运动条件下铅铋反应堆热工水力特性研究

为研究运动条件下铅铋反应堆热工水力特性,开发了运动条件铅铋反应堆瞬态分析系统程序,并完成了对设计的5 MW自然循环小型模块化铅铋反应堆的建模,分析了运动条件对反应堆自然循环热工水力特性的影响。计算结果表明,倾斜条件下,堆芯流量减小,堆芯出口温度升高,在计算最大倾斜角度下,流量减小20%,冷却剂堆芯出口温度升高20 ℃。起伏条件下,起伏幅度和起伏周期越大,对反应堆影响越大,由于系统阻力影响,流量变化较起伏加速度有小于1 s的延时。摇摆条件下,摇摆角度越大和摇摆周期越小,对反应堆影响越大,燃料包壳峰值温度较稳态值高20 ℃以内,对反应堆正常运行时安全性影响较小。     

非能动应急堆芯余热排出系统对核电厂失去厂外电源事故的影响

研究建立了蒸汽发生器二次侧非能动应急堆芯余热排出系统热工水力特性的物理与数学模型，并编制了计算机程序。以中国秦山核电站的数据为依据，计算和分析了在失去厂外电源事故典型工况下，该系统投入运行时对瞬态热工水力特性的影响。     

倾斜与摇摆条件下一体化反应堆自然循环特性研究

通过建立海洋条件下的附加力模型与控制体空间坐标求解模型,开发了基于RELAP5/MOD3.1程序的海洋条件热工水力分析程序。研究了海洋条件下一体化反应堆IP200的自然循环特性,分析倾斜与摇摆条件对自然循环的影响。计算结果表明:倾斜会降低堆芯流量,导致左右侧环路冷却剂流量不一致,影响直流蒸汽发生器的换热特性;摇摆情形下,环路的附加压降主要由切向力贡献;摇摆轴偏离中心位置以及倾斜和摇摆的叠加运动均会打破环路间的热工水力对称性,增大堆芯流量的波动幅度。     

西安脉冲堆反应性引入事故瞬态安全特性分析

为研究西安脉冲堆（XAPR）在意外引入反应性且停堆系统失效事故下的瞬态安全特性，本文基于XAPR的结构和运行特点，建立了适用于XAPR的瞬态热工水力分析模型，并开发了用于XAPR安全特性分析的瞬态热工水力程序TSAC-XAPR。利用TSAC-XAPR程序对反应性引入事故进行模拟计算，结果表明：当XAPR在额定功率范围内运行时，发生反应性引入事故后，堆芯能依靠自身的固有反馈机制使脉冲堆重新达到稳定运行状态；当运行功率过高尤其是超过临界值时，反应性引入事故将导致脉冲堆关键热工水力参数发生振荡，无法再次达到稳态。此外，不同反应性引入方式将影响堆芯参数在反应性引入过程中的变化趋势，但并不影响其最终稳态值。     

摇摆条件下热工水力程序的研制与验证

在RELAP5/MOD3.3程序的基础上,通过添加计算摇摆因素的模块和引入新的流动传热模型以对原程序进行修正,从而建立了摇摆条件下的热工水力分析程序。利用实验结果对理论模型和程序计算结果进行了校核和验证。结果表明：本文采用的流动传热模型可准确计算出摇摆条件下的摩擦阻力系数和传热系数,建立的热工水力分析程序也可对摇摆条件下的热工水力系统进行模拟。     

CPR1000全厂断电事故瞬态特性分析

用RELAP5/MOD3.4程序对CPR1000压水堆一回路系统进行整体建模,分析全厂断电事故下一回路主要参数的瞬态热工水力特性,并将RELAP5模型计算结果与THEMIS程序的计算结果进行对比,二者符合得较好。计算结果表明：该模型可较准确地模拟CPR1000在事故下的热工水力特性。     

竖直与倾斜条件下3×3棒束通道内流场数值模拟研究

为了研究竖直及倾斜条件对堆芯热工水力特性,采用RANS模拟对棒束通道进行了数值模拟,分析了静止及倾斜条件下棒束通道内流场特征及温度分布。模拟结果表明:在入口流速相同的情况下,倾斜会使棒束通道间隙处主流速度略微减小,且倾角越大,间隙处主流流速越大。倾斜条件使得棒束通道内温度场分布发生改变,随着倾斜角度的不断增大,主流最大温度不断增加,导致棒的壁面温度增大,不利于反应堆安全。     

