华龙一号“177堆芯”特点分析

华龙一号（HPR1000）压水堆核电厂最显著的技术特征是反应堆采用由177个燃料组件构成的堆芯（简称“177堆芯”）,具有完全的自主知识产权。为深入分析其特点,本文介绍了“177堆芯”的主要技术特征,并在燃料组件及控制棒组件数目方面与157个燃料组件构成的堆芯（简称“157堆芯”）进行了对比分析;对2种典型反应堆堆芯（“177-A堆芯”与“177-B堆芯”）装载方案的异同进行了叙述和评价。结果表明,与“157堆芯”相比,“177堆芯”在安全性和经济性方面更有优势;2种典型堆芯的首循环装载布置各有所长,在可燃毒物选材上,“177-B堆芯”优于“177-A堆芯”。最后,从取消堆芯中央位置控制棒组件、设置堆芯径向金属反射层、实施无中子源启动、分批装载自主化燃料组件以及优化堆芯活性段长度等5个方面给出了HPR1000反应堆堆芯的优化建议。      

Calculation of the thermohydraulic parameters of a core comprised of unjacketed fuel assemblies

A computer program for calculating the thermohydraulic parameters of a core with jacketless fuel assemblies as a single mass of fuel elements is developed on the basis of the Kedr program for the channelwise computation. The Kedr-A program algorithm employs the principle of decomposition (partition) to the computed region of the core (1/12th part). The computational space is divided into a definite number of subregions – symmetry elements with repeatable geometric structure of the lattice of fuel elements and other structural components of the core. The thermohydraulic parameters of the cells in each section of the core are calculated iteratively over the symmetry elements of the jacket-less fuel assemblies of 1/12th part of the core of a nuclear reactor with water coolant. The symmetry elements are interrelated by the conditions at the boundaries connecting theses regions. The computational algorithm is checked by comparing with experimental data on the mixing of the coolant obtained on a technological stand consisting seven jacketless fuel assemblies.     

百万千瓦级压水堆核电站长燃耗堆芯钆可燃毒物优化研究

对百万千瓦级参考核电站长燃耗堆芯（１８个月换料）采用的可燃毒物（钆）含量与堆芯燃料管理主要结果进行了分析研究。该研究采用先进的燃料管理程序系统,对不同可燃毒物含量和不同可燃毒物棒根数的燃料组件进行了计算,给出了组件无限增殖因子（ｋｉｎｆ）随燃耗的变化关系,据此对参考堆芯采用相同的装载进行了４种方案燃料管理计算。计算结果表明,对于堆芯燃料管理,采用低可燃毒物含量、含可燃毒物棒数多的装载方案明显优于高可燃毒物含量、含可燃毒物棒少的堆芯装载方案。     

Analysis of accidents leading to local flow stagnation in parallel fuel channels of RBMK-type reactors

The RBMK-type nuclear power reactors, still operating in Russia, are graphite-moderated with vertical fuel channels, using low-enriched nuclear fuel. The main challenge, which leads to the overheating of the fuel assemblies, fuel channels and other core components in channel type nuclear reactors, is a misbalance between heat generation in core structures and heat sink, which can appear due to the loss of coolant accident. In this accidental case, the emergency core cooling system ensures the core cooling. In RBMK-type reactors this system consists of hydro-accumulators and a number of pumps, taking water from large water reservoirs. This equipment injects water into the fuel channels through the group distribution headers at high pressure. However, the direct supply of cold water from emergency core cooling system into fuel channels is possible only if check valves on group distribution headers are closed properly. If these check valves failed, the part of water would be lost through the break, the flow stagnation in channels could occur, which might lead to overheating of fuel assemblies in the fuel channels. This paper presents the results of deterministic safety analysis, performed using system thermal hydraulic code RELAP5. Using this code the reactor cooling system of RBMK-1500 was modelled and analyses of loss of coolant accidents with failure of few check valves in group distribution headers were performed. The results of the calculations are used for the development of symptom-based emergency operating procedures for RBMK-1500. The basic principles that describe the complex distribution of water flows in vertical forced circulation circuit with parallel fuel channels can be adjusted for the RBMK-1000 reactors, still operating in Russia.     

