Investigation of space-dependent neutron-flux fluctuation excited by a strong absorber

The time response of neutron flux to the sudden displacement of an absorber in a reactor is studied by 1-D one-group treatment, and by applying the method to an example, it is shown that the flux depression around the absorber follows rapidly the displacement of the absorber, but there is some time delay in reaching a steady flux in regions farther from the absorber.The space dependence of the neutron noise induced by vibrating a strong absorber is studied by considering the flux depression which follows the moving absorber. The correction to the amplitude of the peak for the vibrating strong absorber is derived. It is shown that the bi-frequency term in the first order of displacement of the absorber vanishes because of the characteristic of the Green's function.     

Determination of thermocouple transfer-functions and fluid-flow velocities by temperature-noise measurements in liquid sodium

The time response of TI-, TM- and intrinsic thermocouples has been investigated in sodium by determination of frequency dependent thermocouple transfer functions and related delay times. Experimentally temperature fluctuations (temperature noise), generated in sodium by heated channels or injection of hot sodium, have been used as driving sources. The measured transfer functions and delay times have been compared to predictions from theoretical models. Good agreement was found. Intrinsic and TM-type thermocouples are best suited if fast response is required. Delay times in the range of one millisecond or upper break frequencies of 1000 Hz were determined for intrinsic thermocouples.

The determination of sodium flow velocities by the transit time correlation technique showed systematic error sources for analysis at low frequencies due to intermediate heat storage between the thermocouple positions. Statistical errors have been determined as function of thermocouple distance, frequency range and time of analysis. Minimum errors are obtained for thermocouple distances from 100 to 150 mm and in the frequency range 90 to 120 Hz. A transit time correlator, which automatically tracks the maximum of the cross-correlation function and displays the velocity, has been successfully used for signals with correlation coefficients greater than 0.3.     

Design and development of hydrogen isotope sensor in liquid PbLi

Indian LLCB – TBM uses liquid Lead-Lithium (Pb-Li) as tritium breeder, neutron multiplier and coolant. Tritium bred in liquid PbLi stream has to be recovered by tritium extraction system. Therefore, a reliable sensor with quick response time for measurement of hydrogen isotope is essential.A hydrogen isotope sensor in liquid Pb-Li, based on permeation of hydrogen isotopes through metal (sensor material) is designed. The capsule shaped sensor, made of iron membrane coated with Pd from inside (downstream side), allow hydrogen isotope to permeate through it. The design work mainly includes the selection of proper material, its thickness and surface conditions, which is to be supported by numerical calculations for optimization of maximum permeation flux, fast response and fabricability. The numerical calculation utilizes a physical model having recombination of two hydrogen isotope atoms at the surface and atomic diffusion through the bulk. In this work, design calculations based on numerical simulation and fabrication procedure of the hydrogen isotope sensor are presented.     

压水堆严重事故下封头热斑计算

压水堆堆芯熔化事故情况下,下封头热斑会造成压力容器局部过热,导致临界热流密度发生。利用FLUENT软件对堆芯熔化事故时的下封头热斑进行计算,从流动和换热角度预测热斑导致的下封头薄弱环节。计算结果表明:堆芯熔化事故时,压力容器下封头存在两处最薄弱的位置,分别为下封头正下方正对外部冷却水位置和氧化壳与压力容器交界处。特别是在氧化壳与压力容器交界处,由于多种原因导致临界热流密度发生,使得该处熔化严重。通过设置延伸小管和附加冷却水可延迟压力容器壁面熔穿的时间。     

竖直管道内间歇式两相流动沸腾特性分析

自然循环或重力注水过程的热功率、冷却剂流量等操作条件较小,易出现各种流动不稳定现象,影响核反应堆事故的发展进程,间歇式流动沸腾现象就属于其中的一种。以去离子水为工质,采用2×2加热棒束,对内径为32 mm竖直通道内的间歇式流动沸腾现象进行了实验研究,分析了不同热流密度下间歇式流动沸腾不稳定现象的变化规律,讨论了热流密度对间歇式沸腾周期的影响。结果表明,在一定的热流密度条件下,当加热通道内流体达到饱和并过热时,会发生周期性地剧烈喷涌及冷液回流现象,期间伴随泡状流、弹状流、搅混流及环状流等多种流动形态;间歇喷涌周期取决于沸腾停滞时间,随热流密度的不断增大,沸腾停滞时间缩短,间歇喷涌周期也缩短。当热流密度增大到一定程度时,间歇式流动沸腾现象消失,从而转变为另一种两相流动不稳定现象。     

