核供热堆不能完全停堆的失水事故中主回路自然循环断流实验

５ＭＷ核供热堆和２００ＭＷ核供热堆的主回路是一体化的自然循环系统。在破口失水事故中，当液位降至低于主换热器入口上沿以后会发生主回路冷却剂自然循环的断流过程，影响堆芯的冷却和系统的稳定性。当发生失水事故而且反应堆又不能安全停堆时这种影响更大。在５ＭＷ核供热堆热工水力模拟回路ＨＲＴＬ－５上进行了实验。     

200MW低温核供热堆应急电力系统的可靠性分析

简要介绍了２００ＭＷ低温核供热堆应急电力系统的设计特点，并用故障树分析方法，对其应急电力系统的可靠性进行了分析，从理论上论证出现有简化的２００ＭＷ低温核供热堆应急电力系统设计方案的安全母线供电可靠性指标在保留小数点后４位要求时，与采用一般核电厂应急电力系统设计方案时相一致。     

200MW核供热堆核能海水淡化及接口方案的研究

简要介绍了核能海水淡化的必要性以及２００ＭＷ核供热堆的技术安全特点，并从该堆的特点出发，探讨了它与海水淡化厂的１４种接口方案，通过经济、技术及安全性方面的分析比较，从中选取了适合２００ＭＷ核供热堆的较为理想的接口方案为：蒸汽发生器＋ＭＥＤ海水淡化厂（单一产水方案）和蒸汽发生器＋汽轮发电机组＋ＭＥＤ海水淡化厂（水电联产方案）。     

200MW核供热堆水力稳定性实验数值模拟分析

为确定２００ＭＷ核供热堆的安全性并为今后发展奠定研究基础，在改装的俄罗斯ＫＣ实验装置上，完成了该核供热堆主回路系统水力稳定性实验研究，然后应用ＲＥＴＲＡＮ－０２程序完成主要实验工况数值模拟计算，理论计算和实验测量的对比结果表明，两者间的较好的符合，获得有益于核供热堆安全设计的一些结论。     

NHR-200燃料组件进口阻力特性实验研究

在２００ＭＷ核供热堆水力学实验回路上完成了ＮＨＲ－２００燃料组件进口阻力特性实验研究。采用１：１的实验本体,模拟条件下为几何形状,雷诺数相同。     

基于微机的核供热堆模拟器PCNHR1.0的软件系统研究

为了使更多的技术人员形象地了解低温堆供热站的原理，在国际原子能机构的支持下，清华大学核能技术设计研究院研究开发了基于微机的２００ＭＷ低温核供热堆模拟器。它采用两回路、一维漂移流热工水力学模型，点中子堆物理以及控制系统模型，能对核供热堆稳态运行、瞬态过程和事故进行仿真，仿真精度接近系统分析结果。在奔腾或以上的微机上，ＷＩＮＤＯＷＳ９５／９８／ＮＴ操作系统下，能对过程进行实时仿真，而且大多数过程能达到     

核动力系统模拟技术的研究

简要回顾了清华大学核研院在系统模拟技术方面所开展的主要工作，重点介绍了基于ＲＥＴＲＡＮ－０２程序研究开发的２００ＭＷ核供热堆紧凑型模拟器和基于网络计算技术的开发的１０ＭＷ高温气冷堆网络并行模拟原型系统。     

控制棒水力驱动系统的设计和研究

分析了２００ＭＷ核供热堆控制棒水力驱动系统的设计特点;系统中主要设备的设计特点及特性;旁路自调节结构的设计及其高温下的补偿作用以及系统温度特性的实验结果。经对实验结果的分析表明：ＨＤＳＣＲ和各设备的设计合理,运行可靠;各设备的设计不仅降低了设备的加工难度及安装难度,而且改善了系统的温度特性;系统安全能满足２００ＭＷ核供热堆对控制棒驱动机构的要求。     

200MW核供热堆工业供汽的经济分析

对２００ＭＷ低温核供热堆用于工业开发集中供应压力为１．５ＭＰａ左右的饱和蒸汽方案的经济性进行了较全面的分析和比较。结果表明，２００ＭＷ低温核供热堆用于工业供汽的各项经济评价指标都很好，内部收益率为２０．５５％，净现值（贴现率１０％）为７．４５亿元；投资回收期和贷款偿还期均在项目建设成后５年以内。与低温核供热用于冬季供暖相比，其经济效益有非常明显的改善。同时，在经济发达地区的工业开发区，利用核能进行     

核供热堆不凝结气体排放份额试验研究

叙述了核供热堆上空腔不凝结气体在发生排放事故工况下排放特性的试验研究。试验模拟了核供热堆排放工况的主要参数，着重研究了在排放过程中的氮气排放份额，及氮气对排放背压的影响。本试验是在５ＭＷ核供热堆热工水力学试验回路（ＨＲＴＬ－５）和排放回路上进行的，系统压力为１．５ＭＰａ，初始氮气分压为０．３４ＭＰａ。采用静态校验法，成功地获得了氮气的排放份额。这些试验数据为核供热堆的安全分析提供了重要的试验数据。     