海洋条件下反应堆热工水力参数的子通道计算

针对海洋条件下反应堆的子通道热工水力分析,建立了海洋运动附加力模型和瞬态入口边界,将起伏、摇摆及复合运动的附加力关系式用于子通道模型的轴向和横向动量方程,并应用到COBRAⅢC程序将其改造为适应海洋条件的反应堆子通道分析程序。作为验证,计算了加热实验通道和"奥陆"堆在起伏运动情况下热通道的临界热流密度比(CHFR)、出口空泡份额和冷却剂流量,并与文献结果对比。还详细计算了"奥陆"堆在起伏、不同摇摆中心和复合运动情况下,热通道的CHFR和不同位置子通道出口的热工水力参数。研究表明:海洋条件下反应堆的子通道热工水力参数随运动呈周期性变化;起伏运动对子通道的压降影响较大,摇摆运动对子通道冷却剂的流量和温度影响较大。     

铅铋冷却沸水快堆热工水力系统安全分析程序开发

针对铅铋冷却沸水快堆（PBWFR）主回路系统建立了系统热工水力分析的数学物理模型，并开发了适用于PBWFR的热工水力系统安全分析程序SACOL。利用SACOL对PBWFR的稳态和瞬态热工水力特性进行了研究，并重点模拟了无保护超功率事故（UTOP）。计算结果表明:PBWFR在稳态时具有足够的安全性，但在UTOP中，功率短时间的迅速升高会导致包壳温度超过安全限值。      

钠冷快堆瞬态热工水力及安全分析程序开发

钠冷快堆是第4代核反应堆的主力堆型，瞬态热工水力及安全特性是其设计研发和安全评审的重要工作，需要专用的分析工具。本文基于模块化建模思想，建立了钠冷快堆系统关键部件的热工水力模型和辅助模型，采用具有高稳定性和自动变步长能力的Gear算法，开发了钠冷快堆瞬态热工水力及安全分析软件THACS，并通过了国际基准题EBR-Ⅱ的有保护失流事故实验SHRT-17的初步验证。结果表明，THACS程序能较好模拟此实验的瞬态过程，具备钠冷快堆瞬态热工水力及安全分析的能力，可为我国钠冷快堆研发提供支持。     

Effects of ship motions on natural circulation of deep sea research reactor DRX

DRX is a very small integral type PWR (750 kW_t) for a scientific deep-sea research bathyscaph. The core having a small amount of steam is cooled by natural circulation and pressurized by self-pressurization. During operation of the bathyscaph in a deep sea or near the water surface, a ship inclination or ship motions will affect the reactor behavior. This paper describes the effect of a heeling or a heaving on the thermal hydraulic behavior of reactor system, which is analyzed by improved so to simulate the effect of ship motions. The dynamics has a feature of nuclear power-natural circulation flow coupling under the condition of external forces. The analysis shows that ship inclination induces the core flow to decrease but reactor power recovers to the initial level without help of the reactor automatic control system. The heaving makes the core flow and the reactor power oscillate in phase with heaving, which are different from a density wave oscillation. Oscillation amplitudes of the flow and the power have peaks at the heaving period of 5 s. The peaks are due to resonance of the natural circulation flow and the heaving. An effective measure to suppress this oscillations due to heaving is to pressurize the primary loop by filling non-condensable gas. The density wave oscillation occurs when the reactor power increases over the rated power, and the boundary of its occurrence is analytically revealed. Under the condition of both density wave oscillation and heaving, the system shows to oscillate with the overlapped effect.     

自然循环铅基快堆一回路系统驱动力的关键影响因素研究

为深入研究影响自然循环铅基快堆一回路系统驱动力的关键因素,以自然循环铅基快堆SNCLFR-10为研究对象构建描述反应堆一回路自然循环稳态运行模型;从理论上量化分析冷/热池的热量传递、热源和热阱温度非线性分布、反应堆压力容器壁散热3种因素对自然循环能力的影响,并开展了相关数值模拟验证。结果表明,数值模拟结果与本研究理论计算值吻合较好;3种自然循环能力影响机制耦合作用将降低SNCLFR-10系统自然循环能力,导致自然循环流量与功率之间不再满足理论所得的1/3次方关系。      

