CNP650压水堆不调硼负荷跟踪可行性研究

海南昌江核电厂等CNP650压水堆采用Mode-A控制模式,该模式采用黑体控制棒,有很好的基负荷运行能力,但负荷跟踪能力相对较差。而对一些具有小电网的国家或地区,负荷跟踪运行能力具有一定的市场需求。不调硼负荷跟踪通过棒控系统自动完成,大大减轻了操纵员负担;负荷跟踪过程基本不需要频繁地调硼操作,允许简化化学和容积控制系统设计,减少了废液处理成本。为此,在CNP650压水堆上进行了不调硼负荷跟踪研究。负荷跟踪过程主要有两个控制任务:一是反应性补偿;二是功率分布控制。根据不调硼负荷跟踪的控制任务,重新进行了控制棒的设计、分组和布置,设置两套独立控制的控制棒组(功率补偿棒组和轴向偏移控制棒组),分别用于堆芯反应性控制和轴向功率分布控制,以实现不调硼负荷跟踪。使用SCIENCE程序包进行典型的12h~3h~6h~3h、100%—50%—100%功率水平的日负荷循环计算来进行不调硼负荷跟踪分析。计算步骤为:进行三维堆芯模型计算;根据三维堆芯模型建立一维堆芯模型;在一维模型基础上,进行模拟计算。完成了海南昌江核电厂平衡循环寿期末典型的日负荷循环不调硼运行分析,模拟计算结果表明在CNP650压水堆上不调硼负荷跟踪运行模式是可行的。     

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.     

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。      

Evaluation of Reactivity and Xe Behavior during Daily Load Following Operation

A boiling water reactor (BWR) has an excellent load following capability provided by a core flow control, which is used for changing a reactor power level and for compensating the subsequent Xe concentration change.

The core characteristics during load following operations are investigated in detail, using our reactor core simulator. Comparisons of changes of the Doppler reactivity, the void reactivity and the Xe reactivity during transients are performed. Also the features of Xe transient during load following operations are shown.

It has been shown that the core flow change required to compensate the Xe reactivity change produces much greater change of the void reactivity than that required for power level changes, and that the resulting local power change in the lower part of the core is greater than that in the upper part, because the Xe concentration change in the lower part is hardly compensated by the core flow control. Also the effects of power level changes, cycle patterns, and initial concentration of Xe and I on the Xe transient behavior have been investigated.     

Development of a boron concentration prediction model using multi-cell simulation of the automatic load follow operation

Load follow operations have been performed by manually changing the boron concentration in the reactor core and moving Control Element Assemblies (CEAs) for controlling the power and the power distribution. The manual operation of load follow requires experience and predictions related to core behavior following power changes because CEA movements distort the power distribution and a boron concentration control is also inaccurate and difficult due to the long time delay in boration/dilution operations. A boron concentration prediction model, accurately predict boron concentration in the reactor coolant system, including the chemical and volume control system (CVCS) following boration or dilution, was developed in order to enhance the boron concentration control during load follow operation. The model was developed using a multi-cell concept and integrated with the KOPEC Integrated Systems Performance Analysis Code (KISPAC), which is a system code used for design purposes. Boron concentration behavior was analyzed to verify the model for both direct and indirect injection using SKN 3&4 data. The load follow operation was simulated and the results were compared with the measured data obtained during the startup period. The developed model accurately predicted boron concentration behavior for all subsystems in the reactor control system and CVCS.     

Core physics analysis of 100% MOX core in IRIS

International Reactor Innovative and Secure (IRIS) is an advanced small-to-medium-size (1000 MWt) Pressurized Water Reactor (PWR), targeting deployment around 2015. Its reference core design is based on the current Westinghouse UO₂ fuel with less than 5% ²³⁵U, and the analysis has been previously completed confirming good performance for that case. The full MOX fuel core is currently under evaluation as one of the alternatives for the second wave of IRIS reactors. A full 3-D neutronic analysis has been performed to examine main core performance and safety parameters, such as critical boron concentration, peaking factors, discharge burnup, reactivity coefficients, shut-down margin, etc. In addition, the basis to perform load follow maneuvers via the Westinghouse innovative strategy MSHIM has been established. The enhanced moderation of the IRIS fuel lattice facilitates MOX core design, and all the obtained results are within the operational and safety limits considered thus confirming viability of this option from the reactor physics standpoint.     

Extension of Load Follow Capability of a PWR Reactor by Optimal Control

The problem of extending that part of the fuel life cycle during which a reactor is capable of sustaining load-follow operation is formulated as an optimal control problem. A two-node model representation of pressurized water reactor dynamics is used, leading to a set of non-linear ordinary differential equations. Differential Dynamic Programming is used to solve directly the resulting nonlinear optimization problem and obtain the trajectories of soluble boron concentration and control rod insertion. Results of computations performed for a reference reactor are presented, showing how the optimal control policy stretches the capability of the reactor to follow an average daily load curve towards the end of the fuel life cycle.     

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.     

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。     

A fast-running core prediction model based on neural networks for load-following operations in a soluble boron-free reactor

A fast prediction model for load-following operations in a soluble boron-free reactor has been proposed, which can predict the core status when three or more control rod groups are moved at a time. This prediction model consists of two multilayer feedforward neural network models to retrieve the axial offset and the reactivity, and compensation models to compensate for the reactivity and axial offset arising from the xenon transient. The neural network training data were generated by taking various overlaps among the control rod groups into consideration for training the neural network models, and the accuracy of the constructed neural network models was verified. Validation results of predicting load following operations for a soluble boron-free reactor show that this model has a good capability to predict the positions of the control rods for sustaining the criticality of a core during load-following operations to ensure that the tolerable axial offset band is not exceeded and it can provide enough corresponding time for the operators to take the necessary actions to prevent a deviation from the tolerable operating band.     

