An Experimental Observation of the Thermal Effects and NO Emissions during Dissociation and Oxidation of Ammonia in the Presence of a Bundle of Thermocouples in a Vertical Flow Reactor

Ammonia (NH3) dissociation and oxidation in a cylindrical quartz reactor has been experimentally studied for various inlet NH3 concentrations (5%, 10%, and 15%) and reactor temperatures between 700 K and 1000 K. The thermal effects during both NH3 dissociation (endothermic) and oxidation (exothermic) were observed using a bundle of thermocouples positioned along the central axis of the quartz reactor, while the corresponding NH3 conversions and nitrogen oxides emissions were determined by analysing the gas composition of the reactor exit stream. A stronger endothermic effect, as indicated by a greater temperature drop during NH3 dissociation, was observed as the NH3 feed concentration and reactor temperature increased. During NH3 oxidation, a predominantly greater exothermic effect with increasing NH3 feed concentration and reactor temperature was also evident;however, it was apparent that NH3 dissociation occurred near the reactor inlet, preceding the downstream NH3 and H2 oxidation. For both NH3 dissociation and oxidation, NH3 conversion increased with increasing temperature and decreasing initial NH3 concentration. Significant levels of NOX emissions were observed during NH3 oxidation, which increased with increasing temperature. From the experimental results, it is speculated that the stainless-steel in the thermocouple bundle may have catalysed NH3 dissociation and thus changed the reaction chemistry during NH3 oxidation.

Nanotechnology in Nuclear Reactors: Innovations in Fusion and Fission Power Generation

This article explores the transformative potential of nanotechnology and MMs(memory metals)in enhancing the design and operation of nuclear reactors,encompassing both fission and fusion technologies.Nanotechnology,with its ability to engineer materials at the atomic scale,offers significant improvements in reactor safety,efficiency,and longevity.In fission reactors,nanomaterials enhance fuel rod integrity,optimize thermal management,and improve in-core instrumentation.Fusion reactors benefit from nanostructured materials that bolster containment and heat dissipation,addressing critical challenges in sustaining fusion reactions.The integration of SMAs(shape memory alloys),or MMs,further amplifies these advancements.These materials,characterized by their ability to revert to a pre-defined shape under thermal conditions,provide self-healing capabilities,adaptive structural components,and enhanced magnetic confinement.The synergy between nanotechnology and MMs represents a paradigm shift in nuclear reactor technology,promising a future of cleaner,more efficient,and safer nuclear energy production.This innovative approach positions the nuclear industry to meet the growing global energy demand while addressing environmental and safety concerns.

Thermal–hydraulic analysis of space nuclear reactor TOPAZ-Ⅱ with modified RELAP5

With the advantages of high reliability, power density, and long life, nuclear power reactors have become a promising option for space power. In this study, the Reactor Excursion and Leak Analysis Program 5(RELAP5), with the implementation of sodium–potassium eutectic alloy(NaK-78) properties and heat transfer correlations, is adopted to analyze the thermal–hydraulic characteristics of the space nuclear reactor TOPAZ-Ⅱ.A RELAP5 model including thermionic fuel elements(TFEs), reactor core, radiator, coolant loop, and volume accumulator is established. The temperature reactivity feedback effects of the fuel, TFE emitter, TFE collector,moderator, and reactivity insertion effects of the control drums and safety drums are considered. To benchmark the integrated TOPAZ-Ⅱ system model, an electrical ground test of the fully integrated TOPAZ-Ⅱ system, the V-71 unit,is simulated and analyzed. The calculated coolant temperature and system pressure are in acceptable agreement with the experimental data for the maximum relative errors of 8 and 10%, respectively. The detailed thermal–hydraulic characteristics of TOPAZ-Ⅱ are then simulated and analyzed at the steady state. The calculation results agree well with the design values. The current work provides a solid foundation for space reactor design and transient analysis in the future.