Measurement of the coolant flow rate using correlation of temperature fluctuations

The transit time of the coolant, and thus its velocity, has been measured using the temperature fluctuation at the outlet of a reactor core. An impulse response function estimation is introduced, which substitutes the widely used cross-correlation measurement technique. It is shown in theory and practice that the time delay estimation is improved when using the impulse response function instead of the cross-correlation function in parameter estimation. Extremely low velocities (down to 2 cm/sec) have been measured in a natural circulation regime in a research reactor.     

Core melt down and vessel failure: a coupled problem

Analyses of lower head failure have been performed for a variety of core slump scenarios that result from three contrasting reactor accident sequences in a PWR. The cases cover a range of thermalhydraulic conditions in the vessel and core debris characteristics. The results show lower head failure occurs at a time which depends on the internal thermal-hydraulic conditions and debris characteristics. Failure may be local or global and may be due to one or more of the following processes: creep; plasticity (including thermo-plasticity); and melt-through. At low to moderate pressure, creep damage accumulates over a wide area, leading to probable global failure. Local plastic deformation becomes increasingly important at higher pressures or following a pressure spike, with a possibility of local failure. Local melting can occur before failure if there is a large concentrated heat flux. A question of particular interest for future study is raised by the CORVIS experiments, namely that the deformation can cause a gap to open between the structure and debris crust and hence increase the thermal resistance. Modest estimates of the gap resistance show a significant delay in failure. A coupled treatment of the thermal and mechanical response is needed to assess the dynamic gap behaviour effectively.     

Active radiative liquid lithium divertor concept

Developing a reactor compatible divertor has been identified as a particularly challenging technology problem for magnetic confinement fusion. Application of lithium (Li) in NSTX resulted in improved H-mode confinement, H-mode power threshold reduction, and reduction in the divertor peak heat flux while maintaining essentially Li-free core plasma operation even during H-modes. These promising Li results in NSTX and related modeling calculations motivated the radiative liquid lithium divertor (RLLD) concept [1]. In the RLLD, Li is evaporated from the liquid lithium (LL) coated divertor strike point surface due to the intense heat flux. The evaporated Li is readily ionized by the plasma due to its low ionization energy, and the poor Li particle confinement near the divertor plate enables ionized Li ions to radiate strongly, resulting in a significant reduction in the divertor heat flux. This radiative process has the desired effect of spreading the localized divertor heat load to the rest of the divertor chamber wall surfaces, facilitating divertor heat removal. The modeling results indicated that the Li radiation can be quite strong, so that only a small amount of Li (∼a few mol/s) is needed to significantly reduce the divertor peak heat flux for typical reactor parameters. In this paper, we examine an active version of the RLLD, which we term ARLLD, where LL is injected in the upstream region of divertor. We find that the ARLLD has similar effectiveness in reducing the divertor heat flux as the RLLD, again requiring only a few mol/s of LL to significantly reduce the divertor peak heat flux for a reactor. An advantage of the ARLLD is that one can inject LL proactively even in a feedback mode to insure the divertor peak heat flux remains below an acceptable level, providing the first line of defense against excessive divertor heat loads which could result in damage to divertor PFCs. Moreover, the low confinement property of the divertor (i.e., <1 ms for Li particle confinement time) makes the ARLLD response fast enough to mitigate the effects of possible transient events such as large ELMs.     

Measurement of response time and detection of degradation in pressure sensor/sensing-line systems

A team evaluated several methods for remote measurement of the response time and detection of degradation (blockage or air in lines) of pressure sensor/sensing line systems typical of nuclear power plants. A method was developed for obtaining the response time of force-balance pressure transmitters by briefly interrupting the power supply to the transmitter. The data thus generated are then analyzed in conjunction with a model to predict transmitter response to an actual pressure perturbation. The research team also evaluated a pressure perturbation method for determining the asymptotic delay time of a pressure-sensing line and found that this method yields accurate results for essentially unblocked sensing lines. However, these pressure perturbation tests are not recommended for use in nuclear power plants because they are difficult to implement on-line. A third method for remote measurement applied noise analysis methods that yielded accurate estimates of asymptotic delay times for blockage or air in sensing lines. Even though noise analysis methods worked well in the laboratory, it is recommended that further evaluation be performed in operating nuclear plants.     

