Corrosion of candidate materials for supercritical water-cooled reactor

Supercritical water reactor(SCWR) was proposed as a GenerationⅣconcept for building large capacity nuclear power plants.Comparing with the present GenerationⅡandⅢlight water reactors,SCWR possesses great advantages of 10%higher efficiency,simpler system design,better sustainability,and so on. However,the selection of materials for fuel cladding and reactor internals of SCWR is facing a great challenge. Corrosion in supercritical steam is of the first important issue to be solved to meet the stringent requirement of the reactor internal components.Corrosion screening tests were conducted on candidate materials for nuclear fuel cladding and reactor internals of supercritical water reactor(SCWR) in static and re-circulating autoclave at the temperatures of 550,600 and 650℃,pressure of about 25 MPa,deaerated or saturated dissolved hydrogen(STP). Nickel base alloy type Hastelloy C276,austenitic stainless steels type 304NG,AL-6XN,HR3C.NF709 and SAVE 25,ferritic/martensitic(F/M) steel type P92,P122 and 410,and oxide dispersion strengthened steel MA 956,are tested.This paper presents corrosion rate,and focuses on the formation and breakdown of corrosion oxide film,and proposes the future trend for the development of SCWR internal structure materials.

Control system design for a pressure-tube-type supercritical water-cooled nuclear reactor via a higher order sliding mode method

Nuclear power plants exhibit non-linear and time-variable dynamics.Therefore,designing a control system that sets the reactor power and forces it to follow the desired load is complicated.A supercritical water reactor(SCWR)is a fourth-generation conceptual reactor.In an SCWR,the non-linear dynamics of the reactor require a controller capable of control-ling the nonlinearities.In this study,a pressure-tube-type SCWR was controlled during reactor power maneuvering with a higher order sliding mode,and the reactor outgoing steam temperature and pressure were controlled simultaneously.In an SCWR,the temperature,pressure,and power must be maintained at a setpoint(desired value)during power maneuvering.Reactor point kinetics equations with three groups of delayed neutrons were used in the simulation.Higher-order and classic sliding mode controllers were separately manufactured to control the plant and were compared with the PI controllers speci-fied in previous studies.The controlled parameters were reactor power,steam temperature,and pressure.Notably,for these parameters,the PI controller had certain instabilities in the presence of disturbances.The classic sliding mode controller had a higher accuracy and stability;however its main drawback was the chattering phenomenon.HOSMC was highly accurate and stable and had a small computational cost.In reality,it followed the desired values without oscillations and chattering.

Continuous Production of Biodiesel from Soybean Oil Using Supercritical Methanol in a Vertical Tubular Reactor：I.Phase Holdup and Distribution of Intermediate Product along the Axial Direction

Production of biodiesel with supercritical methanol is a green synthesis process.A study was carried out in a vertical tubular reactor with a length of 3700 mm and a diameter of 20 mm at 275-375°C,15 MPa,and molar ratio of methanol to soybean oil of 40︰1.The phase holdup,intermediate product,yield and axial distribution of methyl ester(ME) were investigated.Methanol and oil were mixed non-uniformly due to the formation of biodiesel and difference in their densities,even when the reaction system was in the supercritical state.From top to bottom,the phase holdup of methanol increased and that of oil decreased.As temperature increased,the concentrations of monoglyceride and diglyceride decreased gradually and the ME yield increased.When the temperature reached 300°C,the critical temperature of the system,the ME yield was 50%.Further increase in temperature led to a sharp in-crease of ME yield.However,at 375°C after 1200 s of reaction time,the decomposition rate of ME was greater than its formation rate,reducing the ME yield.

Treatment of Acetonitrile by Catalytic Supercritical Water Oxidation in Compact-Sized Reactor

The objective of this research was to study the treatment of acetonitrile by catalytic supercritical water oxi-dation in a compact-sized tubular reactor, with an internal volume of 4.71 mL. Manganese dioxide was used as the catalyst and H2O2 was used as the oxidant. The oxidation of acetonitrile in supercritical water was studied at 400-500 oC, 25-35 MPa, the flow rate of 2-4 mL/min, the initial concentration of acetonitrile 0.077-0.121 M and the %excess O2 of 50-200%. As a result, the products were mainly N2, CO2 and CO and acetonitrile can be decomposed > 93 % within a very short contact time (1.45-6.19 s). The oxidation process was carried out with respect to the conversion of acetonitrile by 25 factorial design. Regression models were obtained for correlating the conversion of acetonitrile with temperature and flow rate. The complete oxida-tion can be achieved at a condition as moderate as 400 oC, 25 MPa with the flow rate of 2 mL/min.

