浮动核电厂反应堆供电系统分析

The safety of the floating nuclear power plant is closely related to the merits of the reactor power supply system. In order to improve the safety factor of floating nuclear power plant, it is necessary to analyze the reactor power supply system. In this paper, the configuration of the reactor power supply system of the floating nuclear power plant is analyzed, and the auxiliary power system and nuclear emergency power system under two schemes are compared. The results show that the optimized scheme 2 is better than scheme 1 in reliability and safety, and scheme 2 is more economical. The optimization scheme proposed in this paper can provide a direction for the design of the reactor power supply system of the subsequent nuclear powered ships and has a strong reference significance.     

浮动核电厂反应堆供电系统分析

浮动核电厂的安全性与反应堆供电系统的优劣紧密相连,为提高浮动核电厂的安全系数,需对反应堆供电系统进行分析。本文结合反应堆供电系统设计的要点,分析浮动核电厂反应堆供电系统的配置,对比了2种方案下的辅助电力系统和核应急电力系统。结果表明,优化后的方案2在可靠性和安全性方面都较方案1更优,且方案2也更为经济。本文提出的优化方案可为后续核动力船舶反应堆供电系统的设计提供参考和借鉴。     

Mode-C运行与控制模式设计技术研究

本文基于运行与控制模式设计,结合核电厂的运行需求,针对国内压水堆核电厂以基负荷运行方式为主、负荷跟踪运行需求较少的特点,首次开展了与之适应的Mode-C运行与控制模式设计。通过控制策略设计、控制棒设置设计、核电厂运行方式设计、核电厂运行范围设计等设计步骤,研究Mode-C运行与控制模式的设计技术。结果表明：采用Mode-C模式的压水堆核电厂能根据负荷变化需求选择执行单变量自动控制模式或双变量自动控制模式,实现了设定的控制策略,Mode-C运行与控制模式的设计技术在反应堆物理专业方面是可行的。     

大型先进压水堆核电站堆芯装载方案设想

核电站堆芯装载方案是反应堆堆芯设计的重要基础,它首先必须满足核安全的要求,同时还要尽可能地提高经济性。通过分析国内、外百万千瓦级核电站的堆芯装载,对反应堆输出功率、燃料组件数、堆芯平均线功率密度进行比较,给出我国大型先进压水堆核电站示范工程反应堆堆芯装载方案的设想,为技术决策提供参考。     

基于频域法的核电站堆功率调节器的设计

堆功率调节系统是核电站反应堆控制的核心。本文以反应堆为对象,建立堆功率调节系统仿真模型,采用频域法设计核电站堆功率调节器。该方法相对于传统PID控制结构,具有设计步骤清晰、简单,以及更加宽泛的工业控制要求等优点。仿真结果表明,该方法跟踪调节性能好、鲁棒性强,能消除不可测干扰,对反应堆参与电网调峰具有现实意义。     

基于和睦系统的ACPR1000核电机组保护系统优先级管理系统方案设计

通过梳理DI&C-ISG-04标准和ACPR1000核电机组对优先级管理系统（CIS）的功能要求，参考CPR1000核电机组CIS方案中的经验反馈，分析CIS设计中的诸多难点，提出了一套基于和睦系统（FirmSys）的ACPR1000核电机组保护系统CIS方案。相比CPR1000核电机组，ACPR1000核电机组的CIS方案采用了多种设计手段进行优化，如自主化和国产化设计、更优的防软件共因故障设计、更可靠的在线定期试验设计、模块化及多接口设计、就地功能设计。该设计方案已成功应用于ACPR1000核电机组，运转良好。因此本研究提出的CIS方案切实可行。      

月球基地核电源系统方案研究

月球基地的建立首先需解决能源供给问题,核电源系统具有功率大、寿命长、环境适应性强等优点,是月球基地及其他深空探测任务中理想的能源提供方案。本文分析了目前可用于月球基地的能源情况,有针对性地提出40 kWe月球基地核电源系统的设计理念,经初步优化设计,给出了该系统的流程和总体设计参数,并从物理、屏蔽、热工以及结构方面对系统进行了分析和论证。结果表明,该系统方案合理可行,能满足安全和寿期要求。     

Coordination control and simulation for small nuclear power plant

Compared with nuclear electric factory, marine nuclear power plant has some particular features including smaller size, faster response, and stronger load following capacity etc. This paper focuses on marine nuclear power plant. Based on static mathematical models of some important parts such as reactor core, steam generator etc., a coordination control system is designed to implement its rapid following and response when power changes. According to the Matlab/Simulink simulation, this new scheme improves fast response capacity of the control system, which contributes to the practical system design.     

