M310压水堆的改进--不调硼负荷跟随研究

对大亚湾核电站和岭澳核电站的M310压水堆进行了不调硼负荷跟随研究.使用西屋公司APA堆芯核设计软件.从分析负荷跟随运行时的反应性变化入手,根据不调硼负荷跟随的需要重新设计控制棒价值和控制棒分组,在不改变M310压水堆现有控制棒数量和位置的前提下,实现不调硼负荷跟随.通过人为引入燃耗倾斜,并改进过渡过程,使M310压水堆不仅在实施不调硼负荷跟随时轴向偏移能够满足G模式的梯形图,同时还具备良好的实时反应能力.将这种不调硼负荷跟随加G模式梯形图的运行模式称为BTP运行模式(BTP为"不调硼"的汉语拼音缩写).从原理上证明在M310压水堆上BTP运行模式是可行的.     

Mode-C从基负荷到负荷跟踪的过渡策略研究

Mode-C在基负荷运行时与Mode-G类似,而在负荷跟踪运行时与MSHIM类似,综合了Mode-G和MSHIM的优点。但为满足反应性控制需求,Mode-C在循环末期由基负荷向负荷跟踪过渡时容易出现轴向偏移等难以控制的问题。在负荷变化过程中,控制棒抽插、中子注量再分布和氙反馈等影响反应性和轴向偏移控制的各因素之间是紧密耦合的,难以采用将各影响因素分离后单独计算分析的方法。为此,以某双环路压水堆的典型日负荷跟踪为例,通过对临时调硼、提前调硼和基负荷深插K棒组3种策略进行直接模拟计算研究。结果表明:只有基负荷深插K棒组策略可快速平稳过渡到负荷跟踪工况,且该策略还有循环寿期长、过渡时调硼量少及无需修改参考轴向功率差等优点。     

长寿期小型棒控压水堆控制棒方案研究

小型棒控压水堆舍弃了可溶硼,并高度依赖控制棒与可燃毒物棒控制堆芯的反应性。为研究控制棒对堆芯关键性能的影响,本文以核动力破冰船用KLT-40模型为对象,以轴向功率偏移、堆芯寿期、燃料利用率与径向功率峰因子为指标,开展长寿期小型棒控压水堆控制棒布置与动作策略设计分析。首先,基于OpenMC程序开发带棒燃耗程序;其次,比较堆芯带控制棒与无控制棒运行时的堆芯寿期等指标;最后,分析不同动作策略对轴向功率偏移等指标的影响。结果表明：控制棒将堆芯寿期从590 EFPDs（等效满功率天,Effective full power days）延长至650～698 EFPDs;低价值棒组优先动作策略使轴向功率偏移程度由-0.69与+0.80分别下降至-0.29与+0.52。因此,要准确计算长寿期压水堆寿期必须采用带控制棒燃耗计算策略,并且通过合理的动作策略能够有效减小控制棒带来的轴向功率偏移。     

CNP600堆芯调峰运行时轴向功率偏差影响因素及控制策略研究

随着我国能源的发展,核电将面临着越来越多的调峰运行压力,如采用CNP600堆型的海南昌江核电站,轴向功率偏差ΔI是调峰运行中堆芯功率控制的难点之一。针对CNP600堆芯,分析了慢化剂温度、控制棒棒位、可溶硼浓度等参数对ΔI的影响规律。以此为基础,采用程序模拟了CNP600 3种可能的调峰运行方案,提出了ΔI的控制策略,给出了功率变化过程中ΔI的具体控制方案。研究结果可为CNP600堆芯调峰运行时的堆芯ΔI的控制提供技术支持。     

压水堆核电厂负荷跟踪系统设计与特性研究

压水堆核电厂通过功率控制系统调节反应堆的反应性,以达到负荷跟踪的目的。本文设计的功率控制系统利用模糊控制器对棒速和硼浓度的联合控制做出最优选择,并利用功率补偿通道加快响应速度。MATLAB(Matrix Laboratory)的仿真结果证明该系统具有优良的负荷跟踪特性。利用电力系统分析综合程序(PSASP)的自定义模型功能,将该控制系统模块接入核电厂全系统模型,仿真结果表明压水堆核电厂的负荷跟踪能力可达到行业技术标准,并能满足电网的日负荷调峰要求。     

