高温气冷堆主氦风机惰转特性研究

高温气冷堆主氦风机与压水堆主泵一样,均为反应堆一回路的关键设备.在反应堆正常工况下,两者具有相同的功能要求,但在事故工况时,因反应堆的运行特性不同,其功能要求各异.目前,对压水堆主泵的惰转特性已有大量的实验研究和实际运行结果,但有关主氦风机惰转特性的实验研究与理论研究还很缺乏.本文结合风机的气动特性与高温气冷堆一回路的阻力特性,从理论上研究高温气冷堆主氦风机的惰转特性,建立主氦风机惰转时的流量与转速的预测公式,并给出其数值预测结果,为高温气冷堆设计的初步安全分析提供依据.     

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.     

核电厂一回路系统模拟实验设计与分析

研究核电站特定运行工况下,一回路系统传热流动的规律。应用Ishii模化方法模拟压水堆核电厂的一回路系统,设计出主泵与关联系统耦合实验回路的主要热工参数。同时,应用机理性程序对设计的实验回路进行分析。结果表明,基于Ishii模化方法设计的实验回路主要参数合理可行;模型可以研究反应堆原型事故运行瞬态工况下,一回路各系统间传热流动相互影响规律。     

Parametric evaluation of large-scale high-temperature electrolysis hydrogen production using different advanced nuclear reactor heat sources

High-temperature electrolysis (HTE), when coupled to an advanced nuclear reactor capable of operating at reactor outlet temperatures of 800-950 °C, has the potential to efficiently produce the large quantities of hydrogen needed to meet future energy and transportation needs. To evaluate the potential benefits of nuclear-driven hydrogen production, the UniSim process analysis software was used to evaluate different reactor concepts coupled to a reference HTE process design concept. The reference HTE concept included an intermediate heat exchanger and intermediate helium loop to separate the reactor primary system from the HTE process loops and additional heat exchangers to transfer reactor heat from the intermediate loop to the HTE process loops. The two process loops consisted of the water/steam loop feeding the cathode side of a HTE electrolysis stack, and the sweep gas loop used to remove oxygen from the anode side. The UniSim model of the process loops included pumps to circulate the working fluids and heat exchangers to recover heat from the oxygen and hydrogen product streams to improve the overall hydrogen production efficiencies.The reference HTE process loop model was coupled to separate UniSim models developed for three different advanced reactor concepts (a high-temperature helium cooled reactor concept and two different supercritical CO₂ reactor concepts). Sensitivity studies were then performed with the objective of evaluating the affect of reactor outlet temperature on the power cycle efficiency and overall hydrogen production efficiency of the integrated plant design for each of the reactor power cycles. The results of these sensitivity studies showed that overall power cycle and hydrogen production efficiencies increased with reactor outlet temperature, but the power cycles producing the highest efficiencies varied depending on the temperature range considered.     

Thermal hydraulic simulation of reactor of HTR-PM based on thermal-fluid network and SIMPLE algorithm

This paper provides the model development and its verifications for the reactor thermal-hydraulic transient model for the High Temperature Gas-cooled Reactor Pebble-bed Module (HTR-PM). A thermal-fluid network is constructed to simulate the complexity of the flow and heat transfer structure in the reactor of HTR-PM. SIMPLE algorithm is applied to solve the conservation equations of the thermal-fluid network to simulate the transient behavior of the high speed turbulent helium flow. A calculation process is proposed for coupling the high speed helium flow and the complex heat transfer structure. A FORTRAN code was developed based on the solution method and the thermal-fluid network. Several test cases including two steady states and a Control Rod Withdrawal accident are simulated by this code and the results are compared to those obtained by a safety analysis code, namely THERMIX. The good agreement between the two codes indicates that the proposed model and solution method based on SIMPLE algorithm is reasonable and applicable for simulating the thermal-hydraulic behavior in reactor of HTR-PM.     

