岭澳核电站2号机组一回路流量测量偏差分析及处理

针对岭澳核电站2号机组调试期间发生的二环路3台流量计指示全部超量程故障现象．从反应堆冷却剂流量测量原理入手．对一回路流量测量回路及现场实测数据进行了分析研究,并进行了多次现场试验及功率平台验证,指出二环路流量测量故障的根本原因是其对应弯管流量计的机械制造偏差．提出了对现有流量变送器的量程进行迁移并重新计算和整定相应测量回路参数作为临时解决措施．同时提出了在第一个换料大修期间更换适当量程的流量变送器作为永久解决方案。     

Soft-measuring method of the core exit temperature of coolant for the 200 MW nuclear heating reactor

This paper reports a soft-measuring method of the core exit temperature of coolant for Chinese 200 MW nuclear heating reactor (NHR-200). The primary and vice sheath thermocouple are immersed in a space orthogonal slot that is located at one side of the support grid plate for fuel assemblies. The core exit temperature of the coolant is evaluated by using these two thermocouple's measurement temperatures. The experimental study gives the formula for evaluating the core exit temperature of the coolant. The space orthogonal slot is smooth to decrease the hydraulic resistance of the coolant, thus a part of coolant was steady flowing through the space orthogonal slot on the support grid plate. So the coolant temperature difference between the center region of fuel assembly and the measurement end of the primary sheath thermocouple is small. The flow of the coolant in the slot increases also the sensitive length of temperature of the sheath thermocouple. All of these ensure the measurement reliability and accuracy. The maximum measurement error of the core exit temperature of the coolant for the NHR-200 is 1.7 K.     

Heats of Vaporization of UO2 and UO from UO2

In an earlier paper, a stochastic model of a power reactor has been proposed by the present author on the premise that the coolant-flow through a core is usually accompanied by random variations in the flow-rate, which are eventually largely responsible for the internal reactivity fluctuations.

In the present work, this model is extended to three different reactor systems: (a) where there exists a relaxation process corresponding to the effect of buoyant flow; (b) where a control or fuel element vibrates randomly, due to coolant flow-rate fluctuations; (c) where there are fluctuations in the inlet temperature with a non-white spectrum.

The noise spectra are derived for various state quantities with use made of the Langevin procedure. The theory is illustrated by referring chiefly to the neutron noise spectra, and comparing with the results of observations.

It is shown that the noise sources in question contribute significantly to the spectra, as compared with a low frequency component due to an inherent noise source in the coolant flow. In particular, a strong resonance peak of the spectra arises from the coupling between the random mechanical vibrations and coolant the flow-rate fluctuations.     

反应堆冷却剂系统流量测量试验研究与设计

华龙一号（HPR1000）设置了反应堆冷却剂泵进出口压差表用于测量反应堆冷却剂系统（RCS系统）环路流量,取消了二代改进型核电机组设置的弯管流量计。环路流量测量方式的改变直接影响RCS系统流量测量试验的实施。通过研究主泵的运行特性和系统的阻力特性,提出了基于主泵电功率测量RCS系统流量的试验方法。结合理论分析结果和工程实践经验,给出了反应堆冷却剂惰走流量试验的试验方法和验收准则。研究表明,主泵电功率法可以测量RCS系统的流量,反应堆冷却剂惰走流量可以通过主泵惰转过程的转速变化进行验证。      

一回路动态排气剩余空气体积标准值提升后冷却剂流动特性研究

针对动态排气后提升一回路剩余空气体积标准值的改进方案,提出含高溶解度空气的冷却剂在主泵启动瞬态下的压力预测方法和是否释放为两相分离流动的判断方法,对一回路及其辅助系统进行热工水力建模,空气体积标准值提升为24标准立方米（1标准立方米=1.293 kg）后,对主泵启动的瞬态过程进行了仿真,得到了一回路主要节点压力变化规律;结合冷却剂中气体溶解-释放模型,得到饱和氮气溶解度、氧气溶解度变化规律。结果表明,主泵启动瞬态过程中,一回路主要节点压力均在机组运行正常范围内,一回路中溶解的氮气、氧气不会释放成为两相流动。因此,就流动特性而言,空气体积标准值提升到24标准立方米可行。      