多模式机械补偿运行特性研究

经济、灵活及高自动化的运行控制策略是先进核电厂的设计目标之一。为适应这一发展趋势,近年来国际上提出了机械补偿运行控制的设计理念。机械补偿是一种主要通过控制棒的移动补偿堆芯反应性变化和控制轴向功率分布的先进控制策略。本文分别在基本负荷、负荷跟踪及启动/再启动运行模式下对机械补偿运行特性进行了研究。研究表明：机械补偿能自动实现堆芯的有效控制,并使功率峰因子保持在较低水平,是一种灵活、有效、经济的运行控制策略。此外,针对机械补偿运行对核电厂设计的影响进行了初步分析。     

177燃料组件堆芯反应堆机械补偿控制策略研究

177燃料组件堆芯反应堆通常采用G模式运行,负荷跟踪期间需要调整堆芯硼浓度。受硼回收系统能力限制,仅在85%寿期内具备负荷跟踪能力。为改善177燃料组件堆芯反应堆负荷跟踪能力,扩大可进行负荷跟踪的寿期范围,基于177燃料组件堆芯进行了机械补偿控制策略的研究。设计了不同控制棒组布置方案,从控制棒组价值、对功率峰的影响、负荷跟踪过程中控制能力等方面进行了分析。基于优化的控制棒组布置方案和机械补偿控制策略,进行了全寿期基负荷运行、90%寿期末日负荷循环负荷跟踪以及启动过程模拟。结果表明,在适当的控制棒组布置方案下,177燃料组件堆芯可实施机械补偿控制策略,负荷跟踪能力达到了国际先进水平。      

MSHIM运行模式在M310机组的初步应用研究

以大亚湾核电站1号机组为研究对象,尝试将机械补偿控制策略(MSHIM)运行模式应用于M310核电厂。分析表明,M310核电厂具有基负荷的MSHIM运行能力,具备一定的不调硼负荷跟踪能力,但G1、G2、G3棒组和R棒组存在控制能力不足的问题。在现有控制棒数量及布置前提下,通过重新分组并定义控制棒组,有可能在M310机组上实现MSHIM运行与控制策略。     

利用Pu—238添加燃料代替可溶硼控制反应性的运行策略

在无可溶硼（SBF）压水堆堆芯中,引入了用Pu-238添加燃料控制反应性和功率分布的新概念。在SBF堆芯中,尽管不可避免地广泛使用可燃毒物和控制棒控制反应性,但是在堆芯功率分布控制方面存在着相当大的难度。因此,实际的SBF运行还要等待很长时间。在此项工作中,已经确认,通过引入Pu-238添加燃料可以大大地抑制剩余反应性。与从早期的600MWe SBF堆芯设计工作所获得的结果相比较,使用Pu-238添加燃料的600MWe无可溶硼压水堆（SBF PWR）堆芯概念设计使堆芯反应得到了很好控制。尤其是,借助于本研究研制的简单的轴向区域图,利用Pu-238的浓缩区域成功地进行了轴向功率形状的控制。并且,在1300MWe SBF PWR堆芯设计中也试用了Pu-238添加燃料,在这种堆型中,如果可溶硼控制不可用,与较小型反应堆比较起来,其功率分布控制要更困难。结果表明,即使在大型压水堆中,不使用可溶硼也可以成功地控制堆芯反应性和功率分布。因此,在SBF堆芯设计中,一个控制问题难点,可以通过引入新燃料概念得以很好地缓解。进一步预计本研究引入的Pu-238添加燃料、简单的轴向区域图和控制运行策略将直接用于 实际的SBF堆芯设计。     

压水堆负荷跟踪运行的新模式

综述了压水堆负荷跟踪运行的控制方式，并指出了其优缺点，对压水堆负荷跟踪运行的新模式－ＭｏｄｅＫ的控制棒分组在堆芯中的分布，以及控制棒提棒过程进行了分析，认为其基本控制原理正是大系统控制理论中化整为零，分别对待思想的具体应用，从而提高了自动化程度。     

Improvement of core control strategy for CPR1000: Load follow without boron adjustment

