事故分析中钠沸腾模型的改进

选择单组件瞬间全堵事故作为分析对象,并选取法国SCARABEE系列实验中的BE+1实验进行模型验证,事故模拟中的钠沸腾模型运用两流体六方程模型,子通道的径向和轴向网格均采用交错网格法进行网格划分,模型求解中针对子通道横向速度处理不足的缺点,根据拉梅算子展开原理提出改进方案,并通过对BE+1实验的模拟,验证模型改进的合理性。     

钠冷快堆单个燃料组件冷却剂沸腾的数值模拟

在正常功率下快堆单个燃料组件的瞬间完全堵流可能会产生相当严重的后果 ,对其后续事故序列及其潜在的破坏能力进行预测是必要的。对模拟这种现象的SCARABEEBE +1实验在包壳流动之前的阶段进行了数值模拟。程序中采用了两流体、六方程模型来描述沸腾及两相流动 ,应用子通道方法来对基本方程进行离散化 ,以半隐数值方法进行了求解。计算结果与实验观测相吻合 ,这表明该程序可以比较准确地预测单个燃料组件在瞬间完全堵流之后 ,包壳流动之前的行为。     

钠沸腾噪声探测研究

文章叙述了钠沸腾噪声探测研究进展,建立了离线和在线均可进行的高频和低频信号采集和处理系统,引进、开展、改进和编制了信号处理、故障诊断、事故报警和自回归模型分析等软件包。应用这些硬软件对水和钠沸腾噪声进行了探测和分析。结果表明,沸腾噪声信号的自功率谱密度(APSD)的幅值明显大于沸腾时的值,用自回归模型判别因子分析,可实现钠沸腾在线实时诊断和监护。     

体热源沸腾池的建模及其验证

在事故保护系统和自动停堆系统失效的假设下,快堆中一大类事故可能会发展到熔融池和沸腾池阶段,此阶段的特征是:液态钢和液态燃料为池内主要成分,以燃料的裂变热为体热源,整个池子被附着在冷壁面上的UO2固化壳包裹,当其中钢的温度超过沸点时,便开始沸腾。建立了一个半经验模型来描述体热源沸腾池的行为。模型中,用漂移速度模型来预测空泡份额分布,用修正后的Greene关系式计算平均传热系数并在此基础上根据实验结果确定局部传热系数,用焓方法求解包裹沸腾池的固化壳的温度场及厚度。对SCARABEE BF2实验(单组分UO2沸腾池)及BE+2(UO2 钢混合沸腾池)进行了模拟计算,计算结果与实验结果基本吻合。     

环形通道内液钠沸腾机理实验研究

实验研究和理论分析了负压下液态金属钠在环形通道中垂直上升流动时的沸腾机理及特性,得到了各种工况下大量有关液钠沸腾的数据及曲线。着重分析讨论了液钠沸腾曲线特征及沸腾机理。实验参数如下:热流密度:12.77W/cm~2-78.73 W/cm~2;液钠流速:0.047 m/s-0.674 m/s;饱和压力:850 Pa-9870 Pa;进口过冷度:63.12℃-270.97℃。     

快堆沸腾池数学模型

单组件盒内的沸腾池是快堆燃料组件瞬时堵流事故发展的一个重要阶段,这个阶段之后将会导致熔融物向组件盒外的传播.为了了解沸腾池的内部机理,本文建立了单组分沸腾池机理模型:采用漂移速度模型预测池内空泡份额的分布,用焓方法求解包裹沸腾池的燃料固化壳的温度场及厚度.根据不同的流型,对沸腾池和壁面间的换热Greene关系式进行了一些修正.结果表明,沸腾池的形成是由于冷却剂的排热能力降低,而形成的内部产热量和外部排热量的不平衡而导致的;这个热量的不平衡量是产生气泡的根源.Greene经验关系式适用于没有产生气泡之前的熔融池,形成沸腾池之后,要根据不同的流型对其做相应的修正.     

