核热推进系统热工过程及堆芯关键技术分析

核热推进是利用核反应堆产生的裂变能将氢气加热到高温高压状态,然后从喷管高速喷出产生巨大推力的新型推进方式,具有大推力、高比冲等特点,被认为是未来最有希望实现载人深空探测的技术之一。历史上有固体堆芯、液体堆芯以及气体堆芯等主要设计,其中固体堆芯技术最为成熟。总体来说,核热推进系统中反应堆尺寸较小、堆芯功率密度大、温度较高,因此需要有良好的热工水力设计来保证堆内的热量安全导出。本文通过对美国的NERVA、PBR和MITEE、CERMET堆芯以及苏联的RD-0410核热火箭发动机系统进行简单介绍,归纳总结了主要热工过程,并对这些过程中所涉及到的堆芯关键技术问题进行了分析,为今后我国空间核热推进系统的研究和设计提供一定的借鉴。     

基于超临界二氧化碳布雷顿循环的钠冷快堆方案研究

超临界二氧化碳布雷顿循环因热效率高、布置紧凑等特点受到了广泛的关注,多种核能系统将其列为备选的动力循环系统。为研究基于超临界二氧化碳布雷顿循环的钠冷快堆系统的特点,本文在调研和分析的基础上,从反应堆回路数目、动力循环方式、系统参数选取及设备材料选型等方面开展了分析与对比,针对给定的系统配置方式初步分析了系统主要参数特点,并对应用于钠冷快堆的超临界二氧化碳动力循环系统的发展提出了建议。     

CERMET-SNRE堆芯物理计算分析

核火箭发动机功率高、寿命长、比冲大,在执行深空探测和星际航行任务时具有不可替代的优势。小型化是核火箭发动机的一个重要趋势,基于此提出了一种使用钨基金属陶瓷燃料的小型核火箭发动机（CERMET-SNRE）堆芯方案,并采用蒙特卡罗程序（MCNP）进行了精确建模,计算了相关物理参数。计算分析结果表明：CERMET-SNRE堆芯能谱硬,燃耗浅,后备反应性足够,功率分布合理,控制鼓与安全棒价值足够,发射掉落事故下有效增殖因数小于0.98,堆芯方案合理,满足设计要求。     

核电厂最终热阱冷冻水系统研究

针对特殊地域环境下可能出现的重要厂用水系统(SEC)夏季水温过热的现象,提出了核电厂SEC系统采用机械通风冷却塔二次循环冷却方式及采用直流冷却方式下最终热阱冷冻水系统的选取方案,并对方案进行对比,确定最终热阱冷冻水系统采用板式热交换器的方式。通过理论分析,给出采用板式热交换器方式时冷冻水系统制冷量的确定方法。     

核热推进堆芯方案的发展

核热推进利用核裂变能加热工质,比冲可达化学火箭的2倍多,在空间活动中有广阔的应用前景。在美国和俄罗斯的研究过程中,对多个核热推进堆芯方案进行了较深入的研究。本工作介绍了这些堆芯方案的情况,详细说明了其设计特点,并总结了堆芯方案的发展趋势。     

摇摆对自然循环核热耦合平均功率的影响

通过对摇摆条件下单相自然循环核热耦合的模拟计算,分析研究了摇摆条件下自然循环核热耦合系统平均功率的影响因素。分析结果表明：摇摆条件下,考虑核反馈后的系统平均功率与系统的平均质量流量、系统的平均传热系数成正比,与慢化剂与燃料的温度反馈系数比成反比。摇摆参数对系统平均功率的影响和慢化剂与燃料的温度反馈系数比有关。当系统平均传热系数的变化对反应性的影响起主导作用时,摇摆越剧烈,系统平均功率越大;当平均阻力系数的变化对反应性的影响起主导作用时,摇摆越剧烈,系统平均功率越小。     

应用于钠冷快堆的超临界二氧化碳动力转换系统研究

超临界二氧化碳（SCO₂）布雷顿循环由于高效、紧凑和可避免钠水反应等特性而成为钠冷快堆的理想动力转换系统。本文以1 200 MWe大型池式钠冷快堆为系统热源,钠回路温度及热负荷为循环系统运行边界,对比研究了不同SCO₂布雷顿循环系统性能和关键设备性能的变化规律。研究发现,级间冷却再压缩循环与钠冷快堆热源特性匹配性最佳,且循环效率最高（40.7%）。进而研究了不同运行参数对级间冷却再压缩循环效率的影响规律,给出了循环系统效率对各关键影响因素的敏感度,发现循环系统效率对冷端参数的敏感度最强,其次为分流比和透平入口参数,对主压缩机级间压比的敏感度最弱。     