Simulation study on adaptive modeling in a reactor heat-transfer channel

Adaptive modeling of the dynamic relationship between coolant flow-rate and outlet temperature in an FBR heat-transfer channel has been studied by digital simulation. The aim of this study was to detect heat-production anomalies in a faster, more direct way by, to a great extent, eliminating the flow-rate effect on outlet temperature.     

Thermal-hydraulic analysis of accidents leading to local coolant flow decrease in the main circulation circuit of RBMK-1500

There are a few transient and loss-of-coolant accident conditions in RBMK-1500 reactors that lead to a local flow decrease in fuel channels. Because the coolant flow decreases in fuel channels (FC) leads to overheating of fuel claddings and pressure tube walls, mitigation measures are necessary. The accident analysis enabled the suggestion of the new early reactor scram actuation and emergency core cooling system (ECCS) initiation signal, which ensures the safe shutdown of the reactor and compensates the stagnation flow. Analysis of such conditions is presented in this paper. Thermal-hydraulic analysis was conducted using the state-of-the-art RELAP5 code. Results of the analysis demonstrated that, after implementation of the developed management strategy for destruction of local flow stagnation, the Ignalina nuclear power plant (NPP) would be adequately protected following accidents, leading to local coolant flow decrease in the primary circuit.     

A method of evaluating local turbulent diffusivity of heat in a gridded fuel subassembly

In evaluating the turbulent diffusivity of heat associated with the coolant flow past a grid spacer within an FBR fuel subassembly, a heat diffusion technique is usually employed. However, measurement of subchannel bulk coolant temperature using thermocouples usually involves difficulty due to a steep and non-linear temperature gradient in the subchannels adjacent to a heater pin.A series solution of the heat conduction equation for the coolant flow in subchannels past a grid spacer and a heated section of a dummy fuel pin was derived under a slug flow approximation where the boundary conditions on dummy fuel pins were satisfied by means of the point-matching technique. The solution may be utilized in analyzing the turbulent diffusivity of heat within subchannel coolant flow as a function of distance from a grid spacer based on the measured temperature distribution on the wall of dummy fuel pins, which may be obtained without affecting the subchannel coolant temperature.In an illustrative example, the turbulent diffusivity of heat was most exaggerated at about 50 mm beyond a grid spacer and was approximately five times larger than the corresponding diffusivity without a grid spacer.     

Improved core design of the high temperature supercritical-pressure light water reactor

A new coolant flow scheme has been devised to raise the average coolant core outlet temperature of the High Temperature Supercritical-Pressure Light Water Reactor (SCLWR-H). A new equilibrium core is designed with this flow scheme to show the feasibility of an SCLWR-H core with an average coolant core outlet temperature of 530 °C.In previous studies, the average coolant core outlet temperature was limited by the relatively low temperature outlet coolant from the core periphery. In order to achieve an average coolant core outlet temperature of 500 °C, each fuel assembly had to be horizontally divided into four sub-assemblies by coolant flow separation plates, and coolant flow rate had to be adjusted for each sub-assembly by an inlet orifice. However, the difficulty of raising the outlet coolant temperature from the core periphery remained.In this study, a new coolant flow scheme is devised, in which the fuel assemblies loaded on the core periphery are cooled by a descending flow. The new flow scheme has eliminated the need for raising the outlet coolant temperature from the core periphery and removed the coolant flow separation plates from the fuel assemblies.     