钠冷快堆中熔融池模型的建立与验证

快堆在超设计基准事故下运行时,会导致钠沸腾和干涸,如果不能及时停堆,接着就会产生燃料元件的熔化坍塌,在组件盒下部形成熔融池.为了对熔融池给出合理的安全分析,采用机理建模的方法,建立了完整的熔融池模型,并在法国的SCARABEE系列实验中的BF1三种功率的实验上进行了验证,和实验吻合较好,通过和所验证过的GEYSER及BF2等实验模型进行比较,得出了有关熔融池机理的相关结论.通过排热和温升等相关数据的比较,对熔融池向外的排热形式给出了合理分析,并得出了相关结论.     

永磁式钠流量计的研制

研制了量程为5m~3/h和0.5m~3/h的两种永磁式钠流量计,介绍了它们的结构、分度特性与阻力特性的理论计算、标定试验及误差分析,给出了分度特性的解析表达式。     

快堆单个燃料组件完全堵流事故的建模及其验证

为了预测正常功率下快堆单个燃料组件入口完全堵流所导致的事故序列,根据SCARABEE-N系列实验建立了相关的计算模型.冷却剂的沸腾及其两相流动的描述采用两流体模型;包壳的流动、燃料的熔化及其塌陷采用类似SURFASS程序的简单方法处理.对于事故后期形成的UO2-钢混合沸腾池,采用一维半经验模型描述,即:用漂移速度模型来预测空泡份额分布;用修正后的Greene关系式计算沸腾池和壁面之间的传热系数;用焓方法(enthalpy method)求解包裹沸腾池的固化壳的温度场及厚度.为了验证本文建立的模型,对SCARABEE BE 1实验结果进行了校核计算,其结果与实验结果基本吻合.     

JEFF-3.1.1 Nuclear Data Validation for Sodium Fast Reactors

The JEFF-3.1.1 Nuclear Data Library is the latest version of the Joint Evaluated Fission and Fusion Library. We present the status of the validation of this library using the Monte Carlo Code TRIPOLI 4.5 and the deterministic code package ERANOS 2.2 for fast reactor calculations. For this purpose, we reanalyze a selected set of integral experiments performed in the MASURCA mock-up (CEA/CADARACHE), in the ZPPR mock-up at ANL (USA), and in the SUPERPHENIX Power Reactor. Furthermore, we also present the analysis of pure sample irradiation experiments PROFIL and PROFIL-2 performed in the PHENIX reactor, as this kind of experiment provides a direct feedback on nuclear capture data. We observe good performances of these calculation tools for criticality calculations and fuel inventory prediction. From this validation work, some required improvements on nuclear data are highlighted, as well as the need for new specific integral experiments. The main trends observed are the following:

—Reactivity of clean and fresh cores: the results obtained with JEFF-3.1.1 are consistent with those obtained using JEFF-3.1 within 80 pcm, but there is an overestimation of the calculated reactivity of all the experiments between 40 and 800 pcm depending on the spectrum hardness (Pu content) and fuel composition, the discrepancy being larger in hard spectra. Additional investigations are in progress to understand this behaviour.

—Integral capture cross sections (PROFIL and PROFIL-2):/C/E - 1/≤ 3%, except for ²⁴¹Pu (C/E ≈ 1:08), ²⁴²Pu (C/E ≈ 1:18), ²³⁷Np (C/E ≈ 0:92), ²⁴³Am (C/E ≈ 0:93), ²⁴⁴Cm (C/E ≈ 1:35); impact of the trends observed on the individual fission products put in PROFIL and PROFIL-2 is ≈ 4% on the part of Δ_ρburnup due to fission products.     