先进压水堆一回路地震反应谱分析及参数敏感性研究

压水堆一回路系统包含压力容器、蒸汽发生器、主泵、稳压器、主管道和波动管等重要部件,各部件在地震激励下的动态响应与整个系统的结构形式密切相关。本文从系统的角度,以非能动先进压水堆一回路为研究对象,运用ANSYS建立了其三维有限元模型,在模态分析的基础上,进行了三正交方向输入下的反应谱分析,得到了系统在地震载荷下的响应。并对反应谱输入角度和支撑刚度进行了敏感性研究,给出了这些特性参数对结构设计和分析的指导性意见。此外,通过直接积分法得到系统的地震时程响应,并与谱分析的模拟结果进行了对比,同时也为主泵等单个部件的详细地震分析提供位移、加速度输入。最后通过三维实体模型与集中质量模型抗震计算结果的比较,说明建立三维实体模型的必要性。本文为核电站一回路重要设备的结构分析提供了技术支持。     

Thermal power calibrations of the IPR-R1 TRIGA reactor by the calorimetric and the heat balance methods

Since the first nuclear reactor was built, a number of methodological variations have been evolved for the calibration of the reactor thermal power. Power monitoring of reactors is done by means of neutronic instruments, but its calibration is always done by thermal procedures. The purpose of this paper is to present the results of the thermal power calibration carried out on March 5th, 2009 in the IPR-R TRIGA reactor. It was used two procedures: the calorimetric and heat balance methods. The calorimetric procedure was done with the reactor operating at a constant power, with primary cooling system switched off. The rate of temperature rise of the water was recorded. The reactor power is calculated as a function of the temperature-rise rate and the system heat capacity constant. The heat balance procedure consists in the steady-state energy balance of the primary cooling loop of the reactor. For this balance, the inlet and outlet temperatures and the water flow in the primary cooling loop were measured. The heat transferred through the primary loop was added to the heat leakage from the reactor pool. The calorimetric method calibration presented a large uncertainty. The main source of error was the determination of the heat content of the system, due to a large uncertainty in the volume of the water in the system and a lack of homogenization of the water temperature. The heat balance calibration in the primary loop is the standard procedure for calibrating the power of the IPR-R1 TRIGA nuclear reactor.     

熔盐堆主回路系统氙的动力学特性研究

基于Mathematica7.0为熔盐堆（MSR）主回路系统建立了一套含流动项及在线去除功能的氙（135Xe）的动态分布数值分析程序，针对2?MW MSR的一种设计方案，分析了不同流量、不同启停堆功率、不同在线去除效率情况下135Xe浓度随时间的动态变化特性。结果表明:相较于静态燃耗模型，流动燃耗模型的135Xe带来的负反应性要低约32.2%；额定流量下主回路系统135Xe浓度分布均匀，只有当主回路系统体积流量小于2.24 cm3·s-1时，流动效应才会对主回路系统内135Xe浓度分布有显著影响；当鼓泡系统的在线去除份额约为0.1%时可以使堆芯135Xe带来的负反应性降低至-38.3 pcm?（1 pcm =10-5），其总的去除效率可以达到86.0%；不同功率水平瞬时停堆工况下，堆芯135Xe浓度单调下降，停堆约50 h后135Xe基本消失，相当于引入+254 pcm反应性，停堆过程无碘坑出现，停堆后再启堆过程不必担心碘坑启动的问题。135Xe去除效率对整个系统135Xe总量有一定影响，在去除份额从0.0001%~20%的变化范围内，135Xe的总活度与静态燃耗模型相比相应增加了0.67%~8.75%。      

Thermal power calibrations of the IPR-R1 TRIGA reactor by the calorimetric and the heat balance methods

Since the first nuclear reactor was built, a number of methodological variations have been evolved for the calibration of the reactor thermal power. Power monitoring of reactors is done by means of neutronic instruments, but its calibration is always done by thermal procedures. The purpose of this paper is to present the results of the thermal power calibration carried out on March 5th, 2009 in the IPR-R TRIGA reactor. It was used two procedures: the calorimetric and heat balance methods. The calorimetric procedure was done with the reactor operating at a constant power, with primary cooling system switched off. The rate of temperature rise of the water was recorded. The reactor power is calculated as a function of the temperature-rise rate and the system heat capacity constant. The heat balance procedure consists in the steady-state energy balance of the primary cooling loop of the reactor. For this balance, the inlet and outlet temperatures and the water flow in the primary cooling loop were measured. The heat transferred through the primary loop was added to the heat leakage from the reactor pool. The calorimetric method calibration presented a large uncertainty. The main source of error was the determination of the heat content of the system, due to a large uncertainty in the volume of the water in the system and a lack of homogenization of the water temperature. The heat balance calibration in the primary loop is the standard procedure for calibrating the power of the IPR-R1 TRIGA nuclear reactor.     

