AP1000核电厂堆芯寿期延长研究

在AP1000核电厂寿期末,维持满功率运行所需的临界硼浓度已经达到约7×10-6。为实现寿期末核电厂满功率运行,必须采取堆芯寿期延长措施。在基准工况下通过控制汽轮机调节阀开度和降低反应堆冷却剂平均温度引入正反应性,可使核电厂满功率多运行17 d。此外,对慢化剂温度系数和高压给水加热器的关闭列数进行敏感性分析,结果表明,慢化剂温度系数越负,反应堆平均温度降温速率越小,堆芯预期寿期越长。在2种敏感性工况下核电厂寿期末分别可满功率多运行约12 d和54 d。     

压水堆核电厂机械补偿策略下控制系统AFD的校准及验证

压水堆核电厂机械补偿控制策略的棒阴影效应导致保护系统指示的轴向通量偏差（AFD）变为外围组件加权值而非堆芯平均值,因此基于堆芯平均AFD的控制系统无法像传统电站那样直接使用保护系统输出值。基于该问题,探索了控制系统堆芯平均AFD几种线性的指示途径,研究了不同控制棒位置及机组氙震荡瞬态对线性关系的影响,提出了一种“偏差修正”的堆芯平均AFD校准方法。经过机组调峰过程验证,结果表明使用该方法可以消除控制棒位置、机组功率变化及氙浓度变化对控制系统AFD指示的影响,能够满足系统指示精度的要求。因此这种方法可以用于机械补偿控制策略控制系统AFD的校准。      

反应堆启动和停闭的模拟实验

本文介绍了在北京核电厂模拟器上模拟不同堆芯寿期下的反应堆启动,并对一些非正常情况下的现象和问题进行了研究,从物理上检验了该模拟器的模拟性能.     

根据核电厂工况进行应急防护决策

介绍了美国核管会 (NRC)和国际原子能机构 (IAEA)有关在严重事故期间根据核电厂工况进行防护决策的方法 ,特别是制定核电厂应急行动水平、根据核电厂工况进行堆芯损坏评价、估计源项和确定防护行动的方法。为提高我国核电厂应急响应的有效性 ,提高在事故期间进行防护决策的科学性 ,建议我国应尽快掌握和研究制定核电厂应急行动水平、事故期间评价堆芯损坏和估计源项的方法学     

AP1000非能动堆芯冷却系统测试实验及评估

反应堆堆芯冷却系统是核电厂安全分析的重要内容,新设计的电厂必须通过实验验证其事故工况下保持堆芯覆盖和导出热量的可靠性。本文详细介绍了APl000核电厂非能动堆芯冷却系统的测试实验和美国核管会(NRC)的评估结果。     

可燃毒物提高小型压水堆堆芯寿期研究

针对长寿期堆芯的应用需求,开展了提高小型压水堆堆芯寿期研究。以棒状燃料为对象,对不同栅格尺寸和不同可燃毒物的选取进行计算,得出小型压水堆堆芯寿期相关影响因素。通过对不同尺寸的燃料栅格进行输运-燃耗计算,得到燃耗最佳栅格尺寸。以燃耗最佳栅格尺寸建立组件,并选择转换性能好的锕系核素~(240)PuO_2作为可燃毒物,利用~(240)Pu吸收中子转换成易裂变核素~(241)Pu的特性,对堆芯实现反应性控制和寿期延长。本研究通过对燃料栅格尺寸和可燃毒物的合理选择,提高了燃料利用率,达到延长堆芯寿期的目的。     

英国先进气冷堆存在严重安全问题

【英国《卫报》2006年7月5日报道】根据《信息自由法》于2006年7月5日公开的一些文件表明,英国政府核视察员已对国内日益老化的先进气冷堆(AGR)的安全性提出严重质疑,因为部分AGR反应堆的堆芯已出现重大裂缝。这些文件表明,核安全局(NSD)已经对欣克利角B核电厂及其他核电厂的反应堆堆芯退化问题发出警告。NSD还对英国能源公司(BE)进行了批评。BE目前运营着包括欣克利角B核电厂在内的7座AGR核电厂(共计14座反应堆)、1座压水堆核电厂以及1座燃煤电厂。根据文件显示,BE不了解反应堆堆芯的损坏程度,无法监控堆芯的恶化,也不十分明白出…     

压水堆中使用分立型铀、钍燃料组件的堆芯物理特性研究

通过对分立型铀、钍燃料组件 ,使用在秦山 30 0MW电功率压水堆核电厂中堆芯物理特性的探讨 ,寻找2 3 2 Th在PWR中可能利用的途径。为此 ,特采用铀、钍燃料组件分立的双进料系统的装卸料方法 ,其堆芯寿期分别为铀组件 3个循环 ;钍组件 1 0个循环。并以秦山核电厂为参考电厂 ,进行了 1 0个循环的燃耗计算 ,每一循环装料时均有 4个钍组件进堆。计算结果表明 :到第 1 0循环寿期末 ,堆芯中 40个钍组件所含的2 3 3 U总量已达到 2 1 2 6kg ,可直接参与堆芯的链式反应 ,从而达到利用2 3 2 Th的目的。并可同全铀组件堆芯比较中看出 ,分立型铀、钍组件混装堆芯每一循环 (第 1 0循环后 )可少装 2 0 0多kg2 3 5U ,这样就为钍 铀燃料循环展示了光明的前景。当然如果要达到实际应用 ,仍有许多工程技术问题亟待解决     

