estimate characterization of the delayed photoneutron effect in a prototypical FHR design. It is found that delayed photoneutrons have a worth of approximately 9 pcm in the reference FHR system, and that their importance is slightly higher than that of fission neutrons due mostly to their softer spectrum. Existing emission group structures from literature have been compared to computational data and found to be lacking, particularly in their representation of groups with short half-lives. Therefore, a new 13-group structure is proposed for use in FHR dynamic calculations.

Thermal‐hydraulic analysis of gas‐cooled space nuclear reactor power system with closed Brayton cycle

Space nuclear reactor power system (SNRPS), especially megawatt power level, is very attractive for the future civil and military space demands. The high-temperature gas-cooled reactor coupled with closed Brayton cycle (CBC) can achieve high-power output for space applications. For the gas-cooled SNRPS with CBC, dynamic models for all components, including reactor, turbine, compressor, alternator, ducting, pump, recuperator, gas cooler and heat pipe radiator, are developed in this article. Then, a transient system analysis code (SAC-SPACE) is developed to perform the safety characteristics analysis of the SNRPS. In the full-power steady-state analysis, the maximum fuel temperature has a margin of 1323 K from the melting point. Moreover, the transient responses of the SNRPS under the mechanical failure of one Brayton loop (MFOBL1) and the reactivity insertion accident (RIA) are simulated and analyzed. During the MFOBL1, the maximum fuel temperature is 1275 K lower than the melting point, and finally, the reactor power stabilizes at 56.8% of the rated power. For the RIA, as long as the reactivity introduced into the core does not exceed 0.61 $, the SNRPS can reach a new steady-state safely by itself without other protection strategies. It can be concluded that the SNRPS has good self-stability due to the negative reactivity feedback of the reactor. This article may provide useful theoretical supports for the design and safety analysis of the gas-cooled SNRPS with CBC.

Analysis of the Jet Engine Propulsion and Thrust relative to the use of the Water Molecule Power Reactor System

As civilization evolves through time, humans had begun to unfold the capabilities of the mind to create things and it had evolved through the great span of time from ancient world to the modern era but human civilization cannot be considered advance yet until it is able to reach the nearest planet from earth. As human had begun to embrace the modern world of digital technology and space travel, it had become a burden for mankind as to how would it be able to reach the moon and the nearest planet from earth. The primary concern for space travel is fuel. The fuel that would be capable of transporting a ship from this planet earth to the next nearest planet in the solar system, the answer to this question might be here on earth as well, the very element that came from space and form our beautiful and amazing planet, it is water. Water is a very vital element to human existence although many take it for granted but water is truly our life and time will come that water will be a very valued commodity in this planet.

Analysis of the Jet Engine Propulsion and Thrust relative to the use of the Water Molecule Power Reactor System

As civilization evolves through time, humans had begun to unfold the capabilities of the mind to create things and it had evolved through the great span of time from ancient world to the modern era but human civilization cannot be considered advance yet until it is able to reach the nearest planet from earth. As human had begun to embrace the modern world of digital technology and space travel, it had become a burden for mankind as to how would it be able to reach the moon and the nearest planet from earth. The primary concern for space travel is fuel. The fuel that would be capable of transporting a ship from this planet earth to the next nearest planet in the solar system, the answer to this question might be here on earth as well, the very element that came from space and form our beautiful and amazing planet, it is water. Water is a very vital element to human existence although many take it for granted but water is truly our life and time will come that water will be a very valued commodity in this planet.

Response of p-type and n-type Half-Heusler thermoelectric materials to fast neutron irradiation

The thermoelectric materials working in radioisotope thermoelectric generators or space reactor power system will be exposed to neutron irradiation in space missions. Herein, the influences of neutron irradiation with doses of 1013n/cm2 and 1015n/cm2 on the transport properties of p- and n-type Half-Heusler thermoelectric alloys were investigated. It was found that, neutron irradiation showed negative effects on thermoelectric properties of p-type Hf0.5Zr0.5CoSb0.85Sn0.15 and p-type Hf0.5Ti0.5CoSb0.85Sn0.15. But neutron irradiation had no effects on thermoelectric properties of n-type Hf0.5Zr0.5NiSn0.985Sb0.015 and n-type ZrNiSn0.985Sb0.015. After 1015n/cm2 neutron irradiation, the electrical conductivity, power factor and ZT of p-type Half-Heusler alloys decreased by 7–22%, which was caused by the decreased carrier concentration. However, the radiation-induced changes in the thermoelectric properties proved to be essentially healed by annealing at 773 K ~ 873 K. The anti-annealing effects were also studied.