船用堆自动棒动态反应性Simulink仿真模块设计

船用堆运行中功率频繁、剧烈变化需要自动控制棒频繁调节。针对该特点及现有反应堆系统微机仿真程序存在的控制棒反应性描述不合理、不准确的问题,设计了船用堆自动棒动态反应性Simulink仿真模块。该模块作为船用堆物理热工参数快速计算Simulink程序的子模块,应用于船用堆典型动态过程仿真表明：该模块能够模拟动态过程中的自动棒棒位和相应的动态反应性,适用于船用堆物理热工参数快速计算,对船用堆动态过程的仿真和物理热工参数快速计算有重要意义。     

高温和超高温气冷堆动力转换方案研究

动力转换单元是高温和超高温气冷堆的重要组成部分。本文对高温和超高温气冷堆的动力转换单元进行研究。从4个关键参数(反应堆出口温度、反应堆入口温度、压缩比和主蒸汽参数)入手,对5个循环方案进行比较分析。综合考虑各种工程因素,上位循环为简单氦气透平循环、下位循环为有再热的蒸汽轮机循环的联合循环方案是具有竞争力的,其中下位循环在高温气冷堆范围是亚临界参数循环,在超高温气冷堆范围是超临界参数循环。联合循环可实现高温和超高温气冷堆热量的高效率转化,且反应堆入口温度在反应堆压力壳材料允许的范围内,具有足够的安全性。     

高温气冷堆板翅式回热器的分布参数模型

板翅式回热器是高温气冷堆直接氦气透平循环中的重要部件.针对逆流板翅式回热器,提出了一个一维的分布参数模型来研究其动态性能,并用整场离散、整场求解的方法求解了该模型.通过对回热器高温侧入口温度扰动的模拟和分析,发现与扰动同侧出口温度的响应过程可以用一个带滞后的惯性环节来表达,而与扰动异侧出口温度的响应过程可以用一个过渡过程时问极短的惯性环节来表达.对于流量扰动而言,回热器出口温度的动态特性也可以采用惯性环节来表达.在此基础上,利用该模型模拟了一个温度和流量耦合变化的响应过程.结果表明该模型不仅能用于简单扰动的模拟,还可以用于复杂扰动的模拟和分析.     

高温气冷堆全耦合系统直接联立求解的方法研究和程序开发

反应堆耦合计算是对现有反应堆各领域数值技术的融合、集成和提升,完整的反应堆核电站系统同时具有多种耦合机制,是一个超大规模非线性强耦合系统,以JFNK/NK为代表的直接联立方法是极具潜力的发展方向。本文在综述国内外反应堆耦合计算研究的基础上,介绍了清华大学核能与新能源技术研究院在高温气冷堆核电站全耦合直接联立求解方法及程序开发方面的研究工作。针对高温气冷堆多物理、多尺度、多部件、多回路、多模块的耦合特点,首次提出了非线性消去直接联立方法等关键技术,研发可以描述多层级耦合结构的统一耦合平台框架,已形成多个中间版本的程序。     

高温和超高温气冷堆动力转换方案研究

动力转换单元是高温和超高温气冷堆的重要组成部分。本文对高温和超高温气冷堆的动力转换单元进行研究。从4个关键参数（反应堆出口温度、反应堆入口温度、压缩比和主蒸汽参数）入手，对5个循环方案进行比较分析。综合考虑各种工程因素，上位循环为简单氦气透平循环、下位循环为有再热的蒸汽轮机循环的联合循环方案是具有竞争力的，其中下位循环在高温气冷堆范围是亚临界参数循环，在超高温气冷堆范围是超临界参数循环。联合循环可实现高温和超高温气冷堆热量的高效率转化，且反应堆入口温度在反应堆压力壳材料允许的范围内，具有足够的安全性。     

HTR-10一回路流量变化试验的模拟

10 MW高温气冷实验堆(HTR-10)是我国第一座高温气冷堆。一回路流量变化试验是HTR-10的三个动态特性试验之一,该试验不仅证明了反应堆的功率自调节性能,也为系统分析程序的验证提供了实测数据。基于实际的试验工况,利用THERMIX程序对一回路流量变化试验进行了模拟,分析了反应堆主要参数的变化。关于反应堆功率,计算结果与试验结果符合得很好,证明程序能够满意地再现HTR-10在该试验中的动态特性。试验过程中,燃料元件中心最高温度始终低于1 230℃的温度限值。     