An experimental study on natural circulation decay heat removal system for a loop type fast reactor

A fully natural circulation-based system is adopted in the decay heat removal system (DHRS) of an advanced loop type fast reactor. Decay heat removal by natural circulation is a significant passive safety measure against station blackout. As a representative of the advanced loop type fast reactor, DHRS of the sodium fast reactor of 1500 MWe being designed in Japan comprises a direct reactor auxiliary cooling system (DRACS), which has a dipped heat exchanger in the reactor vessel, and two units of primary reactor auxiliary cooling system (PRACS), which has a heat exchanger in the primary-side inlet plenum of an intermediate heat exchanger in each loop. The thermal-hydraulic phenomena in the plant under natural circulation conditions need to be understood for establishing a reliable natural circulation driven DHRS. In this study, sodium experiments were conducted using a plant dynamic test loop to understand the thermal-hydraulic phenomena considering natural circulation in the plant under a broad range of plant operation conditions. The sodium experiments simulating the scram transient confirmed that PRACS started up smoothly under natural circulation, and the simulated core was stably cooled after the scram. Moreover, they were conducted by varying the pressure loss coefficients of the loop as the experimental parameters. These experiments confirmed robustness of the PRACS, which the increasing of pressure loss coefficient did not affect the heat removal capacity very much due to the feedback effect of natural circulation.     

The primary loop confinement and pressure boundary system of the HTR-10

The design philosophy and requirements of the HTR-10 reactor building and the primary loop confinement are introduced in this paper. Also introduced are the design, fabrication and the installation of the HTR-10 primary loop pressure boundary system. The primary loop confinement comprises the sealed cavities of the reinforced concrete structure. The main components and the connected gas systems of the primary loop pressure boundary system are contained in the confinement. Under normal operating condition, the inside pressure of the confinement is kept at negative pressure to ensure the sealing function of the confinement. There is a rupture disk of overpressure protection in the confinement wall. After a depressurization accident the pressure of the confinement increases and the rupture disk will break. The air of the confinement is discharged directly to the atmosphere through the accident discharge chimney which is connected to the rupture disk without filter. The main components of the primary loop pressure boundary system consist of the reactor pressure vessel, the steam generator pressure vessel and the hot gas duct vessel. All the above main components are installed in the reactor cavity and the steam generator cavity. They are all nuclear safety class 1 components, whose materials production, design, fabrication, and tests are carried out according to ASME Section III and relevant Chinese nuclear codes.     

基于盲源分离算法的压水堆松动件信号提取方法

压水堆一回路松动件状态监测系统的关键环节在于如何从复杂背景噪声中提取松动件冲击信号。利用基于最大信噪比的盲源分离算法，将压水堆背景噪声和松动件冲击信号从观测到的混合信号中分离，并利用相似系数对该算法的分离效果进行了评估。根据该算法，当分离信号之间关系均独立时，信噪比函数取得最大值。研究结果表明，该算法可高效、准确地实现压水堆一回路松动件冲击信号和背景噪声信号的分离。     

Transient flow analysis in reactor coolant pump systems during flow coastdown period

The rapid flow transient calculation in reactor coolant pump system is important in the safety analysis of a nuclear reactor. An accurate transient analysis of flow coastdown is also important and necessary for the design and manufacture of a reactor coolant pump. Only under the reliable work of a reactor coolant pump the safety of a nuclear power plant can be guaranteed. A mathematical model is developed for solving flow rate transient and pump speed transient during flow coastdown period. The detailed information of the centrifugal pump characteristics is not required. The flow rate and pump speed are solved analytically. The analytic solution of non-dimensional flow rate indicates that non-dimensional flow rate is determined by energy ratio β. The kinetic energy of the loop coolant fluid and the kinetic energy stored in the rotating parts are two important parameters in form of β. When the steady-state flow rate and pump speed are constant, the inertia of primary loop fluid and the pump moment of inertia are also two important parameters in flow transient analysis. For the condition all pump shafts are seized, the flow decay depends on the inertia of primary loop fluid. For the case that pump inertia is very large, the flow decay is determined by the pump inertia. The calculated non-dimensional flow rate and non-dimensional pump speed using the model are compared with published experimental data of two nuclear power plants and a reactor model test on flow coastdown transients. The comparison results show a good agreement. As the flow rate approaches to zero, the increase difference between experimental and calculated value is due to the effect of the mechanical friction loss.