Thermal Experiments on a Model of a Target for an Accelerator-Driven System

The results of experiments performed on a model of a window target of an accelerator-driven system are presented. The model, the special features of the structure, and the measurement systems and methodological approaches are briefly described. A eutectic sodium–potassium alloy is used to simulate the lead–bismuth eutectic alloy. The following characteristics were measured directly in the experiments or obtained by analyzing the experimental data: coolant flow rate, power, absolute coolant temperature as a function of distance from the target membrane, the absolute temperature of the membrane surface as a function of the distance from the membrane center, the standard deviations of the indicated quantities and the pulsations of the coolant and membrane temperatures. The measurements showed that large temperature pulsations are observed on the membrane surface; these must be taken into account when analyzing the strength characteristics of a real target setup.     

Prospects for the development of a direct-flow vessel reactor with superheated steam

A system of new technical solutions, which will make it possible to develop a reactor system using a coolant with supercritical parameters and efficiency 44%, solving the main problems of the nonuniformity of coolant heating in the core, is examined. This sytem contains a core based on fuel micropellets, a multistep fuel assembly with transverse coolant flow, coolant mixing after each step, and continual reloading of fuel micropellets into each fuel assembly. __________ Translated from Atomnaya énergiya, Vol. 100, No. 3, pp. 197–204, March, 2006.     

百万千瓦级核电站轴封型主泵整机集成试验验证优化方案研究

为确保主泵的安全性和可靠性,主泵整机在完成集成设计后需通过试验进行验证。文中介绍了主泵设计与验证的总体思路,提出了主泵工程样机需开展的整机试验项目。基于已有的试验条件,进行了主泵整机集成试验验证方案的优化和可行性分析。分析结果表明,在充分开展主泵各模块、子模块、部件和材料的试验与分析的基础上,可采用小流量试验方案进行主泵整机集成试验验证。     

Experimental Investigations of the Thermohydraulic Parameters of the Flow Section of a Liquid-Metal Target with a Window for a Proton Beam from an Accelerator

The results of computational and experimental investigations of the thermohydraulic characteristics of a liquid-metal target with a tight-fitting stopper, whose shape ensures a constant energy-release volume in the stopper material, are examined. The investigations on water circulation stands included flow visualization and measurement of the velocity and pressure distributions in the flow part of the target structure. The investigations on the liquid-metal circulation stand with lead-bismuth coolant were performed with coolant working temperature 300°C and maximum flow rate up to 7 m³/h. The temperature and the temperature pulsations in the coolant and in the material of the tight-fitting stopper were determined.     

基于CFD分析的文丘里流量计设计优化

质量流量是核电站热功率核算的关键参数之一,核电站一般采用文丘里流量计和孔板流量计同时测量,然而在低流量区文丘里流量计呈现出明显波动,其参数不稳定严重影响核电站的正常运行。本文基于理论分析结合数值分析,发现脉动流动是导致文丘里流量计测量波动的主要原因。基于分析结果,对文丘里流量计提出了优化设计方案,通过在文丘里管上游集成流量调整装置,从而减小脉动流,有效提升文丘里流量计的稳定性。此外,开展了集成不同类型流量调整装置的文丘里流量计压损特性数值研究,结果表明K-Lab型流量调整装置阻力较小。本文提出的方案可有效提升文丘里流量计测量精度。     

Core design of a high-temperature fast reactor cooled by supercritical light water