特征统计算法换料优化程序CSA针对实际工程设计要求的改进研究

为了将自主开发的特征统计算法(CSA)燃料管理优化程序用于实际核电厂堆芯换料设计,需要针对换料设计中的一些特殊工程要求进行相应的改进。本文以岭澳核电站堆芯为计算模型,针对这些工程要求对原有CSA程序进行了改进开发,并分别在无可燃毒物堆芯、有可燃毒物堆芯以及平衡循环堆芯换料设计问题上进行了测试验证。最终的结果证明,CSA程序经过相应的改进后,完全可以真正用于核电厂堆芯换料设计和优化。     

评AP1000堆芯设计

根据美国用户要求文件(URD)对3代压水堆核电厂的某些要求,比较AP600和AP1000核电厂的某些设计参数。建议三门核电厂和海阳核电厂取消机械补偿(MSHIM)基荷运行模式及复杂的堆芯设计。     

An experimental and analytical study of the effect of axial power profile on CHF

The effect of axial heat flux distribution (AFD) on the critical heat flux (CHF) was investigated. CHF measurements were obtained with HFC-134a cooled vertical tubes having four non-uniform and one uniform AFD profiles. The HFC-134a test conditions covered a pressure range from 1.6 to 2.4 MPa, a mass-flux range from 2.8 to 4.7 Mg m⁻² s⁻¹, and an inlet-quality range from −0.9 to 0. The water-equivalent pressure and mass-flux ranges are 10–14 MPa and 4–6.5 Mg m⁻² s⁻¹, respectively.In general, the observed AFD effect on critical power is small at high inlet subcoolings. At low inlet subcoolings, the critical power for the inlet-peak profile is up to 15% higher than that for the outlet-peak profile. A local conditions analysis showed that the AFD has the strongest effect on CHF at high dryout qualities. CHF values for non-uniform AFDs could be 50% lower than those for the uniform AFD. The AFD effect on CHF becomes diminished with decreasing dryout quality.Four different approaches to account for the effect of AFD on CHF were assessed against the experimental values from the current experiment. The boiling-length-average heat-flux approach with the boiling-length starting point at the onset of annular flow (OAF) provided the best prediction of the critical power and the CHF location.     

大亚湾核电站延伸运行及降功率期间反应堆轴向功率偏差的控制及其计算机模拟分析

针对压水堆核电站反应堆轴向功率偏差控制比较困难的情况．分析了引起轴向功率偏差的影响因素．特别是降功率对轴向功率偏差的影响。根据分析结果提出相应的控制反应堆轴向功率偏差的策略．以保证核电站功率瞬变运行满足核安全的需要。用计算机模拟了不同控制策略下降功率过程轴向功率偏差的变化．并与大亚湾核电站现场测量系统的测量值比较．证明不同控制策略下的计算机模拟分析结果是可信的。     

Power peak factor for protection systems – Experimental data for developing a correlation

This paper aims to construct a data set that can be used to train neural networks to furnish the power density peak factor during reactor operation. The inputs considered were those available in the reactor protection systems, namely, the axial and quadrant power differences obtained from measured ex-core detector signals, and the position of control rods. The response of ex-core detector signals was measured in experiments performed in the IPEN/MB-01 zero-power reactor. Several reactor states with different power density distribution were obtained by positioning the control rods in different configurations. The power distribution and its peak factor were calculated for each of these reactor states. The obtained results show that the power peak factor correlates well with the control rod position and the quadrant power difference, and with a lesser degree with the axial power differences. The data presented an inherent organisation and could be classified into different classes of power peak factor behaviour as a function of position of control rods, axial power difference and quadrant power difference. The analysis of the data set indicates that the power peak factor can be determined through a neural network having as input the position of control rods. Regarding only signals of ex-core detectors, the data indicate that a neural network may estimate better the power peak factor if the input vector comprises both the axial and the quadrant power differences.     

Reconstruction of core axial power shapes using the alternating conditional expectation algorithm