Core and blanket thermal-hydraulic analysis of a molten salt fast reactor based on coupling of OpenMC and OpenFOAM

In the core of a molten salt fast reactor(MSFR),heavy metal fuel and fission products can be dissolved in a molten fluoride salt to form a eutectic mixture that acts as both fuel and coolant.Fission energy is released from the fuel salt and transferred to the second loop by fuel salt circulation.Therefore,the MSFR is characterized by strong interaction between the neutronics and the thermal hydraulics.Moreover,recirculation flow occurs,and nuclear heat is accumulated near the fertile blanket,which significantly affects both the flow and the temperature fields in the core.In this work,to further optimize the conceptual geometric design of the MSFR,three geometries of the core and fertile blanket are proposed,and the thermal-hydraulic characteristics,including the three-dimensional flow and temperature fields of the fuel and fertile salts,are simulated and analyzed using a coupling scheme between the open source codes OpenMC and OpenFOAM.The numerical results indicate that a flatter core temperature distribution can be obtained and the hot spot and flow stagnation zones that appear in the upper and lower parts of the core center near the reflector can be eliminated by curving both the top and bottom walls of the core.Moreover,eight cooling loops with a total flow rate of0.0555 m3 s-1 ensur an acceptable temperature distribusure an acceptable temperature distribution in the fertile blanket.

Natural gas pyrolysis in double-walled reactor tubes using thermal plasma or concentrated solar radiation as external heating source

The thermal pyrolysis of natural gas as a clean hydrogen production route is examined. The concept of a double-walled reactor tube is proposed and implemented. Preliminary experiments using an external plasma heating source are carried out to validate this concept. The results point out the efficient CH4 dissociation above 1850 K （CH4 conversion over 90%） and the key influence of the gas residence time. Simulations are performed to predict the conversion rate of CH4 at the reactor outlet, and are consistent with experimental tendencies. A solar reactor prototype featuring four independent double-walled tubes is then developed. The heat in high temperature process required for the endothermic reaction of natural gas pyrolysis is supplied by concentrated solar energy. The tubes are heated uniformly by radiation using the blackbody effect of a cavity-receiver absorbing the concentrated solar irradiation through a quartz window. The gas composition at the reactor outlet, the chemical conversion of CH4, and the yield to H2 are determined with respect to reaction temperature, inlet gas flow-rates, and feed gas composition. The longer the gas residence time, the higher the CH4 conversion and H2 yield, whereas the lower the amount of acetylene. A CH4 conversion of 99% and H2 yield of about 85% are measured at 1880 K with 30% CH4 in the feed gas （6 L/min injected and residence time of 18 ms）, A temperature increase from 1870 K to 1970 K does not improve the H2 yield.

Effect of weld microstructure on brittle fracture initiation in the thermallyaged boiling water reactor pressure vessel head weld metal

Effects of the weld microstructure and inclusions on brittle fracture initiation are investigated in a thermally aged ferritic high-nickel weld of a reactor pressure vessel head from a decommissioned nuclear power plant.As-welded and reheated regions mainly consist of acicular and polygonal ferrite,respectively.Fractographic examination of Charpy V-notch impact toughness specimens reveals large inclusions(0.5-2.5μm)at the brittle fracture primary initiation sites.High impact energies were measured for the specimens in which brittle fracture was initiated from a small inclusion or an inclusion away from the V-notch.The density,geometry,and chemical composition of the primary initiation inclusions were investigated.A brittle fracture crack initiates as a microcrack either within the multiphase oxide inclusions or from the debonded interfaces between the uncracked inclusions and weld metal matrix.Primary fracture sites can be determined in all the specimens tested in the lower part of the transition curve at and below the 41-J reference impact toughness energy but not above the mentioned value because of the changes in the fracture mechanism and resulting changes in the fracture appearance.

Effect of Pre-Deformation Enhanced Thermal Aging on Precipitation and Microhardness of a Reactor Pressure Vessel Steel

Microstructure evolution in neutron irradiated Reactor Pressure Vessel （RPV） steels was experimentally simulated through an improved degradation procedure in this study. The degradation procedure includes austenitizing at 1 150℃ and water quench, deformation 10% and 30% respectively, and then thermal aging at 500℃ for different period of time. The microstructure of the specimens was analyzed in details using transmission electron microscopy （TEM）. The micro-hardness test results showed that all the hardness curves of undeformed, 10% pre-deformed and 30% pre-deformed specimens have two micro-hardness peaks with the first peak value corresponding to different thermal aging time of 1 hour, 5 hours and 10 hours, respectively. It was revealed that the hardness curves were influenced by the precipitation of Cu-rich precipitates （CRPs） and carbides, deposition of martensite and work hardening.