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。     

Use of low neutron leakage fuel loads to decrease the radiation load on a VVÉR-1000 vessel

It is shown on the basis of data obtained at Ukrainian nuclear power plants that fuel loads with low neutron leakage can be used effectively to decrease the radiation load on the reactor vessel. The characteristics of 104 fuel loads and the results of a determination of the radiation load on the vessel are analyzed to develop a criterion according to which a VVéR-1000 fuel load can be classified as a load with low neutron leakage. It is shown that the following condition can be chosen as such a criterion: the run-averaged relative power release in all protruding fuel assemblies must be less than 0.57. Different variants of the arrangement of the VVéR-1000 core are examined and analyzed. It is shown that placing burned-out fuel assemblies along the periphery of the core and decreasing the number of neutrons leaving the core do not always result in a lower neutron load on the reactor vessel. __________ Translated from Atomnaya énergiya, Vol. 101, No. 2, pp. 93–97, August, 2006.     

Optimization studies of fuel loading pattern for a typical Pressurized Water Reactor (PWR) using particle swarm method

A three dimensional multi-energy group computer model PRISHA, which solves the neutron diffusion equations using finite difference method is developed for Pressurized Water Reactor (PWR). This computer code can find an optimum loading of a group of fresh fuel assemblies along with fuel assemblies of different exposures. The successive line over relaxation (SLOR) method is used to solve neutron diffusion equations. After validation of this part of computer code against an IAEA – PWR benchmark problem with 177 fuel assemblies in the core, particle swarm optimization (PSO) method is incorporated in the code for finding the optimum fuel loading pattern. A typical PWR core with 157 fuel assemblies, where 289 fuel pins are arranged in 17 × 17 rectangular arrays in a fuel assembly, was analyzed using this computer model for two cycles using PSO method. Different numbers of particles and iterations were used in PSO method. The results are found to be not very sensitive to either the number of particles or the number of iterations used in PSO method for considered case. However, a number of experiments have to be performed to arrive at the best global fitness parameter. Reasonably low power peaking factors were obtained for both the cycles.     

超高通量快中子试验堆堆芯初步概念设计

针对先进核能系统发展需要,提出了超高通量堆的堆芯概念设计。本文采用板型燃料、正方形燃料组件设计,设置宽流道保证堆芯冷却剂占有较高的体积份额。堆芯采用52盒燃料组件,设置8盒控制棒组件和较厚的反射层。通过堆芯概念设计方案评价,结果表明堆芯循环长度可达100EFPD(等效满功率天),所提出的超高通量堆的最大中子注量率可达到1.08×10¹⁶ cm^-2·s^-1。     

Evaluating the effect of using different sets of enrichment for FAs on fuel management optimization using CA

In nuclear reactor core design, achieving the optimized arrangement of fuel assemblies (FAs) is the most important step towards satisfying safety and economic requirements. In most studies, nuclear fuel optimizations have been performed by using a finite number of different types of FAs. However the effect of FA numbers with different enrichments and the difference between their maximum and minimum enrichment values can be important and should be evaluated in the optimization process.     

Analysis of decay heat removal from RBMK-1500 reactor in decommissioning phase by natural circulation of water and air

The main problem in nuclear energy is providing of safety at all stages of lifetime of nuclear installations in conditions of normal operation, accidents and at shutdown. Ignalina NPP, located in Lithuania, is one of the latest with RBMK reactors at highest capacity. Ignalina NPP has two units, both are closed for decommissioning now (in 2004 and 2009). Both units are equipped with RBMK-1500 reactors, the thermal power output is 4200 MW, the electrical power capacity is 1500 MW for each. In RBMK-1500 reactor the fuel assemblies remain for long time inside reactor core after the final shutdown. The paper discusses possibility of heat removal from the RBMK-1500 core at shutdown condition by natural circulation of water (1) and air (2) inside the fuel channels. In first case the decay heat from fuel assemblies is removed due to natural circulation of water and the piping above reactor core should be cooled by means of ventilation in the drum separator compartments. To warrant free access of air in to fuel channels (in the second case) the reactor cooling system should be completely dry out and the pressure headers and the steam discharge valves in steam lines should be opened. If mentioned conditions will be fulfilled, the reactor core will be cooled by natural circulation of water or air and fuel rods remain intact.     

Tests of a Two-Stage Core at the KV-1 Test Facility: Physical Aspects and Results

Atomic Energy - The physical aspects and main results of reactor tests of a two-stage core consisting of fresh fuel assemblies and a significant number of fuel assemblies from the previous core,...     