船用一体化反应堆自然循环过渡过程控制优化研究

为了研究反应堆强迫循环向自然循环转换过程中功率自动调节方案,利用建立的数学模型对某型船用一体化反应堆自然循环过渡过程进行了理论计算。结果表明:过渡过程中,由于一体化压水堆自然循环工况下反应堆进、出口温度的滞后效应明显,并且冷却剂温度在功率自动调节模型中的权重大,使得反应堆进、出口温度测量点的位置对重要参数的峰值产生较大影响,反应堆进、出口温度测量点设置越接近反应堆活性区,则过渡过程中重要参数的波动峰值越小,过渡过程需要的时间越短,控制过程更优。     

无限大芯部热容板翅式回热器动态性能研究

针对高温气冷堆直接氦气透平循环中的板翅式回热器，研究提出一仅考虑回热器芯部热容的集总参数模型，即无限大芯部热容的集总参数模型，并利用四阶龙格 库塔方法求解该模型，求解过程中考虑温度对气体物性的影响。利用该模型，分析了在入口温度、流量阶跃和斜坡扰动下回热器出口温度的响应过程。在此基础上，分析了高温气冷堆直接氦气透平循环中的功率调节过程及透平甩负荷过程时回热器出口温度和芯部温度的响应过程。     

模块式高温气冷堆超临界循环一次再热方案研究

基于模块式高温气冷堆先进技术和超临界蒸汽动力循环先进技术,研究了高温气冷堆模块与超临界蒸汽动力循环耦合配置方案。结合超临界热力循环理论及模块化高温气冷堆的特性,研究了超临界热力循环方案及相应的循环参数。针对标准一次再热循环,研究了反应堆模块与汽轮机组匹配模式;计算了循环可能达到的效率,并与先进压水堆效率进行了比较。结果表明：模块化高温气冷堆超临界循环效率比压水堆电厂约高30%。本研究结果可作为高温气冷堆超临界循环电站概念设计的理论基础,为进一步的技术研究与方案设计提供依据。     

热管反应堆堆芯缩比模块瞬态特性分析

热管堆具有长寿期、高可靠性等优势,是当下空间核反应堆的研究焦点之一。为研究热管堆瞬态过程中的核热耦合现象,本文基于半物理仿真技术,搭建了针对热管反应堆堆芯缩比模块的核热耦合实验平台,通过实验模块测量了堆芯缩比模块的温度分布,在仿真模块中基于点堆模型计算了输出功率随时间的变化情况。通过耦合实验模块和仿真模块,探索了瞬态条件下堆芯缩比模块核热耦合特性,分析了引入不同初始反应性时堆芯温度、加热功率和剩余反应性的瞬态演变过程,揭示了系统热容量造成的温度迟滞变化效应,即热惯性现象。结果表明,堆芯缩比模块的热惯性随引入的初始反应性的增大及初始功率水平的增加而减小,且与基体材料的热扩散率呈反比。     

A system analysis tool with a 2D gas turbine modeling for the load transients of HTGRS

This paper presents the operational performance and transient response of a high temperature gas-cooled reactor (HTGR) with an emphasis on the gas turbine through a two-dimensional approach. For its operational and transient simulation we use a GAMMA-T in which the system code, GAMMA, is coupled with the two-dimensional turbomachinery model. We also implement several models into the GAMMA-T: the reactor kinetics model, the bypass valve model, and the models of the core, the heat exchangers, the gas turbine, and the piping. The estimations of compressor and turbine performances are based on a two-dimensional axisymmetric throughflow method that is capable of predicting both the transient and steady-state behavior of the power conversion system (PCS). To demonstrate the code capability, we investigated the two representative transients of GTHTR300, which is a 600 MW direct cycle helium cooled reactor consisting of a prismatic block type core, a horizontal single-shaft configuration of turbomachinery, a recuperator, and a precooler: a loss of heat rejection transient corresponding to the failure of the precooler water supply, and a 30% load reduction transient from nominal operation with bypass control. The simulation results demonstrated the controllability and operational stability for the plant.     