Primary Breeding Ratio Analysis of an Improved Supercritical Water Cooled Fast Reactor

The purpose of the study is to analyze the breeding ratio of a supercritical water cooled fast reactor (SCFR) and to increase the breeding core of SCFR. The sensitivities of assembly parameters, core arrangements and fuel nuclide components to the breeding ratio are analyzed. In assembly parameters, the seed fuel rod diameter has higher sensitivities to the conversion ratio (CR) than the coolant tube diameter in blanket. Increasing heavy metal fraction is good to CR improvement. The CR of SCFR also increases with a reasonable core arrangement and Pu isotope mass fraction reduction in fuel, which can achieve more negative coolant void reactivity coefficient at the same time. The breeding ratio of SCFR is 1.03128 with a new core arrangement. And the coolant void reactivity coefficient is negative, which achieves a fuel breeding in initial fuel cycle.

Simulation of heat transfer of supercritical water in obstacle-bearing vertical tube

The heat transfer coefficient is very low at bulk temperatures higher than the pseudo-critical point,because the supercritical pressure leads to a vapor-like fluid.In this paper,the heat transfer downstream an obstacle-bearing vertical tube is simulated by the CFD code of Fluent 6.1,using an adaptive grid in the supercritical condition.The reliable results are obtained by the RNG k-ε model using the enhanced wall treatment.The blockage ratio and local temperature of obstacle affect greatly the heat transfer enhancement,and the resultant influence region and decay trend are compared with the existing equations.

Analysis of prompt supercritical process with heat transfer and temperature feedback

The prompt supercritical process of a nuclear reactor with temperature feedback and initial power as well as heat transfer with a big step reactivity (ρ0>β) is analyzed in this paper. Considering the effect of heat transfer on temperature of the reactor, a new model is set up. For any initial power, the variations of output power and reactivity with time are obtained by numerical method. The effects of the big inserted step reactivity and initial power on the prompt supercritical process are analyzed and discussed. It was found that the effect of heat transfer on the output power and reactivity can be neglected under any initial power, and the output power obtained by the adiabatic model is basically in accordance with that by the model of this paper, and the analytical solution can be adopted. The results provide a theoretical base for safety analysis and operation management of a power reactor.

Flow Instability in Parallel Channels with Water at Supercritical Pressure: A Review

Research into flow instability at both subcritical and supercritical pressures has attracted attention in recent years because of its potential of occurrence in industrial heat transfer systems. Flow instability has the potential to affect the safety of design and operation of heat transfer equipment. Flow instability is therefore undesirable and should be avoided?in the design and operation of industrial equipment. Rahman?et al. reviewed studies on supercritical water heat transfer with the aim of providing references for SCWR researchers. It was found out that most of the CFD studies and experimental studies were performed with single tube geometry due to the complexity of parallel channel geometry. Because studies performed with parallel channel geometry could provide detailed information to the design of the SCWR core, they called for more studies in parallel channel geometry at supercritical pressures in the future. In order to help understand how flow instability investigations are carried out and also highlight the need to understand flow instability phenomenon and equip the designers and operators of industrial heat transfer equipment with the needed knowledge on flow instability, this study carried out a review of flow instability in parallel channels with water at supercritical pressures.

Performance of Heat Transfer Correlations Adopted at Supercritical Pressures: A Review

Research activities involving heat transfer at supercritical pressures have attracted attention in recent years because of possibility of increase in thermal output of heat transfer and industrial equipment. Because of high pressure and temperature conditions associated with heat transfer at supercritical pressures, only few experimental heat transfer studies are being carried out at supercritical conditions. The use of numerical tools for heat transfer and other related studies at supercritical pressures is increasing because of the high-pressure-temperature limitation of experimental studies at supercritical conditions. Heat transfer correlations implemented in these numerical tools are used to obtain numerical heat transfer data to complement experimental heat transfer data provided through experimental studies. In order to further broaden the understanding of fluid flow and heat transfer, this review examines the performance of heat transfer correlations adopted at supercritical pressures. It is found from the review that most of the correlations could predict heat transfer quite well in the low enthalpy region and few of the correlations could predict heat transfer in the high enthalpy region near critical and pseudo-critical conditions (heat transfer deteriorated conditions). However, no single heat transfer correlation is able to accurately predict all the experimental results presented in this work.