Analysis of the turbine deblading in an HTGR with the CATHARE code

The direct coupling of a Gas Cooled Reactor (GCR) with a closed gas-turbine cycle leads to a specific dynamic plant behaviour. This behaviour is described and illustrated through computer analyses performed at CEA with the computer code CATHARE. This analysis requires a 1D code able to simulate the whole reactor, including the core, the vessel, the piping and the components (turbine, compressors, heat exchangers).This paper is devoted to deblading accidents. The problems and solutions encountered in various types of gas-turbines are presented: aero engines, steam turbines (EDF-Porcheville steam turbine accident feedback) and finally the feedback from previous High Temperature Gas Reactor experiments (EVO helium loop, HHT project and other HTR projects) are displayed. From this literature survey, some recommendations are drawn for a future High Temperature Reactor. It is shown that for safety reasons in case of deblading, a horizontal shaft aligned with the reactor vessel is recommended for the turbomachinery.This paper presents simulations of different scenarii performed with CATHARE code:

(a) Turbine deblading with and without reactor trip. In these calculations, a pessimistic assumption has been made: all turbine blades break off.

(b) Total flow blockage. The flow area is entirely blocked by the turbine blades.

(c) Partial flow blockage. Previous conclusions consider the worst cases of deblading and total flow blockage. An intermediate case based on a partial deblading has also been performed.

CATHARE results illustrate that the loss of turbine blades is accompanied by abrupt changes in the Power Conversion System and reactor flow conditions: large axial pressure drop, reverse flow through the core and high rate de-pressurization.     

Experiments on the Heat Transfer and Natural Circulation Characteristics of the Passive Residual Heat Removal System for an Advanced Integral Type Reactor

Experiments on the heat transfer characteristics and natural circulation performance of the passive residual heat removal system (PRHRS) for the SMART-P have been performed by using the high temperature/high pressure thermal-hydraulic test facility (VISTA). The VISTA facility consists of the primary loop, the secondary loop, the PRHRS loop, and the auxiliary systems to simulate the SMART-P, a pilot plant of the SMART. The primary loop is composed of the steam generator (SG) primary side, a simulated core, a main coolant pump, and the loop piping, and the PRHRS loop consists of the SG secondary side, a PRHRS heat exchanger, and the loop piping. The natural circulation performance of the PRHRS, the heat transfer characteristics of the PRHRS heat exchangers and the emergency cooldown tank (ECT), and the thermal-hydraulic behavior of the primary loop are intensively investigated. The experimental results show that the coolant flows steadily into the PRHRS loop and that the heat transfers through the PRHRS heat exchanger and the emergency cooldown tank are sufficient enough to enable a natural circulation of the coolant. The results also show that the core decay heat can be sufficiently removed from the primary loop with an operation of the PRHRS.     

Numerical Study on Tritium Behavior by Using Isotope Exchange Reactions in Thermochemical Water-Splitting Iodine—Sulfur Process

One potential problem in the hydrogen production system coupled with the high-temperature gascooled reactor (HTGR) is transmission of tritium from the primary coolant to the product hydrogen by permeation through the heat transfer tubes. Tritium accumulation in the process chemicals in the components of a hydrogen plant, a thermochemical water-splitting iodine-sulfur (IS) process, will also be a critical issue in seeking to license the hydrogen plant as a non-nuclear plant in the future. A numerical analysis model for tritium behavior in the IS process was developed by considering the isotope exchange reactions between tritium and the hydrogen-containing process chemicals, i.e., H₂O, H₂SO₄ and HI. The tritium activity concentration in the IS process coupled with the high-temperature engineering test reactor (HTTR), the HTTR-IS system, was preliminarily evaluated in regard to the effects of some indeterminate parameters, i.e., equilibrium constants of the isotope exchange reactions, permeability of tritium through heat transfer tubes, tritium and hydrogen concentrations in the secondary helium coolant, and the leak rate from the secondary coolant loop. The results describing how the tritium activity concentration changes with variations in these parameters and which component has the maximum tritium activity concentration in the IS process are described in this paper.     

核电厂蒸汽发生器二次侧三维流场分析

为弄清核电厂蒸汽发生器二次侧的流动和传热特性机理,以确保蒸汽发生器的稳定性,文章采用数值软件根据蒸汽发生器的结构特点和运行模式进行简化建模,利用相似原理,使用相变模块模拟了蒸汽发生器的二次侧汽液两相流的流场分布情况。研究了相同结构下不同给水比例对二次侧流场分布的影响,尤其是对空泡份额分布特性的影响。研究发现,不同的给水工况对直管段的空泡份额分布和流体流速分布都有明显的影响,但对传热管上部区域的空泡份额和速度分布的影响不大。     

Thermal and hydraulic tests of test sections in the helium engineering demonstration loop

The helium engineering demonstration loop (HENDEL) has been constructed and operated to test the large-scale components of the high temperature engineering test reactor (HTTR) under simulated reactor operating conditions. The fuel stack test section (T₁) of HENDEL simulates the fuel stack of the HTTR core and is used to investigate thermal and hydraulic performance. Hot tests with 1000°C helium gas have been conducted using simulated fuel rods having uniform, exponential and cosine axial heat flux distributions. The test results agreed with previously proposed correlations, although the simulated fuel rods had various heat flux distributions and high heat flux rates.