On the use of the multivariate autoregressive and relative power contribution models for determination of transient root cause

The multivariate autoregressive (MAR) model and the relative power contribution (RPC) ratio are used in this work to determine the root causes of a power oscillation event and an apparent positive reactivity insertion transient occurred at the at the BWR/5 Units of the Laguna Verde Nuclear Power Plant (LVNPP) of México. The application of the MAR and PRC models leads to identify dominant frequencies and the contribution from other different signals to the dominant frequencies. The methodology firstly uses a linear model to estimate the response characteristics of the system and the spectra of the noise sources. The estimate of the process linear predictor is obtained by the ordinary least squares method. Then, the model performs a MAR analysis, and the RPC ratio is computed to determine the inter-relationship between the different reactor noise signals. The RPC ratio is an indication of how the fluctuation of one variable depends on other variables, at each frequency.Reactor signals acquired during the two transient operational events are used in the analysis. In the first event, a problem on the position controller of the flow control valve's stem induced the appearance of a power peak of 12% amplitude on the average power range monitors. Actual insertion of positive reactivity did not occur. The signals used for the analysis come from an average power range monitor, the position of the stem in the valve, controller of stem position, and controller of the recirculation flow. For the second transient, power oscillations of about 12% amplitude occurred. Signals from an average power range monitor, total flow through the core and flow through the 10 jet pump of each loop are analyzed. In both cases, some other signals were also used, but since they did not show appreciable influence on the RPC results, they were not considered for final analysis.The RPC results obtained in here confirmed previously known facts about the origin of the transients analyzed. Specifically, for the first transient, a dominant frequency of 1.7 Hz appeared on the power spectral densities of different signals from sensors on the recirculation loop B. At this 1.7 Hz frequency, the RPC ratio showed influence of such loop B spectra to the average reactor power spectrum, but no influence at all of the average reactor power to any of such loop B signals. The root cause, although not a real power transient event, therefore was not of neutronic nature, but related to recirculation flow. In the second transient, the prominent power oscillation frequency (0.54 Hz) was tracked within the spectral data of other signals. The RPC results for this case showed a strong influence of the average reactor power on the flow signals, but only a modest contribution from the flow signals to the average reactor power. The root cause therefore was of neutronic nature, due mainly to a combination of a particular core configuration and control rod pattern change at the moment of the event.     

蒸汽发生器液位控制系统手自动切换的前馈补偿研究

当前蒸汽发生器（SG）液位控制系统手自动切换信号复制回路的设计中,液位控制器运算基准为切换时的汽水失配信号,主给水流量调节阀由手动模式切到自动模式后导致SG液位控制系统失去快速调节给水流量的前馈作用。针对该问题,结合阳江核电厂4号机组SG液位高高跳堆事件,提出了针对手自动切换操作方式和系统设计的2种优化方案。针对操作方式的优化,在主给水流量调节阀投自动前,手动平衡汽水流量;针对系统设计的优化,增加汽水失配判断环节和前馈自动补偿环节。通过SG液位扰动试验证明,所提出的优化方案能有效提高手自动切换后控制系统的调节速度、减小超调量,对核电机组安全运行水平提升有重要贡献。      

CPR1000核电机组延伸运行优化控制

延伸运行（SO）是压水堆核电机组灵活运行的重要手段,研究如何提升机组SO模式下的安全性和经济性具有重要意义。针对某中国改进型三环路压水堆（CPR1000）核电机组某次SO模式下一回路平均温度、堆芯热功率、堆芯轴向功率偏差和温度调节棒棒位等重要参数存在波动的案例,研究表明波动的主要原因是由于该CPR1000核电机组的汽轮机高压调节阀运行在流量特性曲线的陡峭区,导致阀门开度在外部扰动影响下产生波动,并诱发主蒸汽流量、一回路平均温度等重要参数的波动。结合该核电机组设备的运行特性,提出优化高压调节阀流量特性曲线和优化主蒸汽流量限值等策略来提高机组SO期间安全性和经济性。数台CPR1000核电机组采用SO模式的工程实践案例验证了该策略的有效性。      

液态熔盐堆运行安全特性初步研究

液态燃料反应堆与固态燃料反应堆相比,原理上有较大不同。液态熔盐堆中由于燃料流动带走缓发中子先驱核在堆外衰变导致堆芯反应性降低,且裂变产物在堆外回路中衰变也会引起一回路发热。本文使用熔盐堆中子动力学程序Cinsf1D探讨2 MW熔盐堆的临界动力学特性和安全特性,研究零功率临界下不同熔盐流速启泵和停泵导致的缓发中子先驱核流失所需改变的控制棒棒位。同时还计算了2 MW恒定功率情况下稳态运行及降低流速时一回路温度分布,并模拟了2 MW额定功率下停泵事件。停泵后由于缓发中子损失减少反应堆功率先缓慢增加,然后迅速降低到接近余热水平。停泵后堆芯温度缓慢增加后稳定在安全值以内,说明熔盐堆具有本征安全性。     

加速器驱动次临界堆的安全和控制特性初步研究

加速器驱动次临界反应堆的结构特点使其安全和控制特性有别于临界反应堆。本工作使用数值计算和仿真运行方法,驱动堆的安全和控制特性进行了初步研究。结果表明：驱动堆不易发生瞬发临界,其安全特性优于临界堆,次临界度越深,安全性越好;驱动堆控制回路具有小的时间常数和超调量,调整时间短,控制特性亦优于临界堆。     

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.     