快堆组件盒壁破损机理模型的建立与验证

快堆内发生超设计基准事故后，故障组件盒会发展到沸腾池，事故下一步的传播取决于池壁破损。文章采用机理建模方法，对3种主要盒壁破损机理建立模型，并在法国SCARABEE堆内实验中的BE+3和PV-A实验以及堆外GEYSER实验上进行了模型验证，模型计算结果与实验结果吻合较好。根据模型计算结果，对PV-A实验的池壁破损给予了合理解释，总结出快堆池壁破损的相关结论，并对堆内发生燃料-冷却剂相互作用（FCI）的可能性进行分析，给出了相关结论。     

Calculation Model for Predicting Power Distribution in Fast Reactors

An electrolyzer model for the analysis of a hydrogen production system using a solid oxide electrolysis cell has been developed, and the effects of principal parameters have been estimated via sensitivity studies based on the developed model. The main parameters considered were current density, area-specific resistance, temperature, pressure, molar fraction, and flow rates in the inlet and outlet. A simple model is also estimated for a high-temperature hydrogen production system that integrates the solid oxide electrolysis cell with a very high temperature reactor.     

用于钠泵的螺管型超导磁体设计与实验

本文对钠金属电磁泵的螺管型超导磁体部分进行了详细设计和实验研究.磁体设计指标为水平室温孔长度600mm,直径160mm,中心场磁感应强度5T.对超导磁体所用的超导线进行了短样测试,当温度为4.2K、磁感应强度为5T时,临界电流为464A,大于工作电流.磁体经过绕制、真空压力浸渍工艺后进行10次失超锻炼,在失超电流97 A的电流下测试,磁体中心磁感应强度达到4.66T.若进一步降低励磁速率,磁体中心磁感应强度预计可以达到5T左右,可以满足钠金属电磁泵的要求.     

Optimization of Fuel Assembly Allocation for Boiling Water Reactors

Abstract

A direct search algorithm is applied to the optimization of fuel assembly allocation of BWR with particular consideration given to the nuclear model and the treatment of operating constraints. A simple expression is derived for evaluating the stuck rod margin, based on regression analysis of data obtained by three-dimensional full core analysis, and the expression is applied to optimization procedure.

The practical applicability of the method is confirmed through trial computations for the second and equilibrium cycles of a medium-sized commercial BWR, with an examination based on various initial guesses and objective functions for radial power peaking.     

Impact of Nuclear Data Uncertainties on Innovative Fast Reactors and Required Target Accuracies

It is widely accepted that the current status of neutronics calculations for fast reactor design is such that the present uncertainties on nuclear data should still be significantly reduced, in order to get the full benefit from advances in modeling and simulation. Only a parallel effort in advanced simulation, high-accuracy validation experiments, and nuclear data improvement will provide designers with more general and wellvalidated calculation tools to meet tight design target accuracies to further improve safety and economics. The present paper presents very recent results related to nuclear data uncertainty impact assessment and target accuracy requirements for advanced reactor systems.     

Analysis of Mixed Oxide Fuel Critical Experiments with Neutronics ANalysis Codes for Boiling Water Reactors

Critical experiments of UO₂ and full mixed oxide (MOX) fuel cores conducted at the Tank-type Critical Assembly (TCA) were aNalyzed using BWR design-purpose codes HINES and CERES with ENDF/B files and Monte Carlo fine analysis codes VMONT and MVP with the JENDL-3.2 library.

The averaged values of the multiplication factors calculated with HINES/CERES, VMONT and MVP agreed with those of experiments within 0.3%ΔAk. The values by the design-purpose codes showed a small difference of 0.1%Δk between UO₂ and MOX cores. Monte Carlo code results showed that the JENDL-3.2 library had a tendency to overestimate the multiplication factors of UO₂ cores by about 0.3%Δk compared with those values of MOX cores. The root mean square errors of calculated power distributions were less than 1% for HINES/CERES and VMONT.

These results showed that (1) the accuracy of these codes when applied to full MOX cores was almost the same as their accuracy for UO₂ cores, which confirmed the accuracy of present core design codes for full MOX cores; and (2) the accuracy of the 190-energy-group Monte Carlo calculation code VMONT was almost the same as that of the continuous-energy Monte Carlo calculation code MVP.