Steady thermal hydraulic analysis for a molten salt reactor

The Molten Salt Reactor (MSR) can meet the demand of transmutation and breeding. In this study, theoretical calculation of steady thermal hydraulic characteristics of a graphite-moderated channel type MSR is conducted. The DRAGON code is adopted to calculate the axial and radial power factor firstly. The flow and heat transfer model in the fuel salt and graphite are developed on basis of the fundamental mass, momentum and energy equations. The results show the detailed flow distribution in the core, and the temperature profiles of the fuel salt, inner and outer wall in the nine typical elements along the axial flow direction are also obtained.

Numerical investigations of thermal mixing performance of a hot gas mixing structure in high-temperature gas-cooled reactor

A numerical simulation study was performed to clarify the thermal mixing characteristics of coolant in the core bottom structure of the high-temperature gas-cooled reactor(HTR). The flow field and temperature field in the hot gas chamber and the hot gas duct of the HTR were obtained based on the commercial computational fluid dynamics(CFD) program. The numerical simulation results showed that the helium flow with different temperatures in the hot gas mixing chamber and the hot gas duct mixed intensively, and the mixing rate of the temperature in the outlet of the hot gas duct reached 98 %. The results indicated many large-scale swirling flow structures and strong turbulence in the hot gas mixing chamber and the entrance of the hot gas duct, which were responsible for the excellent thermal mixing of the hot gas chamber and the hot gas duct. The calculated results showed that the temperature mixing rate of the hot gas chamber decreased only marginally with increasing Reynolds number.

Simulation of radiation dose distribution and thermal analysis for the bulk shielding of an optimized molten salt reactor

The Chinese Academy of Science has launched a thorium-based molten-salt reactor(TMSR)research project with a mission to research and develop a fission energy system of the fourth generation.The TMSR project intends to construct a liquid fuel molten-salt reactor(TMSR-LF),which uses fluoride salt as both the fuel and coolant,and a solid fuel molten-salt reactor(TMSR-SF),which uses fluoride salt as coolant and TRISO fuel.An optimized 2 MWth TMSR-LF has been designed to solve major technological challenges in the Th-U fuel cycle.Preliminary conceptual shielding design has also been performed to develop bulk shielding.In this study,the radiation dose and temperature distribution of the shielding bulk due to the core were simulated and analyzed by performing Monte Carlo simulations and computational fluid dynamics(CFD)analysis.The MCNP calculated dose rate and neutron and gamma spectra indicate that the total dose rate due to the core at the external surface of the concrete wall was 1.91μSv/h in the radial direction,1.16μSv/h above and 1.33μSv/h below the bulk shielding.All the radiation dose rates due to the core were below the design criteria.Thermal analysis results show that the temperature at the outermost surface of the bulk shielding was 333.86 K,which was below the required limit value.The results indicate that the designed bulk shielding satisfies the radiation shielding requirements for the 2 MWth TMSR-LF.

Optical Emission Spectroscopic Study During the Evaporation of Aluminium in the Thermal Plasma Reactor

The oxidation of aluminium was studied using optical emission spectroscopy （OES） during the evaporation of aluminium in traces of oxygen in a thermal plasma reactor. The ratio of the measured line intensities of Al-O with that of Al follows the exact trend as of that obtained from the corresponding line intensities in X-ray diffraction spectra of the synthesized samples. In this paper the inherent capacity of emission spectroscopy in evaluating the growth processes under plasma induced reactions is presented.