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.     

Active oxygen control by a PbO mass exchanger in the liquid lead–bismuth eutectic loop: MEXICO

Accurate control of dissolved oxygen concentration is crucial in order to use liquid lead alloys as a coolant of advanced nuclear systems. An oxygen control system based on PbO mass exchanger (PbO MX) technology was implemented in order to control the dissolved oxygen concentration in the liquid lead--bismuth eutectic (LBE) loop MEXICO. The oxygen control system consisted of a packed bed of PbO spheres, an oxygen sensor and a pneumatic control valve. The concentration of dissolved oxygen in the loop was controlled by regulating the LBE flow through the PbO MX using a proportional–integral–derivative (PID) controller with feedback from the oxygen sensor. Highly accurate control of the dissolved oxygen concentration in the loop was achieved by this system.     

Model experimentation and analysis of flow-induced vibrations of PWR internals

In a PWR the reactor coolant flow that goes through the reactor internals and the fuel assemblies is characterized by high turbulence and this flow is able to induce some structural vibration. A few years ago, some nuclear power plants were obliged to shut down for many months, due to the heavy damage caused by vibration. The design of reactors must be carefully checked taking into account the possible interaction between hydraulic excitation and reactor structure response. The reactor assembly of a PWR consists of: (1) a reactor vessel which withstands the internal pressure of the primary fluid and maintains the reactor core; (2) reactor internals which maintain fuel assemblies, guide the control rods and wear a thermal shield in order to reduce the fast neutron exposure of the reactor vessel wall; and (3) fuel assemblies and control rods.The SAFRAN test loop consists of a reduced-scale ( ) model of a reactor vessel, reactor internals, dummies representing fuel assemblies and a system of three loops including pumps and damping tanks connected to the reactor vessel, the purpose of which is to simulate the flow distribution of a three-loop PWR. The scaling laws for designing the model and the test loop are: same geometry and attachment conditions; same flow velocity: V model = V reactor; same Cauchy number, i.e. same ratio of inertia forces to stiffness forces; and same Euler number, i.e. same ratio of inertia forces to pressure forces. Nevertheless, it is not possible to use the same Reynolds number. The ratio between the Reynolds number of the reactor and the Reynolds number of the model, for the same fluid velocity, is 70. This is mainly due to scale ratio and to the viscosity of the fluid in the hot condition. But in most cases, we are above the critical values of Reynolds number where there is a variation of the Strouhal number S = ƒD/V. The measured frequencies in the model will be eight times the frequencies occurring in the reactor. In general, the construction technology used for the model is the same as that used for the reactor. All the structures in contact with the fluid are made of stainless steel. The instrumentation used on the SAFRAN test loop consists of accelerometers, pressure sensors and relative displacement sensors.Vibration phenomena are studied using two different approaches. In the first approach, the vibration properties of the structure are measured by means of tests performed in air and water to obtain, in both cases, frequencies, modes, damping and stiffness values. The hydraulic excitation sources are measured by tests on the loop: frequencies, Δp values, direct- and cross-correlation lengths. During these tests, structures are stiffened in order to prevent their motion. By means of a computer program based on the POWELL method, the structural response can be calculated according to the density of Δp distributed around the structure. The second approach consists of measuring directly the structural response to hydraulic excitations. Comparison of the results given by these two approaches shows: (a) the system non-linearities and (b) the coupling between the fluid and the structure. By using two different approaches a better knowledge of complex phenomena can be gained.     

小型核动力装置自然循环运行特性分析

本文以小型一体化自然循环反应堆为研究对象,用RELAP5/MOD3.2对反应堆系统、中间回路及二回路系统进行建模,对反应堆单双环路切换及偏环路运行时反应堆的自然循环运行特性进行数值模拟研究。计算结果表明:在反应堆自然循环运行工况下,进行单双环路切换及偏环路运行时,堆芯能重新建立稳定的自然循环。双环路切换至单环路后,堆芯出口温度降低,堆芯自然循环平衡流量降低但仍大于初始值的1/2;单环路切换至双环路运行时,堆芯流量、温度均与双环路稳定工况的一致;偏环路运行时故障环路循环流量降低,正常环路自然循环流量升高,堆芯总流量降低的数值为二者之差。