喷嘴参数对超临界水热燃烧特性影响的模拟

反应器喷嘴的作用是维持水热火焰在复杂流场中保持稳定。建立内预热式蒸腾壁反应器(IPTWR)内甲醇超临界水热燃烧过程计算流体力学模型,分析了喷嘴的材料热物性和结构参数对进料混合特性及火焰结构的影响规律。结果表明,喷嘴材料传热特性的提高使水热火焰朝高温和广域的方向发展;氧气-辅热混合段直径由18 mm减小到14 mm,反应器轴向温度峰值由954.84 K升高到981.60 K,火焰位置向远端移动;随着喷嘴缩进深度减小,水热火焰逐渐向喷嘴出口聚拢,表现为火焰收缩现象。在此喷嘴结构参数范围内小直径的短氧气-辅热混合流道有助于水热火焰的温度升高和聚拢稳定。结果可为内预热式蒸腾壁反应器的喷嘴设计提供理论指导。

煤炭超临界水汽化技术研究进展

回顾了近年来煤炭超临界水汽化制氢的进展,介绍了制氢反应机理,系统总结了温度、浓度、停留时间和催化剂等因素对煤炭超临界水汽化制氢的影响及煤炭超临界水汽化反应装置的发展现状。针对现存问题对煤炭超临界水汽化制氢的未来前景进行了展望。

圆管内氦氙混合气体与超临界二氧化碳换热特性对比分析

小型化、高紧凑反应堆系统是陆上多用途能源供给的研究重点,直接热-动循环下的能量转换对冷却剂工质选型提出极高要求。为了明确氦氙混合气体、超临界二氧化碳的工质适用性,本文采用数值模拟方法,对二者在圆管内的换热特性进行计算,对比了2种冷却剂的物性,分析了不同加热功率、入口速度、入口温度对冷却剂换热系数的影响,拟合提出了2种冷却剂工质圆管内换热的经验模型。研究结果表明:反应堆在加热功率超过100 kW/m~2、入口流速大于10 m/s下,对流换热系数较大;氦氙混合气体入口温度在1 000~1 200 K、超临界二氧化碳在入口温度550~600 K附近时,对流换热系数存在极大值;在高雷诺数区(Re>10~4),2冷却剂的修正关系式与计算值吻合良好。本文计算能够为新型反应堆开发提供数据及模型基础。

超临界水氧化技术中反应器腐蚀问题研究现状

本研究介绍了超临界水氧化过程中反应器的腐蚀问题以及国内外关于此类问题的研究。目前研究主要使用特殊结构的反应器以降低腐蚀速率,或者利用晶界工程技术提升材料的抗氧化性解决腐蚀问题,但没有对反应器的腐蚀机制进行研究。因此,本文提出通过试验研究和数值模拟分析探究反应器的腐蚀机制。这对超临界水氧化技术中的发展和应用具有重要的意义。

超临界水氧化反应器数值分析

蒸发壁式反应器作为超临界水氧化系统的核心设备,反应区温度及反应器出口温度是指导反应器结构设计的关键因素。通过建立反应器三维几何模型,采用CFD数值模拟手段,计算蒸发壁式反应器内部流场分布,了解分布规律,为反应器结构设计提供依据,保证设备安全稳定运行。

Development of a displacement-reactivity feedback model for dynamic behavior simulation in fast burst reactor

In this study, a displacement-reactivity feedback model, which can directly represent the inherent ‘‘thermal expansion extinction effect'' of fast burst reactors(FBRs),was developed with the aid of the static neutron transport component of the FBR-MPC code. Dynamic behaviors of bursts in the Godiva I reactor were simulated by coupling the simplified multiphysics models consisting of the point kinetic equations for neutronics, adiabatic equation for temperature, and thermoelastic equations for displacement/stress with the developed model. The results were compared with the corresponding experimental data and those obtained using the traditional fission yield(temperature rise)-reactivity feedback models. It was found that the developed model can provide good results for the bursts with no or a small inertia effect. For the bursts with a prominent inertia effect, the smaller burst width and asymmetric distribution of the fission rate curve, noticed in the experiments but not evident using the traditional models, can be reproduced. In addition, the realistic oscillations in reactivity and fission rate caused by the core vibration, as well as the deeper sub-prompt criticality in the plateau following the burst, can be observed. Therefore, the developed displacement-reactivity feedback model can be expected to be an effective tool for calculating the dynamic behaviors of bursts.