A high-temperature large fast reactor cooled by supercritical water (SCFR-H) is designed for assessing its technical feasibility and potential economical improvement. The coolant system is once-through, direct cycle where whole core coolant flows to the turbine. The goal is to achieve the high coolant outlet temperature over 500°C. We study the reactors with blankets cooled by ascending and descending flow. SCFR-H adopts a radial heterogeneous core with zirconium-hydride layers between the driver core and the blankets for making coolant void reactivity negative. The coolant outlet temperature of the core with blankets cooled by ascending flow is low, 467°C. The reasons are as follows: (1) the power swing due to the accumulation of fissile material in the inner blankets with burn-up, and (2) local power peak in the assemblies due to the zirconium-hydride layers. The difference in the outlet coolant temperature is more enhanced than the low temperature core where outlet temperature is approximately 400°C. The reason is that the coolant temperature is more sensitive to the enthalpy change than near the pseudo critical temperature, 385°C at 25 MPa. Thus, we design the core with blankets cooled by descending flow to obtain high coolant outlet temperature. The coolant outlet temperature becomes 537°C, which is 70°C higher than that of the core with ascending blanket flow. The thermal efficiency is improved from 43.2 to 44.6%. The coolant mass flow rate per electric power decreases by 14%. This will reduce the size of the balance of plant (BOP) system. The power of the reactor is high (1565 MWe) and the void reactivity is negative.     

System for Observing the Accuracy and Reliability of Temperature Measurements in VVÉR-440

A system for observing the accuracy and reliability of planned temperature measurements in a VVÉR-440 reactor and the advantages of using such a system are described. Planned measurements are measurements of the coolant temperature at the fuel assembly exit and in loops introduced into the in-reactor monitoring system. The system is installed in two V-230 reactors and two V-213 reactors.     

Sensitivity analysis of an impedance void meter to the void distribution in annular flow: a theoretical study

Impedance void meters are frequently used to measure the area-averaged void fraction in pipes. This is primarily for two reasons: firstly, this method is non-intrusive since the measurement can be made by electrodes flush mounted in the walls, and secondly, the signal processing equipment is simple.Impedance probes may be calibrated by using a pressure drop measurement or a quick closing valve system. In general, little attention is paid to void distribution effects. It can be proved that in annular flow, the departure from radial symmetry has a strong influence on the measured mean film thickness. This can be easily demonstrated by solving the Laplace equation for the electrical potential by simple analytical methods. When some spatial symmetry conditions are encountered, it is possible to calculate directly the conductance of the two-phase medium without a complete calculation of the potential. A solution of this problem by using the separation of variables technique is also presented. The main difficulty with this technique is the mixed nature of the boundary conditions: the boundary condition is both of Neumann and of Dirichlet type on the same coordinate curve. This formulation leads to a non-separable problem, which is solved by truncating an infinite algebraic set of linear equations.The results, although strictly valid in annular flow, may give the correct trends when applied to bubbly flow. Finally, the theory provides an error estimate and a design criterion to improve the probe reliability.     

基于相依竞争失效模型的主泵多状态可靠性分析

基于核动力主泵运行环境和性能退化机理,考虑自身振动和外部冲击对其性能退化的影响,建立了主泵冲击与退化相依竞争失效过程的可靠度模型。采用该模型计算了考虑性能退化的主泵在振动和外部冲击条件下的退化状态概率和可靠度,为基于使用环境的核动力主泵的多状态可靠性分析提供了一种有效的分析途径。分析结果可为设计变更和维修优化提供决策依据。     

Plan for first phase of safety demonstration tests of the High Temperature Engineering Test Reactor (HTTR)

Safety demonstration tests using the High Temperature Engineering Test Reactor (HTTR) will be conducted for the purpose of demonstrating inherent safety features of High Temperature Gas-cooled Reactors (HTGRs) as well as providing the core and plant transient data for validation of HTGR safety analysis codes. The first phase safety demonstration test items include the reactivity insertion test and the coolant flow reduction test. In the reactivity insertion test, which is the control rod withdrawal test, one pair out of 16 pairs of control rods is withdrawn, simulating a reactivity insertion event. The coolant flow reduction test consists of the partial loss of coolant flow test and the gas circulators trip test. In the partial loss of coolant flow test, primary coolant flow rate is slightly reduced by control system. In the gas circulators trip test one and two out of three gas circulators are run down, simulating coolant flow reduction events. The gas circulators trip tests, in which position of control rods are kept unchanged, are simulation tests of anticipated transients without scram (ATWS).