摇摆运动下单相自然循环核热耦合特性研究

采用考虑6组缓发中子的点堆中子动力学模型,开发了核反馈模拟模块,并将之与摇摆条件下单相自然循环热工水力计算模型进行合并,基于Matlab软件编制了相应的计算程序,实现了摇摆条件下单相自然循环核热耦合的模拟计算。计算结果表明：摇摆条件下,与不考虑核反馈相比,考虑核反馈后核热耦合效应使系统流量降低,系统功率产生波动;系统功率的平均值随摇摆频率及振幅的增大而降低,而系统功率的振幅则随摇摆周期及振幅的增大而增大。核热耦合效应使燃料元件温度的波动振幅减小,起到了抑制燃料温度波动的作用。     

纵摇和横摇对自然循环的影响

针对分散布置的核船压水堆一回路的特点，分析了船舶纵摇和横摇对自然循环能力的不同影响。以某核船参数为例，编程计数比较了两种摇摆情况下自然循环能力的变化情况。     

摇摆下自然循环矩形双通道系统核热耦合不稳定性研究

将海洋条件热工水力分析程序RELAP5/MC与三维物理瞬态输运程序TDOT-T采用并行方式耦合,对摇摆条件下自然循环矩形双通道系统核热耦合不稳定性进行计算分析。结果表明,系统存在同相和异相2种振荡模式,分别由摇摆运动和密度波振荡(DWO)引起。核反馈对第1类DWO和两相区的同相振荡有抑制作用,但对第2类DWO和单相区的同相振荡几乎没有影响。基于非线性理论对计算结果进行分析,发现耦合核反馈后系统非线性增强,由于摇摆导致系统流量波动与DWO叠加,其现象非常复杂,摇摆条件下的核热耦合不稳定性会出现非线性振子耦合中的同步化与混沌现象。     

Design options for safety depressurization system

The Korean Next Generation Reactor (KNGR) adopted an advanced design feature, a safety depressurization system (SDS) to rapidly depressurize the primary system in case of events beyond the design basis. Two design approaches are considered for the SDS design. The use of bleed valves similar to the ABB-CE System 80+ is design option 1, while in design option 2, the French Sebim valve is considered to provide the combined function of overpressure protection and rapid depressurization. In this paper, thermal hydraulic analysis using a best-estimate version of CEFLASH-4AS/REM is performed for a total loss of feedwater (TLOFW) event to investigate the feasibility of those two design options. For each design option, various feed and bleed (F and B) procedures are investigated for a TLOFW event. For design option 1, the required bleed capacity is determined from the CEFLASH-4AS/REM simulation according to the EPRI Advanced Light Water Reactor (ALWR) requirements. The analysis results demonstrate that the TLOFW event can be mitigated in a proper manner with a sufficient margin using design option 1. For design option 2, the operator action times for initiating the F and B are investigated by varying the number of Sebim valves and high pressure safety injection (HPSI) pumps. If the operator opens two out of the three Sebim valves in conjunction with the four HPSI pumps before a hot leg saturation condition, the decay heat removal and core inventory make-up function can be successfully accomplished. The results of the present investigation demonstrate that the two design options are both feasible.     

Potential application of Rankine and He-Brayton cycles to sodium fast reactors

Traditionally all the demos and/or prototypes of the sodium fast reactor (SFR) technology with power output, have used a steam sub-critical Rankine cycle. Sustainability requirement of Gen. IV reactors recommends exploring alternate power cycle configurations capable of reaching high thermal efficiency.By adopting the anticipated working parameters of next SFRs, this paper investigates the potential of some Rankine and He-Brayton layouts to reach thermal efficiencies as high as feasible, so that they could become alternates for SFR reactor balance of plant. The assessment has encompassed from sub-critical to super-critical Rankine cycles and combined cycles based on He-Brayton gas cycles of different complexity coupled to Organic Rankine Cycles. The sub-critical Rankine configuration reached at thermal efficiency higher than 43%, which has been shown to be a superior performance than any of the He-Brayton configurations analyzed. By adopting a super-critical Rankine arrangement, thermal efficiency would increase less than 1.5%. In short, according to the present study a sub-critical layout seems to be the most promising configuration for all those upcoming prototypes to be operated in the short term (10-15 years). The potential of super-critical CO₂-Brayton cycles should be explored for future SFRs to be deployed in a longer run.     

Effect of Ion Engine Exhaust on the Radiation Field of a Dipole Antenna Part I: Infinite Slab Model for the Exhaust Ion Beam

A theory of the effect of an ion rocket engine exhaust on the radiation pattern of a dipole antenna is presented. The electromagnetic equations are combined with those describing the exhaust plasma beam to calculate its equivalent effective dielectric constant. In part I of this paper, the beam is represented by an infinite slab of a homogenous plasma medium, as is usually considered in space charge neutralization studies of this type of engine. The equations of propagation of electromagnetic waves through the beam medium, are used to calculate the total dipole radiation field. The method of steepest descent is applied for the evaluation of the integrals. It is found that the dipole radiation pattern depends greatly on the beam characteristic parameters.     