超临界二氧化碳核能动力系统的兴起和发展

超临界二氧化碳(S-CO_(2))核能动力系统以S-CO_(2)为工质,通过直接或间接循环将核释热转换为电能或机械能。本文总结了国际上S-CO_(2)核能动力系统的概念初创、研究重启、协同创新3个历史阶段近60年的发展历程,分析了S-CO_(2)核能动力系统在核反应堆设计、材料工艺、热工流体力学、换热器、涡轮发电系统、系统运行、控制及安全策略等方面的研究现状,并提出了S-CO_(2)核能动力系统在基础研究和工程攻关中面临的技术挑战和攻关方向。目前,以中美为代表的能源强国已初步完成S-CO_(2)动力转换系统的实验室级测试,预计可在5~10年内实现中等规模工程示范甚至规模化应用。S-CO_(2)核能动力系统基本具备了走向工程的前提条件,有望引领先进核能技术变革。

超临界二氧化碳冷却反应堆空泡反应性研究

由于二氧化碳本身的慢化能力较弱,因此超临界二氧化碳冷却反应堆通常具有较硬的能谱,在冷却剂丧失事故发生时容易出现由正反应性反馈引入的安全问题。本文针对超临界二氧化碳反应堆的能谱特点,提出了描述其中子循环过程的三因子公式,并将冷却剂丧失对反应性的影响拆分为能谱项和泄漏项,作为进一步研究的理论依据。基于蒙特卡罗模拟方法,对美国麻省理工大学提出的超临界二氧化碳冷却反应堆堆设计方案进行了建模计算和验证,分析了径向反射层、添加慢化材料的影响。研究结果表明:超临界二氧化碳反应堆的设计需要注重能谱的软化与合理的堆芯几何设计,通过分区设置慢化材料的方案可以展平通量、软化能谱,同时降低冷却剂丧失事故引入的反应性;以超临界二氧化碳作为反射层材料,可以通过增大冷却剂丧失事故时的泄漏率在保证中子经济性的同时实现较低的空泡反应性;在进行超临界二氧化碳反应堆设计时,需综合考虑空泡反应性随燃耗的变化,并可以通过优化燃料核素组成来降低空泡反应性。

超临界二氧化碳核能系统负荷运行策略研究

为了确保超临界二氧化碳再压缩布雷顿循环直接冷却核能系统在变负荷工况下的安全经济运行,利用稳态和瞬态回热器换热实验验证了自主研发的瞬态分析程序SCTRAN/CO_(2)在预测布雷顿循环动态特性方面的可行性,并以该程序作为分析工具,开展了超临界二氧化碳再压缩布雷顿循环核能系统分别在反应堆反应性扰动和二次侧冷却水流量扰动两种典型瞬态工况下的开环动态特性研究,开发了主压缩机入口温度恒定、堆芯出口温度恒定及改变循环工质装量的负荷运行策略,研究了系统在100%~50%~100%变负荷工况下的瞬态响应。结果表明,装量控制策略能实现以额定满功率5%/min的速率追踪负荷,并保证系统安全经济运行。此过程中,堆芯压力波动0.6 MPa,堆芯出口温度波动不超过5℃。研究获得的瞬态分析工具、开环特性和控制策略为此核能系统的负荷运行策略研究提供了参考。

用于水下航行器的铅铋堆S-CO_(2)循环发电系统热力学及动态特性分析

将铅铋堆与超临界二氧化碳(supercriticalcarbon dioxidecycle,S-CO_(2))循环相结合的发电系统可为水下航行器稳定、高效供能。然而,目前缺乏水下航行器与铅铋堆S-CO_(2)循环发电系统的匹配设计研究,其热力学分析仍不完善。因此,该文分别构建其热力学模型和动态模型,通过模拟仿真,开展发电系统与4种动力循环形式的匹配研究,发现简单回热循环与铅铋合金的温度匹配性较好,系统热效率在压缩机和透平入口压力分别为7.6和25MPa时达到26.81%,满足系统设计要求。进一步探讨循环关键运行参数对系统热力学性能的影响规律,结果表明,增大循环最高压力和最高温度有利于系统效率的提升。基于此,进行系统关键换热部件的设计,揭示发电系统在温度阶跃扰动条件下的动态响应特性。可知,当堆芯铅铋温度从410℃降至390℃时,循环效率在大约8s内下降1.69%。该文可为应用于水下航行器的铅铋堆超临界二氧化碳循环发电系统设计提供参考。