核-气联合循环发电系统性能仿真分析

针对压水堆核电机组循环热效率较低及电网对核电调峰能力的需求，基于Ebsilon软件，在大亚湾核电站二回路热力系统模型基础上，建立核-气联合循环发电热力系统。以燃气轮机循环效率、联合循环效率作为热经济性指标，评价联合循环系统的性能，并分析环境温度、压力及燃气轮机负荷变化对系统性能的影响。结果表明:核-气联合循环系统热效率相比原核电机组提高13.15%，汽轮机输出功率增加75.49%，工作环境得到明显改善；环境温度降低或压力升高会提高燃气轮机效率及联合循环功率；燃气轮机降负荷时，通过补燃天然气可维持核蒸汽发生器进口温度不变，汽轮机仍有较高的输出功率，负荷可调节范围为56.57%~100%。      

A spatial kinetic model for simulating VVER-1000 start-up transient

An accurate prediction of reactor core behavior in transients depends on how much it could be possible to exactly determine the thermal feedbacks of the core elements such as fuel, clad and coolant. In short time transients, results of these feedbacks directly affect the reactor power and determine the reactor response. Such transients are commonly happened during the start-up process which makes it necessary to carefully evaluate the detail of process. Hence this research evaluates a short time transient occurring during the start up of VVER-1000 reactor. The reactor power was tracked using the point kinetic equations from HZP state (100 W) to 612 kW. Final power (612 kW) was achieved by withdrawing control rods and resultant excess reactivity was set into dynamic equations to calculate the reactor power. Since reactivity is the most important part in the point kinetic equations, using a Lumped Parameter (LP) approximation, energy balance equations were solved in different zones of the core. After determining temperature and total reactivity related to feedbacks in each time step, the exact value of reactivity is obtained and is inserted into point kinetic equations. In reactor core each zone has a specific temperature and its corresponding thermal feedback. To decrease the effects of point kinetic approximations, these partial feedbacks in different zones are superposed to show an accurate model of reactor core dynamics. In this manner the reactor point kinetic can be extended to the whole reactor core which means “Reactor spatial kinetic”. All required group constants in calculations are prepared using the WIMS code. In addition CITATION code was used to calculate the flux, power distribution and core reactivity inside the core. To update the last change in group constants and resultant reactivity in point kinetic equations, these neutronic codes were coupled with a developed dynamic program. This study is applied on a typical VVER-1000 reactor core to show the reactor response in short time transients caused during start-up procedure.     

Model development and simulation of transient behavior of helium turbine system

This paper describes a performance model for the transient analysis of helium turbine system. Governing equations have been derived from integral forms of unsteady basic conservation equations. The one-dimensional model is employed for flow-paths except turbine and compressor, which are considered as zero-dimensional components and volume-less treatment is employed. Component mathematical model results in a set of ordinary differential equations and algebraic equations. The simulation code is established on MATLAB, and the ordinary differential equations are solved a variable order solver of MATLAB, ode15s. The accidents of loss of load and loss of feedwater to precooler and intercooler, the transients of recuperator and the decreasing heat transfer capacity of intermediate heat exchanger are simulated respectively. The analysis of calculated results verifies the present model. The effects of bypass valve size and thermal inertia of the recuperator wall are also studied. The simulation results show that throttle size of bypass valve has important influence on the characteristics of turbine system and should be carefully selected to satisfy the requirement of system control and safety.     

Validation of coupled neutron kinetic/thermal-hydraulic codes. Part 2: Analysis of a VVER-440 transient (Loviisa-1)

Several three-dimensional hexagonal reactor dynamic codes have been developed for VVER type reactors and coupled with different thermal-hydraulic system codes. Under the auspices of the European Union's Phare programme these codes have been validated against real plant transients by the participants from 7 countries. Two of the collected five transients were chosen for validation of the codes. Part 1 of this article consists of validation against VVER-1000 reactor data. This second part is focussed to validation against measured data of ‘One turbo-generator load drop experiment' at the Loviisa-1 VVER-440 reactor. The experiment was performed just after plant modernisation and more measured data was available to validation than in normal operation of real plants. Good accuracy of the results was generally achieved comparable to the measurement accuracy. The confidence in the results of the different code systems has increased, and consequences of certain model changes could be evaluated.     