Azcatl-CRP: An ant colony-based system for searching full power control rod patterns in BWRs

We show a new system named AZCATL-CRP to design full power control rod patterns in BWRs. Azcatl-CRP uses an ant colony system and a reactor core simulator for this purpose. Transition and equilibrium cycles of Laguna Verde Nuclear Power Plant (LVNPP) reactor core in Mexico were used to test Azcatl-CRP. LVNPP has 109 control rods grouped in four sequences and currently uses control cell core (CCC) strategy in its fuel reload design. With CCC method only one sequence is employed for reactivity control at full power operation. Several operation scenarios are considered, including core water flow variation throughout the cycle, target different axial power distributions and Haling conditions. Azcatl-CRP designs control rod patterns (CRP) taking into account safety aspects such as k_eff core value and thermal limits. Axial power distributions are also adjusted to a predetermined power shape.     

HEPS超导高频腔低温恒温器氮低温制冷循环系统的热力学性能分析

高能同步辐射光源（HEPS）氮低温制冷循环系统为储存环隧道内的7个超导高频腔低温恒温器提供液氮冷量。氮低温制冷循环系统的热力学分析目的是针对HEPS超导高频腔低温恒温器的特点，建立系统的热力学分析模型，研究其工作过程，计算分析系统的制冷系数与循环增压比之间的对应关系。同时，考虑到系统的整体布局，进一步分析计算氮低温制冷循环系统中氮气压缩机和低温透平膨胀机的火用损，优化了系统的热力学设计参数。HEPS氮低温制冷循环系统的热力学性能分析与计算结果对于系统最优化设计和后期工程实践具有一定的指导意义。     

Crack arrest in high cycle thermal fatigue crazing

Thermal crazing in high cycle thermal fatigue due to thermal fluctuation in residual heat removal (RHR) system of some nuclear power plants is explained by crack arrest in the depth due to a decreasing stress intensity factor. This is related to high frequencies of thermal loading. An attempt has been made through a parametric study to acquire some knowledge about the loading, knowing the crack depth. For this purpose, analytical as well as finite element simulations of crack propagation in 2D- and 3D-semi-elliptical cracks have been performed. In periodic loading, bounds for the number of cycles to fatigue life are proposed. Moreover, it is shown that in the absence of mean stress, fatigue damage in RHR may be produced in the macroscopic elastic-plastic regime. Finally, it is shown by FE simulations that for a semi-elliptical crack, a small error on stress intensity factor may result in significant error on crack length at high number of cycles, due to error accumulation cycle by cycle. Moreover in this paper is given the reason as to why shielding effect has not been taken into account in the study of crack arrest in RHR.     

Multiplicative model for out-of-phase thermal fatigue degradation of ferritic–martensitic steel MANET-II

Hollow specimens of a Cr-steel have been subjected to triangle temperature cycles and different temperatures. Out-of-phase thermal fatigue lifetime is analysed and compared to isothermal fatigue data close to the mean cycle temperature. Segments associated with different temperature ranges and isothermal fatigue lines display in logarithmic scale only a slight slope difference. The thermomechanical fatigue degradation model suggests that the strain-related degradation per cycle is multiplied by a factor reflecting the particular effect of temperature oscillation.     

Optimum in-core power sharing with multicycle coupling effect

Multicycle optimization was carried out by assuming power sharing of each fuel batch as an independent parameter; that is, the power sharing of each fuel batch was considered as an optimization variable. The steepest descent method was used to optimize the power sharing for multiple cycles. Two different optimizations were carried out, i.e., multicycle and successive single-cycle optimizations. In the former, the power sharing of each fuel batch in each cycle was simultaneously optimized for multiple cycles. In the latter, optimization of the power sharing in a cycle was carried out, and then optimization in the next cycle was carried out. Maximization of discharge burnup and minimization of the number of fresh fuel assemblies were considered as the objective functions. The calculation results qualitatively and quantitatively clarify the implicit adverse effect of the single-cycle optimization, which is usually used in current core designs. Under the calculation conditions of the present study, the difference in the number of fresh fuel assemblies between multicycle and successive single-cycle optimizations is 2–3 fuel assemblies per cycle. Comparison of the power sharing obtained by both methods would provide insights to correct the adverse effect of the single-cycle optimization.     

Waste disposal aspects of the long term cooling characteristics of irradiated nuclear fuels

A study has been made of the long term cooling characteristics of nuclear fuels irradiated in commercial reactor designs of interest within the U.K. In the case of thermal reactors, Magnox, AGR, SGHWR, LWR and HTR systems fuelled with either natural or enriched uranium are considered, together with a fast reactor fuelled with plutonium derived either from the Magnox or the AGR programmes. Alternative uses for Magnox plutonium are considered by simulating a plutonium fuelled HTR thermal system and the development of a $\text{Th}{}^{233}\text{U}$ fuel cycle has been anticipated for both a fast reactor and an HTR.For each system the activities as a function of cooling time are considered on the assumption of U/Pu recovery from the fuel during reprocessing within a year of discharge from the reactor and for the alternative case of no U/Pu extraction. The reprocessing waste products associated with the various fuel cycles have then been compared both on the basis of decay heating and radiological hazard per GW(e) yr. Finally, recycling of transplutonium elements is also considered with a view to reducing the long term heating commitment from the higher actinides.