A new method for estimating heat flux in superheater and reheater tubes

In this paper a procedure on how to estimate the heat flux in superheater and reheater tubes utilizing the empirical formula and the finite element modeling is proposed. An iterative procedure consisting of empirical formulae and numerical simulation is used to determine heat flux as both temperature and scale thickness increase over period of time. Estimation results of the heat flux over period of time for two different design temperatures of the steam and different heat transfer parameters are presented.     

温度传感器动态响应特性的测试与分析

本文分析了热电偶、热电阻等温度传感器动态热传递方程,影响传感器时间常数的因素以及预先测试的喷射装置;综述了时间常数就地测试的LCSR法和噪声分析法以及FFT、功率谱、窗函数、ARMA模型等信号处理技术在动态温度测量中的应用。     

An investigation of transition boiling mechanisms of subcooled water under forced convective conditions

A mechanistic model for forced convective transition boiling has been developed to investigate transition boiling mechanisms and to predict transition boiling heat flux realistically. This model is based on a postulated multi-stage boiling process occurring during the passage time of the elongated vapor blanket specified at a critical heat flux (CHF) condition. Between the departure from nucleate boiling (DNB) and the departure from film boiling (DFB) points, the boiling heat transfer is established through three boiling stages, namely, the macrolayer evaporation and dryout governed by nucleate boiling in a thin liquid film and the unstable film boiling characterized by the frequent touches of the interface and the heated wall. The total heat transfer rates after the DNB is weighted by the time fractions of each stage, which are defined as the ratio of each stage duration to the vapor blanket passage time. The model predictions are compared with some available experimental transition boiling data. The parametric effects of pressure, mass flux, inlet subcooling on the transition boiling heat transfer are also investigated. From these comparisons, it can be seen that this model can identify the crucial mechanisms of forced convective transition boiling, and that the transition boiling heat fluxes including the maximum heat flux and the minimum film boiling heat flux are well predicted at low qualities/high pressures near 10 bar. In future, this model will be improved in the unstable film boiling stage and generalized for high quality and low pressure situations.     

The use of the noise analysis for the detection of the pellet- cladding mechanical interaction

The pellet-cladding mechanical interaction (PCMI) has been classically studied with the cladding elongation signal versus the rod power, but a different method can be used to evaluate the PCMI. It is based on the noise analysis techniques and uses the coherence function between the neutron flux and the cladding elongation noise signals detecting PCMI. This technique has been studied in the Halden experimental reactor. The mean coherence versus average linear heat rate shows two different regions that can be fitted with two straight lines, with the crossing point between them being taken as the threshold for PCMI. In addition, noise analysis in the time domain allows us to measure an elongation response time to a power step. This result can be applied to power instability events in BWRs.     

二回路耗汽运行对船用堆全部电源丧失事故进程缓解分析

针对船用核动力装置的特点,建立了船用堆一、二回路及控制系统的RELAP5模型,用RELAP5/MOD3.2程序对典型船用堆经济巡航工况下发生全部电源丧失事故进行模拟,分析了4种耗汽工况对事故进程的缓解效果。分析表明：事故后合理的耗汽运行方案能明显延缓事故进程,延缓时间为小时级别;耗汽量越小,二回路设备运行时间越长,二回路热阱持续时间也越长,一回路事故进程越缓慢。但太小的耗汽量会引起事故过程中蒸汽发生器（SG）二次侧水位过高甚至满水,进而威胁二回路设备的正常运行。同时,二回路设备数量众多,不同设备的最低运行汽量也不尽相同,应选择最有用和最低耗汽量的设备耗汽运行。本研究能对实际船用堆事故下的应急处理提供参考。     

Properties of Active Water Cooling Atmospheric Plasma Spraying Tungsten Coating

Tungsten coating was deposited onto an actively water-cooled oxygen-free copper substrate by atmospheric plasma spraying technique. The properties of the microstructure, porosity, microhardness and oxidation of the active water cooling atmospheric plasma spraying tungsten coating were characterized and measured. High heat load and thermal response tests using an electron beam facility have been carried out on the tungsten coated mock-ups under the water cooling condition. The heat flux is changed from 3 to 10 MW/m² under the condition that the water flow rate is 2.5, 2.0 and 1.5 m³/h, respectively. It is demonstrated that the mock-ups successfully withstood a heat flux of 15.5 MW/m² at steady state.     