火星表面核反应堆电源方案研究

火星作为地球的近邻,一直是人类进行深空探测和开发的首选目标。登陆火星、建立火星基地并开发利用火星资源首先需解决电力供给问题,核反应堆电源具有功率大、寿命长、环境适应性强等优点,是火星基地及其他深空探测任务的理想能源。通过分析目前可用于火星基地的能源情况,针对性地提出了40 kWe火星表面核反应堆电源的总体方案,并从物理、安全、屏蔽、热工以及结构方面对电源方案进行了分析和论证。结果表明：该电源方案合理可行,能满足安全和寿期要求。     

Modeling the nonlinear dynamic temperature and power oscillations in VVER

The nonlinear dynamic auto-oscillatory regimes of the temperature at the exit from the core and of the power of nuclear reactors are investigated and the quantitative features of the regimes are determined. A physical-mathematical model of the dynamics of a VVER-type nuclear reactor is examined, taking account of the inertial (temperature) and instantaneous (power) feedbacks in the reactor but neglecting the delayed neutrons. Analytical solutions characterizing the auto-oscillatory real-time behavior of the temperature at the exit from the core and of the reactor power are obtained. The period of these oscillations is determined. The model constructed for the dynamics of a reactor is used as a basis to simulate, using the RELAP5/Mod3.2 computer code, certain nonlinear auto-oscillatory processes occurring in the first and second loops of a nuclear power plant with a VVER-440 reactor. It is found that the oscillations of the temperature at the exit from the core and power of the nuclear reactor generate the same period for processes which were calculated.     

Optimal power control system of a research nuclear reactor

The power control system is a key control system for a nuclear reactor, which directly concerns the safe operation of a nuclear reactor. Much attention is paid to the power control system performance of nuclear reactor in engineering. The designers put a high value upon design of an optimal power control system. In this paper, a design method is applied to the design of power control system. According to the optimal control theory, an objective function, quadratic performance index with weight factors is proposed. Then, the objective function is transformed into frequency domain form by use of Paserval's theorem. In frequency domain, an optimal transfer function can be obtained at the lowest value of objective function. The system with optimal transfer function has an optimal performance. The transfer function of the power control system is derived from a typical research nuclear reactor. Using the state feedback theory, the transfer function is synthesized to the optimal transfer function. The simulative results with the optimal controller and with a conventional controller show that the performance of the optimal power control system is largely improved on dynamic characters. The method applied here not only can be used for research nuclear reactor but also can be easily extended to pressurized water reactor power plant and other fields.     

用于海上钻井平台的小型钠冷快堆核电源概念设计方案

根据海上石油钻井平台用户电力需求的特点,介绍了一种基于斯特林热气机发电技术的小型钠冷快堆核电源设计方案,研究了小型钠冷快堆核电源的总体技术方案、主回路冷却系统以及关键设备设计方案,并给出小型钠冷快堆核电源的初步布置方案。研究结果表明：小型钠冷快堆核电源概念设计方案符合海上石油钻井平台用户需求的长周期换料、空间限制等特点。     

Boiling water vessel reactors for nuclear district heating

The thermal and economic performance indices of a present-day low-capacity reactor facility with a boiling-water vessel reactor are presented. The operation of a reactor with natural coolant circulation is highly reliable in different operating regimes of a nuclear heat and power plant.     

海上浮动堆燃料组件进入堆芯仿真分析

与陆上核电厂不同,海上浮动堆换料操作会受海浪环境的影响,因此对换料操作工艺和设备提出了新要求。本文选取海洋核动力平台的海上换料方案,对燃料组件在摇摆工况进入堆芯过程进行了仿真分析。分析结果表明,引入万向节的燃料组件进入堆芯过程中,燃料组件满足强度设计要求。     

Evaluation and optimization of General Atomics’ GT-MHR reactor cavity cooling system using an axiomatic design approach