The in-core structure test section (T₂) also was installed in the HENDEL to verify the performance of the core bottom structure of the HTTR. The tests show that good performance was obtained. Examination of the thermal mixing characteristics indicated that mixing started at the location where the hot helium gas flowed into the hot plenum and that complete mixing was achieved during the downward flow in the outlet hot gas duct. The seal performance testing indicated no change of the leakage flow rate after 4000 hours of operation. The temperature of the metal portion of the structure was below 500°C and uniform around circumferential cross-sections due to the good performance of the thermal insulation blocks.     

细长自然循环系统流动不稳定性实验研究

以水为工质,在常压下对拥有细长回路和较长水平段的自然循环系统进行可视化实验研究,并以典型的实验现象（ P =1.46 kW）为例分析该系统的瞬态运行特性和不稳定性机理。结果表明：阻力系数较大的细长自然循环回路难以产生有效的单相自然循环,只能通过间歇性沸腾和两相流动将热量导出。这是因当回路阻力较大时,过冷沸腾产生的驱动力无法驱动回路产生有效的自然循环,而只有当加热段内流体发生饱和沸腾时才能驱动系统产生循环流动。较大的回路阻力和沸腾过程中产生的系统降压闪蒸是细长自然循环系统难以维持稳定的流动驱动压头从而产生间歇性沸腾和强烈流动不稳定性的根本原因。     

HTR-10蒸汽发生器的建模及分析

为开发适用于球床模块式高温气冷堆HTR-10的模拟机,采用一体化仿真支撑平台vPower建立了蒸汽发生器的实时热工水力模型。模型以传热方程为基础求解两侧工质及金属管壁的温度和焓,以流体网络为基础求解两侧工质的压力和流量。本文讨论了3种节点划分方案,针对不同节点划分方案的适用范围提出了建议并采用96节点划分方案进行后续研究。此外,通过分析确认了模型在稳态工况下主要参数和分布参数的准确性和合理性,并在100%功率稳态工况的基础上模拟了氦气侧流量阶跃的场景,分析了模型中主要参数的变化过程。动态仿真结果表明,氦气流量阶跃会引起一、二次侧参数不同程度的变化,变化幅度与阶跃程度呈正比,且金属管壁和水侧热容、二次侧参数变化速率相对缓慢,模型再平衡时间较短,表明HTR-10采用的螺旋管式直流蒸汽发生器的热惯性相对较小。     

倒U型管蒸汽发生器自然循环回流特性研究

自然循环在核动力工业中有重要的应用。以AP1000自然循环压水反应堆一回路系统的蒸汽发生器为对象,建立自然循环系统蒸汽发生器倒U型管内单相水回流的数学模型,并进行了计算,得出了蒸汽发生器倒U型管内正流和回流流量、正流管管长和回流管管长分布等。研究结果表明,回流能大幅降低自然循环的重位压头、流量和放热量,不利于自然循环。     

Experimental and numerical investigations on the reverse flow phenomena in UTSGs

Aiming at the reverse flow phenomena in the inverted U-tube steam generators (UTSGs), the experimental and numerical simulations are performed. A new method is developed to model the flow and heat transfer in the steam generator based on the system analysis code RELAP5/MOD3.3. The reverse flow phenomenon observed experimentally is simulated well by the new method. The experimental and numerical results show that the reverse flow occurs in the adjacent shorter U-tubes. For single U-tube, the mass flow rate of reverse flow is generally greater than that of normal flow U-tube. When the reverse flow occurs, the negative pressure drop between the inlet and outlet plenums and the heat transfer of the UTSG reduce significantly. The numerical simulations also show that the reverse flow occurs more easily in UTSGs with the bigger tube length ratio.     