We have introduced the alternating conditional expectation (ACE) algorithm in reconstructing 20-node axial core power shapes from five-level in-core detector powers. The core design code, Reactor Operation and Control Simulation (ROCS), calculates 3-dimensional power distributions for various core states, and the reference core-averaged axial power shapes and corresponding simulated detector powers are utilized to synthesize the axial power shape. By using the ACE algorithm, the optimal relationship between a dependent variable, the plane power, and independent variables, five detector powers, is determined without any preprocessing. A total of ∼3490 data sets per each cycle of YongGwang Nuclear (YGN) power plant units 3 and 4 is used for the regression. Continuous analytic function corresponding to each optimal transformation is calculated by simple regression model. The reconstructed axial power shapes of ∼21,200 cases are compared to the original ROCS axial power shapes. Also, to test the validity and accuracy of the new method, its performance is compared with that of the Fourier fitting method (FFM), a typical method of the deterministic approach. For a total of 21,204 data cases, the averages of root mean square (rms) error, axial peak error (ΔF_z), and axial shape index error (ΔASI) of new method are calculated as 0.81%, 0.51% and 0.00204, while those of FFM are 2.29%, 2.37% and 0.00264, respectively. We also evaluated the wide range of axial power profiles from the xenon-oscillation. The results show that the newly developed method is far superior to FFM; average rms and axial peak error are just ∼35 and ∼20% of those of FFM, respectively. ©     

均匀加热全长棒束过冷沸腾工况子通道参数场计算分析

采用壁面热分配模型对PSBT基准题中的5×5均匀加热全长棒束过冷沸腾传热进行了数值模拟研究,分析了均匀加热全长棒束通道中不同子通道和加热元件表面参数沿轴向的发展过程和径向的分布特性。研究发现,角通道和边通道是弱对流区域,其质量流速低于棒束平均值,但由于冷棒功率偏低,消除了流动不均衡性对传热效果的影响。在棒束径向方向,不同位置子通道间参数场存在差异,这是由于位于搅混格架横向导流对角方向的通道具有更有效的通道间对流效果,其传热效果更好。这种流动特性引起的参数差异在角通道中尤为显著。热棒表面过热度明显高于冷棒过热度,且位于非搅混格架横向导流方向的热棒具有更高的壁面过热度。     

Power series solution method for solving point kinetics equations with lumped model temperature and feedback

Point kinetics equations are stiff differential equations, and their solution by the conventional explicit methods will give a stable consistent result only for very small time steps. Since the neutron lifetime in a LMFBR is very short, the point kinetics equations for LMFBRs become even stiffer. In this study the power series solution (PWS) method is applied for solving the point kinetics equations for a typical LMFBR. A Fortran program is developed for accident analysis of LMFBRs with the PWS method for solving the point kinetics and a lumped model for solving the heat transfer equations. A new technique is developed with fixing factor to find out the average temperature at the peak power node (PPN) without performing temperature calculations at all axial nodes in a reactor fuel pin. The temperature at PPN also decides whether the reactor is within the design safety limit (DSL) or it has entered a serious transient that may lead to an accident. The coupled heat transfer and point kinetics models for a peak power node give the average fuel, clad and coolant temperatures. For the transient over power accidents (TOPA), this is the best way for calculating the temperature, with minimum amount of computations. TOPA analyses are carried out with PWS method. It is found that the PWS methodology uses a small number of numerical operations, while the computational time and the accuracy are comparable with the available fast computational tools. This methodology can be used in nuclear reactor simulation studies and accident analysis.     

Long life small CANDLE-HTGRs with thorium

CANDLE (constant axial shape of neutron flux, nuclide densities and power shape during life of energy producing reactor) burnup strategy is applied to small (30 MWth) block-type high temperature gas-cooled reactors (HTGRs) with thorium fuel. The CANDLE burnup is adopted in this study since it has several promising merits such as simple and safe reactor operation, and the ease of designing a long life reactor core. Burnup performances of thorium fuel (²³³U, ²³²Th)O₂ are investigated for a range of enrichment ⩽15%. Discharged fuel burnup and burning region motion velocity are major parameters of its performances in this study. The reactors with thorium fuel show a better burnup performance in terms of higher discharged fuel burnup and slower burning region motion velocity (longer core lifetime) compared to the reactors with uranium fuel.     

The Potential for Advanced Limiter/Divertor Systems

The Advanced Limiter-divertor Plasrna-facing Systems (ALPS) program was initiated in FY 1998 in order to evaluate the potential for improved performance and lifetime for plasma-facing systems. The main goal of the program is; to demonstrate the advantages of advanced limiter/divertor systems over conventional systems in terms of power density capability, component lifetime, and power conversion efficiency, while providing for safe operation and minimizing impurity concerns for the plasma.     

Optimization of Axial Enrichment and Gadolinia Distributions for BWR Fuel under Control Rod Programming

The axial enrichment and gadolinia distributions of BWR (boiling water reactor) fuel are optimized under control rod programming. The objective of the problem is to minimize the average enrichment required to reach a planned EOC (end-of-cycle) with criticality condition and axial power peaking constraint.

A method of approximation programming is employed as the basis for the solution method. Resulting linear programming problem at each iteration step is solved by means of goal programming algorithm. The method is applied to the initial fuel for a typical BWR/5 represented by an axial one-dimensional core model

Two-region analysis leads to the conclusion that the core bottom should be depleted during the cycle so that the power shifts to the core top at EOC. The enrichment and gadolinia distributions are determined to maximize EOC power peaking within a limit. The optimal solution of a 24-region fuel with a power peaking limit of 1.4 saves 10.6% in uranium ore compared with a uniform fuel depleted with a Haling power shape. Half the saving comes from an optimal natural uranium blanket implementation.