Progressive Thermalization Fusion Reactor Able to Produce Nuclear Fusions at Higher Mechanical Gain

In the standard fusion reactors, mainly tokamaks, the mechanical gain obtained is below 1. On the other hand, there are colliding beam fusion reactors, for which, the not neutral plasma and the space charge limit the number of fusions to a very small number. Consequently, the mechanical gain is extremely low. The proposed reactor is also a colliding beam fusion reactor, configured in Stellarator, using directed beams. D+/T+ ions are injected in opposition, with electrons, at high speeds, so as to form a neutral beam. All these particles turn in a magnetic loop in form of figure of “0” (“racetrack”). The plasma is initially non-thermal but, as expected, rapidly becomes thermal, so all states between non-thermal and thermal exist in this reactor. The main advantage of this reactor is that this plasma after having been brought up near to the optimum conditions for fusion (around 68 keV), is then maintained in this state, thanks to low energy non-thermal ions (≤15 keV). So the energetic cost is low and the mechanical gain (Q) is high (>>1). The goal of this article is to study a different type of fusion reactor, its advantages (no net plasma current inside this reactor, so no disruptive instabilities and consequently a continuous working, a relatively simple way to control the reactor thanks to the particles injectors), and its drawbacks, using a simulator tool. The finding results are valuable for possible future fusion reactors able to generate massive energy in a cleaner and safer way than fission reactors.

Thermal Hydraulic Analysis Improvement for the IEA-R1 Research Reactor and Fuel Assembly Design Modification

This paper presents the sequence of activities to improve the thermal hydraulic analysis of the IEA-R1 research reactor to operate in safe conditions after power upgrade from 2 to 5 MW and core size reduction from 30 to 24 fuel assemblies. A realistic analysis needs the knowledge of the actual operation conditions (heat flow, flow rates) beyond the geometric data and the uncertainties associated with manufacturing and measures. A dummy fuel assembly was designed and constructed to measure the actual flow rate through the core fuel assemblies and its pressure drop. First results showed that the flow distribution over the core is nearly uniform. Nevertheless, the values are below than the calculated ones and the core bypass flow rate is greater than those estimated previously. Based on this, several activities were performed to identify and reduce the bypass flow, such as reduction of the flow rate through the sample irradiators, closing some unnecessary secondary holes on the matrix plate, improvement in the primary flow rate system and better fit of the core components on the matrix plate. A sub-aquatic visual system was used as an important tool to detect some bypass flow path. After these modifications, the fuel assemblies flow rate increased about 13%. Additional tests using the dummy fuel assembly were carried out to measure the internal flow distribution among the rectangular channels. The results showed that the flow rate through the outer channels is 10% - 15% lower than the internal ones. The flow rate in the channel formed between two adjacent fuel assemblies is an estimated parameter and it is difficult to measure because this is an open channel. A new thermal hydraulic analysis of the outermost plates of the fuel assemblies takes into account all this information. Then, a fuel design modification was proposed with the reduction of 50% in the uranium quantity in the outermost fuel plates. In order to avoid the oxidation of the outermost plates by high temperature, low flow rate, a reduction of 50% in the uranium density in the same ones was shown to be adequate to solve the problem.

Study for ^(228)Th reduction in thermal reactor with Th-U fuel cycls

By using computer code WIMS/CENDL, the effects of some parameters, core configuration such as fuel element structure, neutron flux and burn-up, are discussed in this paper. It is shown that high neutron flux, small fuel rod diameter, large volume ratio of coolant to fuel, seed-blank heterogeneous core arrangement and 231 Pa chemical separation are necessary for reducing 228Th production in reactor.

Applying chemical engineering concepts to non-thermal plasma reactors

Process scale-up remains a considerable challenge for environmental applications of non-thermal plasmas.Undersanding the impact of reactor hydrodynamics in the performance of the process is a key step to overcome this challenge.In this work,we apply chemical engineering concepts to analyse the impact that different non-thermal plasma reactor configurations and regimes,such as laminar or plug flow,may have on the reactor performance.We do this in the particular context of the removal of pollutants by non-thermal plasmas,for which a simplified model is available.We generalise this model to different reactor configurations and,under certain hypotheses,we show that a reactor in the laminar regime may have a behaviour significantly different from one in the plug flow regime,often assumed in the non-thermal plasma literature.On the other hand,we show that a packed-bed reactor behaves very similarly to one in the plug flow regime.Beyond those results,the reader will find in this work a quick introduction to chemical reaction engineering concepts.