This paper presents the study of load follow operations without boron adjustment for CPR1000. To enable the CPR1000 to perform load follow maneuvers without changing soluble boron concentration, the worth of Rod Cluster Control Assemblies (RCCAs) are reconfigured with their amount and location unchanged according to the reactivity variations during the load follow transient. To ensure the real-time ability of the reactor core, the target axial offset (AO) during load follow operations is set to the same value with that in based load, and the Delta-I is maintained within the special trapezoidal shaped target band around its target value.For the simulation of the reactor core, the time-dependent one-dimensional two group diffusion equations with the reactivity feedback of moderator temperature, Doppler and xenon–iodine are used. The transverse buckling is adjusted at each time interval so that the one-dimensional model can match the average characteristics of the three-dimensional reactor core accurately. To show the superiority of the improved core control strategy for CPR1000 reactor, the load follow results employing the purposed boron-adjustment free control strategy are compared to those obtained with the typical MODE-G control strategy. It has been demonstrated by the simulation results that the load follow capability of CPR1000 reactor is greatly improved due to the elimination of boron concentration adjustments during load change transients. Full load follow capability of the reactor has been extended from the initial 80% of cycle life to more than 90% of cycle life. Thus, the boron adjustment free improvement on the MODE-G core control strategy is feasible for CPR1000, which can improve the economical performance of the plant and simplify the operational process during power maneuvers.     

Axial offset control of PWR nuclear reactor core using intelligent techniques

Improved load following capability is one of the main technical performances of advanced PWR (APWR). Controlling the nuclear reactor core during load following operation encounters some difficulties. These difficulties mainly arise from nuclear reactor core limitations in local power peaking, while the core is subject to large and sharp variation of local power density during transients. Axial offset (AO) is the parameter usually used to represent of core power peaking, in form of a practical parameter. This paper, proposes a new intelligent approach to AO control of PWR nuclear reactors core during load following operation. This method uses a neural network model of the core to predict the dynamic behavior of the core and a fuzzy critic based on the operator knowledge and experience for the purpose of decision-making during load following operations. Simulation results show that this method can use optimum control rod groups maneuver with variable overlapping and may improve the reactor load following capability.     

Small PWR Using Coated Particle Fuel of Thorium and Plutonium

An innovative concept of PFPWR50 for district heating has been studied, which is a small PWR of 50MWt capability using coated particle fuels with conventional zircaloy cladding. This concept takes advantages of fuel integrity against fission products release of coated particle fuels and a high reliability of PWR technology based on the long history of a successful operation. We have investigated burnup characteristics of fuel rods, assemblies, and reactor cores by the calculation code SRAC95 in order to establish a core concept of long life without on-site refueling. The loading pattern of assemblies with various concentrations of burnable poison is optimized to obtain a flat excess reactivity during the core life in order to eliminate a soluble boron control system. The core life of a cycle is about 8.9 equivalent full power years. And we have also studied the applicability of SiC/SiC composite cladding in place of zircaloy cladding, which is now under development for gas cooled fast reactor fuels. It could be applicable to high burnup fuel rods for a long term operation. From the calculation results, it is found out that the burnup characteristics do not change significantly with SiC cladding and contribute to elongate the core life to 9.2 equivalent full power years.     

Advanced operational strategy for the IRIS reactor: Load follow through mechanical shim (MSHIM)

The renaissance of nuclear power brings more attention to advanced reactor designs and their improved performance and flexibility, including their enhanced load follow capability. Reactor control strategy used to perform transients including power changes has impact on the overall control system design. In particular, as the power change is performed within a load follow maneuver, several modifications occur in the core from a neutronic view point: the fuel and moderator temperature change, the xenon concentration and distribution are modified, the power distribution skewed axially, etc. These changes need to be adequately counterbalanced to keep both the core critical and the power distribution acceptable. The traditional approach in PWRs is to compensate for the reactivity change due to the power variation by adjusting the soluble boron concentration and moving a limited number of control rod banks. However, advanced reactors may adopt a different strategy for a variety of reasons. For example, water-cooled reactors that do not use soluble boron in coolant obviously cannot use its adjustment for this purpose. Moreover, Integral Primary System Reactors (IPSRs) using soluble boron, due to their integral design, have a large inventory of primary coolant. Therefore dilution/boration strategy, while in principle an option, becomes expensive for short time changes and leads to large volume of liquid effluent, in particular toward the end of cycle. Therefore, a capability to perform load follow without changing soluble boron concentration is very desirable for a range of reactor designs.International Reactor Innovative and Secure (IRIS) is an advanced medium-size IPSR that has been selected as the reference reactor for the purpose of this study. A capability to perform load follow maneuvers without changing soluble boron concentration has been examined and demonstrated through implementation of the Westinghouse Mechanical Shim (MSHIM) control strategy. A control bank design suited for MSHIM operation has been devised. Nine load follow scenarios covering a wide range of possible operating requirements, including Westinghouse design basis plus others proposed by EPRI for Advanced LWRs, have been successfully performed through the control rod banks movement only, without soluble boron adjustment, and maintaining power peaking factors within the acceptable range. Thus, IRIS provides improved operation by enabling load follow through MSHIM.     