氦气杂质对中间换热器材料的影响初探

中间换热器是高温气冷堆氦气透平间接循环和高温工艺热应用的关键部件.中间换热器属于一回路压力边界,它将堆芯出口温度达900～1 000 ℃氦气的热量传递给二回路氦气,此外还承受一、二回路氦气压差,因此,目前能够用于中间换热器的耐热金属材料非常有限.高温气冷堆一、二回路氦气中含有H_2、H_2O、CO、CH_4等杂质,在高温下,氦气杂质对中间换热器材料的影响主要是氧化、碳化和脱碳,降低材料的机械性能,其影响不可忽视.对于中间换热器设计,现有规范的温度范围需扩展,氦气杂质对材料强度的影响也需考虑.     

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。      

蒸汽发生器实验回路倒流特性仿真分析

针对立式倒U型管蒸汽发生器传热管内出现的倒流现象,基于RELAP5/MOD3.3程序,采用新的控制体划分方案对蒸汽发生器实验段进行建模,模拟实验回路中发生的倒流现象。通过与实验数据进行对比分析,验证建模方案的正确性。在此基础上,分析倒流现象对蒸汽发生器实验段流动传热的影响。结果表明:倒流现象发生在较短管内,对于单个U型管,倒流管的流量高于正流管。倒流发生后,系统进入相对稳定状态,但蒸汽发生器实验段的换热功率和进出口腔室负压降绝对值显著降低。     

Development of TSACO and application to Chinese HCCB TBM cooling system

In this study, a thermal-hydraulic and safety analysis code (TSACO) for helium cooling system has been developed using Fortran 90 language, and the simulation has been performed for the cooling system of the Chinese helium cooled ceramic breeder test blanket module (CH HCCB TBM). The semi-implicit finite difference technique was adopted for the solution of the dynamic behavior of helium cooling system. Furthermore, a detailed illustration of the numerical solution for heat structures and critical model was presented. The code was verified by the comparison of RELAP5 code with the same initial condition, boundary condition, heat transfer and flow friction models. The TBM inlet/outlet temperatures and pressure drop were obtained and the results simulated by TSACO were shown in good agreement with those by RELAP5. Thereafter, the design basis accident in-vessel loss of coolant accident (LOCA), was investigated for the CH HCCB TBM cooling system. The critical flow model was also verified by comparing with RELAP5 code. The results indicated that the TBM can be cooled down effectively. The vacuum vessel (VV) pressure and the mass of helium spilled into the VV maintained below the design limits with a large margin.     

Simulation of two regime sodium flows in a parallel plate channel during buoyancy-driven flow

Due to the remaining decay heat, the reactor core has yet to be cooled after shutdown. As the reactor power is low, the core can be sufficiently cooled by natural convection. The coolant flow is driven by buoyancy, as the heated fluid decreases its density. During buoyancy-driven flow, a reverse flow may take place when a heat sink exists close to the heat source, such as in a wall (edge) or corner subchannels. For simplicity in applying boundary condition, the reverse flow is simulated by two parallel plates, one as a heat source having positive heat flux and the other as heat sink with negative heat flux.     

燃料组件稳态回路辐照考验多热工参数耦合特性研究

燃料组件稳态回路辐照考验是研究组件抗辐照性能的关键环节,鉴于回路辐照考验热工参数对试验运行的重要性,本文针对稳态辐照考验回路,结合燃料组件对回路辐照考验热工参数的需求,分析了一次水流量及入口温度发生变化、装置内部存在换热以及不同换热器运行方式下回路系统的关键热工参数耦合特性。研究表明,随着换热器一次水入口温度和流量的下降,主换热器的最大换热功率有明显下降,在严重偏离设计工况下,主换热器换热能力存在不能满足燃料组件辐照运行的风险。同时,辐照装置内部换热对主换热器换热极为不利。在回路运行温度需求较高,而装置内部换热较强情况下,主换热器二次水流量的设计裕量不能低于装置内的热交换比例,且该裕量取值需趋大。两台主换热器并联运行的优势主要体现在较大一次水流量情况,且存在一个较小的一次水流量,使得单台换热器独立运行与两台换热器并联运行时的换热能力一致,低于该流量比例时,两台换热器并联运行时的换热能力反而弱于单台换热器运行。