Heat Transfer to Steam-Water Annular Dispersed Flow in Vertical Annulus

The heat transfer coefficient and slow burnout heat flux were measured for a stream-water annular dispersed upward flow under pressures up to 3.5 ata in an electrically heated vertical annular channel.

An empirical equation was derived for the heat transfer coefficient as function of mass flow rate, steam quality and heat flux. The dominant mechanism of heat transfer to the annular dispersed two-phase flow is forced convection of liquid film on the heater surface even in the region of low steam quality (down to about 0.03). The observed slow burnout heat flux was near the point of intersection of the lines representing liquid film forced convective heat transfer and nucleate boiling heat transfer on the q vs. δT _sat diagram. A dryout mechanism is proposed in which increasingly violent evaporation comes to impede the rewetting of the dry patches generated on the heater surface, which thus spread to cover the whole surface. A maximum value is observed in the slow burnout heat flux plotted against exit steam quality. This can be explained as the effect of heat removal by droplet exchange between liquid film and steam flow.     

Non-Hydrodynamic Aspects of Pool Boiling Burnout

More than 400 pieces of test sections—flat plates made of various metals and alloys—were placed in saturated water under atmospheric pressure and heated to physical destruction by passing electric current directly through them.

Several non-hydrodynamic parameters have marked effect upon the critical heat flux, which indicates that purely hydrodynamic theories are not directly applicable to general prediction of the critical heat flux.

The critical heat flux data was found to be well correlated with the heat capacity per unit surface area of the test section: the critical heat flux is reduced with decrease of this parameter, while with increasing parameter it approaches a certain asymptotic value.

The fluctuation of surface temperature due to alternate contact with vapor and liquid was calculated by the proposed model, and the trend of dryout duration was estimated from the critical heat flux data.     

导热对不均匀边界放热及最大热流影响的分析和试验

文章阐述了导热和几何尺寸对平均传热系数和最大局部热流的影响,对之进行了分析并提供了研究试验的结果。计算分析和试验都是针对环形流道进行的。流道内侧为肋化管,外侧为光滑管。为进行计算分析,专门发展了二维计算程序“CONDU”。分析表明:导热和几何尺寸对平均传热系数和最大局部热流有重要影响。相似性分析表明,对于实践中有重要意义的某些场合,有必要引进一个与导热和几何尺寸有关的修正函数。此分析的主要结果已得到实验的证实。     

Eddy current effects in a high field dipole

Eddy currents produced by a time-varying magnetic field will introduce time delay and thus affect field quality. This effect leads to drifting of the beam position over time, especially for a compact synchrotron.Simulations and measurements of different dipoles have been performed, to investigate the time delay and field quality. The simulations are conducted using OPERA software. The measurements are conducted using a long coil and Hall sensor. All results show that the magnetic field deviation is up to 0.4% for the dipole with stainless steel endplates. The simulations show that the main sources of eddy current are the field saturation effect and the field component Bz, introduced by the bedstead-type coil. Field correction using a power supply is adopted to reduce the deviation to less than 0.02%.     

Development of Thin Film Thermometer and Its Application to Radiation Loss Measurement in JFT-2 Tokamak

A thin film thermometer with high sensitivity and small response time has been developed. A nickel film was deposited in a vacuum on a backing plate of stainless steel 0.05 mm thick. The pattern of nickel film was so designed to provide sufficient accuracy. A small heat flux of 0.5 W/cm² is measurable with a response time of less than 1 ms. The thermometer was applied to radiation loss measurement in JFT-2 tokamak. The power loss due to the radiation was found to be about 30% of the total power input.