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.

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.

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.

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.

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.

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.

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.

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.

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.

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。

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.

ODS MA754合金传热界面接触热阻实验研究

鉴于ODS MA754合金传热界面的接触热阻参数对全固态堆芯空间反应堆系统的热量导出具有重要影响,研发和设计了高温高压接触热阻实验装置,测量了不同温度(20~800℃)、压力(0~80 MPa)、气体氛围(He、CO_(2))以及试件表面粗糙度(1.6、3.2μm)下ODS MA754合金传热界面的接触热阻,并基于测试获得的宽量程数据点,建立了ODS MA754合金的接触热阻数据库。实验结果表明:随着接触面温度和压力的升高,界面接触热阻降低,且热阻降低的速率逐渐减小;相较于表面粗糙度为1.6μm的试件,粗糙度为3.2μm试件表面的界面接触热阻明显偏大,实验得到的定量关系可为工程样件的加工粗糙度要求提供依据;He气氛下的接触热阻远小于CO_(2)气氛,在0.1 MPa、100℃工况下,He气氛接触热阻约为CO_(2)气氛接触热阻的1/4。该研究结果可为空间反应堆的热工设计提供数据参考。

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.

不同低温等离子体反应器降解VOCs的研究进展

挥发性有机物(VOCs)可对人体及自然环境造成严重危害,低温等离子体技术适用范围广且能在常温常压下分解VOCs.对不同类型的低温等离子体反应器进行分类,并关注其对降解效果的影响,结果表明:在配电参数和配气参数不变的情况下,线-筒式电晕放电反应器和电极的直径、线-板式电晕放电反应器的线线间距和线板间距对降解效率影响较大;圆筒式介质阻挡放电反应器在性能上优于平板式介电阻挡放电反应器;等离子体内催化反应器降解VOCs的能力优于等离子体后催化反应器,但在副产物控制方面不如等离子体后催化反应器;催化剂的选择在VOCs的降解中有重要影响.最后,对等离子体反应器的发展进行了展望.

液态金属冷却快堆子通道分析软件SACOS-LMR研发与工程应用

子通道分析方法是反应堆堆芯设计和热工水力分析的重要手段之一,对于我国提出的压水堆-快堆-聚变堆三步走核能发展战略,开发适用于液态金属冷却快堆热工安全分析的子通道分析程序具有重要意义。本文基于西安交通大学热工水力研究室自主开发的压水堆子通道程序SACOS,通过添加液态金属快堆特有的模型,如绕丝模型、盒间流模型、液态金属对流换热模型等,扩展至适用于液态金属快堆的子通道分析程序SACOS-LMR,该程序具备对液态金属快堆组件开展稳态和瞬态热工水力分析的功能。结合卡尔斯鲁厄开展的37棒钠冷瞬态实验,完成了SACOS-LMR程序的瞬态功能验证。基于验证后的SACOS-LMR程序,对欧洲铅冷快堆(ALFRED)堆芯开展了稳态工况和瞬态事故工况下的热工安全特性分析,计算结果合理,且与同类程序保持一致,表明SACOS-LMR程序可用于液态金属快堆的堆芯设计和热工水力分析研究。

氦氙气冷小堆燃料棒辐射散热特性分析

反应堆堆芯热工安全特性是反应堆系统安全特性最重要的组成部分。氦氙气冷小堆中燃料棒间温差大,辐射散热不可忽略。因此针对氦氙气冷小堆堆芯,不止需要考虑对流换热和热传导两种换热形式,也需要考虑辐射散热对堆芯换热的影响。本文通过建立氦氙气冷小堆堆芯1∶1精细化模型,开展反应堆堆芯内部三维辐射散热特性分析,同时获得辐射角系数。将本文模型计算结果与蒙特卡罗方法结果进行对比验证,两者相对误差小于1%,证明该辐射角系数计算方法的准确性。开展了燃料棒径距比、燃料棒表面温度、燃料棒长度等几何参数敏感性分析,关注其对燃料棒辐射散热特性的影响,并开发出一套具有普适性的堆芯内燃料棒辐射角系数经验关系式。本文通过研究燃料棒间辐射散热特性了解反应堆内辐射散热规律,为后续反应堆堆芯热工安全特性研究提供了技术支撑。

高导热环氧复合材料干式电抗器热点温升的仿真研究

干式电抗器的稳定运行影响新型电力系统的输电可靠性。干式空心电抗器包封材料整体由浸有环氧树脂的玻璃纤维丝经高温固化而成。本文采用多物理场耦合有限元方法,考虑干式空心电抗器的包封材料热导率对其热点温升的影响,建立了环氧复合材料的COMSOL微观仿真模型和外电路约束下的干式空心电抗器电-磁、流-热耦合计算模型。将电磁场下的损耗作为热源计算温度场与流场分布,研究在25℃环境温度下常规/高导热环氧复合材料对干式空心电抗器热点温升的影响规律。结果表明:高导热环氧树脂对复合材料热导率的提升效果显著;包封材料本体及周围空气温度场区域中热点温升最大值为103.75℃,出现在内部第4层包封材料的上端处;不同热导率的复合材料对降低干式电抗器的热点温升有明显差异,其中干式电抗器在高导热环氧树脂复合材料下的热点温度降低了7.55℃。