Comparison of two models for a high temperature reactor coupled to a gas turbine

This paper presents two independent dynamic models of a nuclear gas turbine power plant. Both the high temperature nuclear reactor (HTR) and its energy conversion system (ECS) based on a direct Brayton cycle have been modelled. One model utilises RELAP5 for the ECS, the other Aspen Custom Modeler (ACM). The reactor model used in both models is a point kinetic model derived from a detailed reactor model. The ECS model is described and compared componentwise, with an emphasis on the turbomachinery. The total plant models are compared with each other by calculating two representative transients: one load rejection transient and one transient with the system at part load.     

Comparison of two models for a pebble bed modular reactor core coupled to a Brayton cycle

The pebble bed modular reactor (PBMR) plant is a promising concept for inherently safe nuclear power generation. This paper presents two dynamic models for the core of a high temperature reactor (HTR) power plant with a helium gas turbine. Both the PBMR and its power conversion unit (PCU) based on a three-shaft, closed cycle, recuperative, inter-cooled Brayton cycle have been modeled with the network simulation code Flownex.One model utilizes a core simulation already incorporated in the Flownex software package, and the other a core simulation based on multi-dimensional neutronics and thermal-hydraulics. The reactor core modeled in Flownex is a simplified model, based on a zero-dimensional point-kinetics approach, whereas the other model represents a state-of-the-art approach for the solution of the neutron diffusion equations coupled to a thermal-hydraulic part describing realistic fuel temperatures during fast transients. Both reactor models were integrated into a complete cycle, which includes a PCU modeled in Flownex.Flownex is a thermal-hydraulic network analysis code that can calculate both steady-state and transient flows. An interesting feature of the code is its ability to allow the integration of an external program into Flownex by means of a so called memory map file.The total plant models are compared with each other by calculating representative transient cases demonstrating that the coupling with external models works sufficiently. To demonstrate the features of the external program a hypothetical fast increase of reactivity was simulated.     

Dynamic model for control system design and simulation of a low temperature nuclear reactor

A lumped parameter dynamic model for the primary-loop and the U-tube steam generator of a low temperature power reactor is developed based on the fundamental conservation laws of fluid mass, energy and momentum. The dynamic model is formulated by coupling the point kinetics with reactivity feedback and the thermal-hydraulics of the reactor. The developed dynamic model is implemented on a personal computer using MATLAB/SIMULINK. Numerical simulation results for steady-state and transient responses are then presented, which show that the steady-state precision of the newly developed dynamic model is acceptable and the trend of the transient responses is correct. In addition, the “swell and shrink” behavior of the U-tube steam generator is also verified by numerical simulation. This newly established model can be utilized to control system design and simulation for the low temperature power reactor.     

Validation of coupled neutron kinetic/thermal–hydraulic codes. Part 1: Analysis of a VVER-1000 transient (Balakovo-4)

Three-dimensional hexagonal reactor dynamic codes have been developed for VVER type reactors and coupled with different thermal–hydraulic system codes. In the EU Phare project SRR1/95 these codes have been validated against real plant transients by the participants from several countries. Data measured during a test in the Balakovo-4 VVER-1000 have been analysed by coupled codes. In the test, one of two working feed water pumps of the steam generators was switched off at nominal power. The steady-state assembly powers measured before and after this transient are reproduced by the codes with a maximum deviation of about 5%. The time behaviour of the most safety-relevant parameters, such as total fission power, coolant temperatures and pressures is well modelled. Thermal–hydraulic feedback effects observed in the measurement are described by the codes in a consistent manner. The analyses have shown, that an accurate treatment of the heat transfer from the fuel rods to the coolant is important. In all, the results have increased the confidence in the coupled code analyses of VVER-1000 transients.