Implementation of control rod movement and boron injection options by using control variables in RELAP5/PARCS V2.7 coupled code

In present neutron kinetics codes, control rods banks do not have the possibility of dynamic movement during the simulation of a transient; besides it is necessary to send the boron concentration from the thermal-hydraulic code to the neutronic code to account for changes in cross-sections due to boron dilution. For instance, control rod movements are pre-programmed with simple instructions introduced before the beginning of the calculation. Hence, control rod positions are not related to the core characteristics and the control systems at any time of the simulation. This work presents the changes introduced in RELAP5/PARCS v2.7 codes to achieve that control rods and the boron injection become more dynamic and realistic components in such kind of simulators. Furthermore, in order to test the modifications introduced in both codes, it has been analyzed a boron injection transient in a typical PWR Nuclear Power Plant. The thermal-hydraulic model includes all the primary loop components of a PWR, the core fuel assemblies modeled with PIPE components, pumps, steam generators, pressurizer, etc. The neutronic representation of the reactor has been made in a one-to-one basis fuel channel model for the whole core.     

压水堆堆芯设计特点及其演变

综合论述了压水堆堆芯设计中的化学补偿反应性、标准化无盒大型燃料组件、棒束型控制棒、可燃毒物和采用多区堆芯装料等基本问题。并以上述５大问题为基础,简要叙述了负荷跟踪运行给堆芯设计带来的有关设计问题。此外,简要介绍了当前压水堆堆芯的改进设计及演变过程。     

Implementation of control rod movement and boron injection options by using control variables in RELAP5/PARCS V2.7 coupled code

In present neutron kinetics codes, control rods banks do not have the possibility of dynamic movement during the simulation of a transient; besides it is necessary to send the boron concentration from the thermal-hydraulic code to the neutronic code to account for changes in cross-sections due to boron dilution. For instance, control rod movements are pre-programmed with simple instructions introduced before the beginning of the calculation. Hence, control rod positions are not related to the core characteristics and the control systems at any time of the simulation. This work presents the changes introduced in RELAP5/PARCS v2.7 codes to achieve that control rods and the boron injection become more dynamic and realistic components in such kind of simulators. Furthermore, in order to test the modifications introduced in both codes, it has been analyzed a boron injection transient in a typical PWR Nuclear Power Plant. The thermal-hydraulic model includes all the primary loop components of a PWR, the core fuel assemblies modeled with PIPE components, pumps, steam generators, pressurizer, etc. The neutronic representation of the reactor has been made in a one-to-one basis fuel channel model for the whole core.     

Application of Three Axial Offsets Trajectory Method for Load Follow Operation Control in PWRs

During load follow operation of PWRs, it is required to control the core power distribution and to reduce the amount of cost due to the usage of control devices such as boron adjustment and control rod. Since occurrence of xenon spatial distribution oscillation following the change of the reactor power can cause oscillation in the power distribution, one task in the core power distribution control is to suppress xenon oscillation as effectively as possible. A lot of studies have been done to solve the problem, some of which use complex mathematical treatments. On the other hand, the three axial offsets trajectory method, which uses a simple mathematical treatment based on two points reactor model, has been proved to be effective for xenon oscillation control. In this study, we examined the feasibility of application of the three axial offsets trajectory method in the load follow operation by comparing with conventional strategies such as boron priority control and control rods priority control. In order to increase the effectiveness of control means, we propose a new method that is constructed by considering the superiority of each control strategy.