含有可燃毒物的压水堆核电站堆芯装料设计优化研究

含可燃毒物的压水堆装料优化是燃料管理优化研究中的难点,应用通常的脱耦方法和优化算法效率低、全局性差。研究提出局部脱耦方法用以简化问题规模、缩小搜索空间,选择特征统计算法进行优化方案的搜索。利用局部脱耦方法结合特征统计算法研制出压水堆核电站堆芯LP和BP耦合装料优化程序CSALPBP。使用该程序对大亚湾第10循环和第12循环进行了装料优化计算。结果表明CSALPBP程序在求解含可燃毒物的压水堆装料优化问题方面具有很高的搜索效率和很好的全局性,能够较好地解决含可燃毒物的压水堆堆芯装料优化难题。     

径向倒料式驻波堆堆芯概念设计

本工作在综合分析日本CANDLE堆和美国TerraPower公司设计的TP-1堆的基础上,提出沿径向倒料的驻波堆堆芯设计初步方案,通过高、低富集度组件在堆芯的混合布置展平功率,降低了堆芯的功率峰因子和组件的最大燃耗深度。通过倒换料,实现了增殖 燃耗波的传递。为能有效地利用行波堆增殖产生的易裂变核素,采用更换包壳的新技术,实现核燃料的持续利用。     

六边形套管型燃料堆芯临界质量测量试验结果验证分析

为验证核设计程序对燃料组件、铍组件和铝组件的计算可靠性,对六边形套管型燃料堆芯（HCTFR）临界质量测量试验数据进行了验证计算和偏差分析。通过分析不同位置铝组件的反应性差异,提出了新的近活性区铝组件计算模型,将铝组件近活性区布置方案的计算偏差从2.2%降低至0.1%,为堆芯核设计程序的工程验证奠定了较好的基础。      

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。      

控制棒导向管外流道旁流特性实验研究

冷却剂流经核反应堆堆芯时，绝大部分通过燃料组件内部流过，带走裂变能量。另外一小部分作为旁流经过燃料组件外侧流道、控制棒导向管外侧及内侧流道流出。为确保反应堆在正常运行工况下的安全性，必须限制堆芯旁流流量。本文通过开展导向管外侧流道阻力特性实验研究，在不同流量工况下获得了分段压差，并进一步拟合了雷诺数与阻力系数的关系式。实验结果表明，导向管外侧流道压力损失主要集中在堆芯下栅格板处，当反应堆额定工况运行时，单组导向管外侧流量仅为0.196 m³/h。     

Optimal fuel utilization during decommissioning of nuclear power plants with RBMK reactors

When RBMK reactors are decommissioned successively in the same nuclear power plant, part of the fuel of the stopped reactors can be transferred to other units and additionally burned in continuing operations. The problem of minimizing the consumption of fresh fuel by optimal distribution of the additionally burned fuel over the reactors is examined. The limitations on the refueling rates, the holding time of fuel assemblies prior to transfer, the service life of fuel assemblies, and certain characteristics of reactors are taken into account. It is shown that the reuse of fuel in other units permits saving from one to almost two thousand fresh fuel assemblies and that the effect of optimizing the additional burn regime can reach several hundreds of saved fuel assemblies. __________ Translated from Atomnaya énergiya, Vol. 102, No. 5, pp. 284–290, May, 2007.     

Preliminary study of the tight lattice pressured heavy water reactor loaded with Pu／U and Th／U mixed fuels

To improve nuclear fuel utilization efficiency and prolong fuel cycle burn-up,a tight ptich lattice pressured heavy water reactor was investigated as an alternative of next generation of power reactors.It is shown that the high conversion ratio and negative coolant void reactivity coefficient are challenges in the reactor core physics designs.Various techniques were proposed to solve these problems.In this work.a tight pitch lattice and mixed fuel assemblies pressured heavy water reactor concept was investigated.BY utilizing numerical simulation technique,it is demonstrated that reactor core mixed with Pu/U and Th/U assemblies can achieve high conversion ratio(0.98) ,long burn-up(60GWD/t)and negative void reactivity coefficients.     

基于遗传算法的钠冷快堆堆芯流量分区优化设计方法

钠冷快堆采用封闭组件，流量分区是实现堆芯出口温度展平的重要途径。传统的流量分区优化设计方法的计算量随组件数的增加呈指数增长，不适用于解决大型问题。本文建立了流量分区设计的最优化模型，并设计了基于最优个体保存策略的遗传算法，以燃料最高温度限值和包壳温度限值为边界条件，搜索出使活性区平均出口温度最高以及活性区总流量最小的最优流量分区方案，为解决大型钠冷快堆堆芯流量分区优化设计问题提供了新的途径。