Simulation of the Traweling Wave Burning Regime on Epithermal Neutrons

New results of two computer experiments on modeling of superthermal neutron-nuclear combustion of natural uranium for two different flux densities of external neutron source and duration of half a year each are presented. The simulation results demonstrate the dependence of the autowave combustion modes on the parameters of the external source.

低浓度甲烷热氧化流向变换反应器的动态行为及余热回收研究

废弃煤矿的低体积分数甲烷(1%-3%)通常被直接排放到大气中,但其较高的升温潜势带来了严重的环境问题。在流向变换反应器中直接热氧化甲烷是一种有吸引力的解决方案,但潜在的爆炸和不稳定燃烧等风险限制了其应用。阐明低含量甲烷在流向变换反应器中热氧化的动力学行为是开发工业级反应器的基础。为此,采用数值模拟的方法分析了低含量甲烷热氧化流向变换反应器的自热操作边界,深入研究了热空气导出量对流向变换反应器行为的影响。结果显示,甲烷体积分数超过0.2%即可实现自热操作;甲烷体积分数从0.5%提升至3.0%,最高床温仍维持在1200℃左右。当甲烷体积分数超过0.5%,可以回收部分热量;相同甲烷含量条件下,最高床温随着热气抽出量的增加而增加;随着甲烷体积分数从0.5%提升到3.0%,允许导出的热空气从12.5%几乎线性地增加到32%。进一步研究发现,以30-50 s的时间间隔反向流动可以确保甲烷的完全转化和床温稳定。上述结果表明,甲烷体积分数在1%-3%时,采用热氧化处理可以实现余热回收;此外,通过调整换向时间和热空气导出量可以实现床层温度控制。

钠冷快堆小栅板联箱压降对组件流量分配影响研究

钠冷快堆堆芯采用大栅板联箱、小栅板联箱和组件的三级流量分配方式,小栅板联箱的压降影响组件的流量分配,进而影响堆芯的安全,因此进行钠冷快堆小栅板联箱压降对组件流量分配影响研究有重要意义。根据小栅板联箱压降造成组件流量分配偏差的机理,提出了理论计算模型和堆芯组件优化设计的方法,并针对中国实验快堆(CEFR)堆芯进行了组件压降的优化设计,通过优化设计降低了CEFR燃料组件流量分配负偏差。结果表明,在进行钠冷快堆堆芯热工水力设计时,需要结合实际堆芯布置分析组件压降设计值的优化方向,并进行敏感性分析,以确定组件的最优设计压降,将小栅板联箱压降对组件流量分配影响降低到最低程度。本文结果可为钠冷快堆堆芯热工水力设计提供参考。

CAP1400核主泵叶轮动应力计算及疲劳寿命预测

为实现核主泵叶轮疲劳寿命预测,考虑叶轮高温高压的恶劣运行工况建立流-热-固耦合计算模型,应用ANSYS CFX软件对核主泵叶轮内部流动的压力载荷和温度载荷进行非定常数值计算,在ANSYS Workbench中实现载荷向结构的传递,并对叶轮动力响应疲劳载荷开展研究.利用雨流计数法对叶片危险部位的载荷数据进行统计分析,进一步结合Palmgren-Miner理论对核主泵叶轮的最小疲劳寿命周期进行预测.研究结果表明:叶轮在旋转过程中承受周期性交变应力的作用;叶轮叶片进、出口边与前、后盖板交接处容易发生内部应力集中,最大应力出现在叶片出口边与前盖板交接处,为142.57 MPa;叶片各危险部位承受应力波峰和波谷的时间基本一致;叶轮产生的疲劳为应力疲劳,疲劳破坏首先发生在叶片进口边与后盖板交接处;计算得到叶轮的疲劳寿命为277.94 a.研究结果可为叶轮的动态强度优化和疲劳设计提供一定参考.