Testing of improved algorithms controlling energy release in VVÉR-1000 in maneuvering regimes at the Tianwan nuclear power plant (China)

The salient features of improved algorithms adopted at the Tianwan nuclear power plant for controlling the energy released in a VVéR-1000 core are examined. The optimal configuration of the controlling groups is chosen in the first two power-generating units, the reactor power can be varied automatically under the control of an automatic power regulator, the boron regulation system makes it possible to determine the first-loop makeup automatically, and a modern version of the Imitator Reactora program has been installed. The results of testing the algorithms in the No. 1 unit in regimes with single and cyclic (daily) power maneuvers are presented. The tests of single power maneuvers were combined with dynamic tests of equipment. The operation of the power-generating unit in a daily load schedule was tested separately. Five daily load-change cycles were conducted during these tests. __________ Translated from Atomnaya énergiya, Vol. 103, No. 5, pp. 277–282.     

AP1000机组首次临界试验验证计算

首次临界试验是压水堆核电厂调试启动过程的关键环节,旨在确认核反应堆堆芯能按照设计要求达到预期的临界运行状态。本文利用西安交通大学自主研发的NECP-Bamboo程序系统对AP1000机组堆芯的首次临界试验的设计结果进行了验证计算,并与AP1000堆芯的核设计结果进行了比较。计算结果表明：预估临界状态下的硼浓度的偏差为-15 ppm,控制棒积分价值的最大偏差为-52 pcm,硼微分价值的偏差不超过0.2 pcm/ppm,反应性温度系数的偏差不超过1 pcm/K。本文计算结果的精度与高保真计算程序KENO（概率论方法）和VERA（确定论方法）的计算精度相当,为确保AP1000堆芯调试启动阶段的核安全提供了进一步的数据支撑。     

Integrated modular water reactor (IMR) design

Integrated modular water reactor (IMR) has been developed as one of the advanced small-scale light water reactors, with a thermal output of 1000 MW. The IMR adopts natural circulation and self-pressurization in the primary cooling system, and a reactor vessel built-in steam generators. The core design has been performed using the current light water reactor technology. Thermal-hydraulic sensitivity analyses have been done from the viewpoint of the departure from nucleate boiling (DNB) limitation. The IMR core, with 97 21×21-type fuel assemblies and natural circulation in the primary coolant system, shows a good nuclear and thermal-hydraulic behavior and good allowable margins for the DNB phenomenon. The reactivity change with burnup is about 1%Δk by using burnable absorbers, and only 12 rod-cluster-controls are used through the operating cycle. The 20 m-height reactor vessel encloses steam generators in vapor and liquid portions. Plant dynamic analyses have been also performed in order to evaluate the IMR behavior from the viewpoints of plant operation and control. This study shows that the IMR will operate with enough margins for the core safety and will be stably controlled for load demand changes expected during normal operations.     

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.     

反应堆机组停堆检修后重返满功率运行的堆芯控制策略研究

核电站压水堆(PWR)在寿期末长时间停堆后重返满功率的运行过程中,堆芯控制存在困难,常常不能够安全快速地进入满功率运行。本文通过对影响堆芯控制的各种因素和物理过程进行分析,提出了一套基于临界点状态选择,在升功率期间采用提升功率控制棒和硼浓度稀释相结合的堆芯控制策略,并进行了实际验证。对重返临界点的选择和影响因素的分析直接影响后期操作,是堆芯控制策略的基础。策略对相同状况下的PWR堆芯控制有一定的指导意义。     

小型模块化快堆中含铪控制棒的设计与分析

快堆一般采用以碳化硼（B4C）为吸收剂的控制棒进行反应性控制。小型模块化快堆中子泄漏率较大,增殖能力偏弱,单位燃耗反应性损失较大。模块化反应堆运行周期较长,且需要紧凑型堆芯设计,控制棒数量有限。因此,小型模块化快堆需要高10B富集度的B4C进行反应性控制。由于吸收剂燃耗深、功率密度高且导热能力受辐照削弱严重,B4C的安全使用寿命有限。本文通过对比硼化铪（HfB2）、氢化铪（HfH162）和传统B4C为吸收剂的控制棒的反应性价值、堆芯功率分布、堆芯反应性反馈系数、控制棒温度裕度与吸收剂燃耗深度,发现HfB2有更高的安全裕度和更长的安全使用寿命。HfH162控制棒略微改善了功率分布,但其高温氢气解离问题有待进一步研究。     

用硼中毒法进行脉冲堆控制棒效率和总后备反应性测量

本文介绍了在脉冲堆零功率物理实验中,利用硼中毒法测量反应性的原理和方法,给出了脉冲堆堆芯的硼微分价值、控制棒效率和总后备反应性的实验结果。还利用硼中毒法和脉冲中子源法配合进行脉冲堆控制棒之间反应性干涉效应的实验研究,取得了初步结果。     

压水堆负荷跟踪的模糊控制系统

模糊控制理论的发展促进了模糊控制器在压水堆负荷跟踪中的应用。用模糊逻辑控制器和常规ＰＩＤ（比例－积分－微分）控制器相结合，并以输出增闪随功率调整的策略，解决了反应堆负荷跟踪问题，本方法解决了时变非线性对象的闭环控制并克服了基于模型的控制方法的不足，仿真结果显示本文提出的方案不但具有优良的动态特性，而且具有很高的稳态精度，使负荷跟踪控制的自动化程度大为提高。     

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.     

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.