The development of the Generation IV (Gen-IV) nuclear reactors has presented social, technical, and economical challenges to nuclear engineering design and research. To develop a robust, reliable nuclear reactor system with minimal environmental impact and cost, modularity has been gradually accepted as a key concept in designing high-quality nuclear reactor systems. While the establishment and reliability of a nuclear power plant is largely facilitated by the installment of standardized base units, the realization of modularity at the sub-system/sub-unit level in a base unit is still highly heuristic, and lacks consistent, quantifiable measures. In this work, an axiomatic design approach is developed to evaluate and optimize the reactor cavity cooling system (RCCS) of General Atomics’ Gas Turbine-Modular Helium Reactor (GT-MHR) nuclear reactor, for the purpose of constructing a quantitative tool that is applicable to Gen-IV systems. According to Suh's axiomatic design theory, modularity is consistently represented by functional independence through the design process. Both qualitative and quantitative measures are developed here to evaluate the modularity of the current RCCS design. Optimization techniques are also used to improve the modularity at both conceptual and parametric level. The preliminary results of this study have demonstrated that the axiomatic design approach has great potential in enhancing modular design, and generating more robust, safer, and less expensive nuclear reactor sub-units.     

AP1000核电厂灵活循环燃料管理策略研究及应用

我国核电装机容量逐年稳步扩增,核电厂参与电网调峰愈加频繁,固定的换料周期逐渐难以满足核电厂经济运行的需求。本文基于AP1000核电厂18个月堆芯装载方案,设计了±1个月和±2个月的灵活周期堆芯装载方案,完成方案的安全性限值与燃料经济性评价,开展完整的安全分析。结果表明,堆芯设计满足安全相关验收准则的要求,全面论证了灵活循环燃料管理策略的安全性和可行性。本研究为AP1000核电厂灵活循环周期运行提供了技术支撑,灵活循环周期运行即将在海阳核电厂中工程应用。     

ACP600主蒸汽管道破裂事故的应对措施研究

针对ACP600取消高压安注系统和浓硼箱、使用一体化钆为可燃毒物、采用Mode-C运行与控制模式等设计改进导致主蒸汽管道破裂（MSLB）事故安全裕量降低的不利情况,对先进三代核电厂ACP600的MSLB事故进行分析研究。为避免MSLB事故下反应堆重返临界后堆芯功率峰值过高导致偏离泡核沸腾比（DNBR）低于限制值,分别从快速注入硼溶液和减缓堆芯冷却率的角度,评价不同的安注系统配置以及停运故障环路主泵对于缓解MSLB事故的作用。研究最佳的缓解方案,并提出增设“蒸汽管线压力低-3”信号停运故障环路主泵的设计优化建议。      

铅铋合金系统氧控旁路性能数值研究

为研究通过固态氧控有效调节铅铋合金（LBE）系统氧浓度的方法,本文通过修正液态LBE腐蚀经验公式,结合氧化铅（PbO）溶解模型,基于集总参数法并使用FORTRAN语言自编程序计算LBE系统氧浓度;据此研究主回路流量、质量交换器（MX）内温度、PbO装量对自主设计的小型LBE系统的MX供氧性能的影响;初步建立MX设计准则,获得一定约束条件下MX供氧性能及其氧控旁路设计参数。本研究可为LBE系统氧控旁路的设计和计算提供参考,同时提供一种高效求解LBE系统瞬态氧浓度和腐蚀的计算方法,为建立氧浓度模型预测系统提供新思路。      

ACP100S浮动核电站总体设计及验证

浮动核电站是利用浮动平台建造的可移动的核电站,从电功率分级上划分属于小型反应堆,可用于发电、淡化海水、供热,能满足区域供电、区域供热、海上石油开采、偏远地区、孤岛等特殊需要。本文在分析目前浮动核电站发展形式及其特点、优势基础上,提出了中核集团ACP100S浮动核电站方案,分析ACP100S设计原则、技术特点、设计参数、总体技术方案、试验和验证以及示范工程进展情况。     

Design of Test Scheme for Failure Rate Verification of Control Rod Drive Mechanism Drop

In the event of the reactor accident, the control rod drive mechanism (CRDM) must be able to drop quickly and achieve the safe shutdown. In order to ensure the safety of the reactor, the nuclear power plant puts toward higher requirements on the failure rate of the CRDM. In this paper, the design algorithm of the test scheme for the failure rate of the CRDM and the corresponding solution method based on the dichotomy method were proposed. The correctness was verified by comparison with the scheme given in GB 5080.5-85. It supplements the design algorithm of the fixed-end censored test scheme for success rate verification test scheme of the success or failure type products in GB 5080.5-85. Based on any given success rate, risk value and discrimination ratio, it can give the censored test protocol in a short time. It also designs the success rate verification test scheme for general success or failure type products to meet the increasing reliability requirements for success or failure type products.