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.     

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。     

用灰色系统理论预测环形通道内液钠的临界热流密度

用灰色系统理论对在两台液钠沸腾实验回路上测得的实验数据进行了钠沸腾临界热流密度（CHF）值影响因素的灰色相关分析,并用GM（1,1）模型对CHF进行预测,选用GM（1,h)模型对CHF进行了建模,计算及预测结果与实验值符合较好。     

快堆中钠环境下熔融物传播模型的建立与验证

快堆在超设计基准事故下运行,会导致燃料元件的熔化,并很快形成燃料和钢的混合沸腾池,池壁破损之后,高温熔融物将向组件盒间隙以及相邻组件中传播,使事故进一步升级.为了弄清熔融物传播的机理,采用了机理建模的方法,分别建立了导热冻结和整体冻结两种机理模型,在英国的SES实验及法国的PV-A实验上进行了验证,得出了有关冻结机理和钠流动的相关结论,通过对传播的相关参数进行比较分析.得出了熔融物传播的相关结论.     

环形通道内液钠的临界热流密度的灰色相关分析和灰色模型的建立(英文)

采用灰色系统理论,对在两台液钠沸腾实验回路上测得的实验数据,进行了影响钠沸腾临界热流密度(CHF)值因素的灰色相关分析,并用GM(1,1)模型对CHF进行了预测。选用GM(1,h)模型对CHF进行了建模。计算及预测结果与实验值符合较好。     

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。     

液态金属钠沸腾两相流动传热特性理论研究

在实验的基础上对液态金属钠沸腾两相流动传热特性进行理论研究。计算对象为环形流道。单相流动区域认为液态金属钠不可压缩;两相流动区域考虑钠蒸汽的可压缩性。两相流动区域选用均匀流模型,求解过程中采用迎风格式进行积分。将模型计算结果与相关实验数据进行对比,结果表明本文模型可用于计算液态金属钠沸腾两相流动传热特性,模型计算结果在一定程度上能完成对实验工作的拓展。     

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。     

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.     

利用子通道程序对快堆堆芯堵流的计算

沸腾是钠冷快堆安全分析工作中必须要考虑的一个问题，而堵流是众多导致发生沸腾原因中十分重要的一个。要对堵流导致的沸腾问题进行计算分析，就必须要发展堵流计算能力。文中用具有两相流计算能力的子通道程序对堵流问题进行了计算，和中国原子能研究院的试验结果以及其它程序计算结果的比较表明，对传统的子通道模型进行改进后程序具有良好的准确性，可以应用于今后的堵流沸腾计算。     

Analysis of low-heat-flux sodium boiling test in a 37-pin bundle by the two-fluid model computer code SABENA

This paper summarizes the development of a new detailed multi-dimensional multi-field computer code SABENA and its application to an out-of-pile low-heat-flux sodium boiling test in a 37-pin bundle. The semi-implicit numerical method employed in the two-fluid six-equation two-phase flow model has proved in solving a wide spectrum of sodium boiling transients in a rod bundle under low pressure conditions. The code is capable of predicting the spatial incoherency of the boiling, dryout on fuel cladding surfaces and fuel pin heat transfer. Essential to the successful application of such a mechanistic model computer code are validational efforts aimed at the LMFBR accident phenomenology analyses. Through the simulation of the natural circulation boiling conditions, this study provides a consistent analytical interpretation of the experimental data. The important influences of such parameters as the inlet flow restriction and bundle geometry have been examined through interpretations of two-phase flow analysis including considerations of the flow instability induced dryout mechanism.     

快堆局部事故时受损燃料组件的径向传热机理

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系列堆内实验建立了模型，针对SCARABEEBE+1实验的计算结果与实测数据吻合，进一步使用该模型对实际快堆中的单组件完全堵流事故进行了预测。结果表明：1）对实际快堆中发生全堵的燃料组件而言，其外部的冷却条件与SCARABEEBE+系列实验非常相似；2）堵流组件向上和向下的传热可忽略不计，径向传热对事故有较强的延缓作用；3）随着时间推进，径向传热的主导机理依次为液态钠单相对流、钠蒸汽在组件盒内壁冷凝、体热源沸腾池散热。     

TRACE simulation of sodium boiling in pin bundle experiments under loss-of-flow conditions

In the framework of PSI's FAST code system, the thermal–hydraulic code TRACE is being extended for representation of sodium two-phase flow. As the currently available version (v.5) is limited to the simulation of only single-phase sodium flow, its applicability range is not enough to study the behavior of a Generation IV sodium-cooled fast reactor (SFR) during transients in which boiling is anticipated. The work reported here concerns the extension of the non-homogeneous, non-equilibrium two-fluid models, which are available in TRACE for steam-water, to sodium two-phase flow simulation. The conventional correlations for ordinary gas–liquid flows are used as basis, with optional correlations specific to liquid metal where necessary. A number of new models for representation of the constitutive equations specific to sodium, with a particular emphasis on the interfacial transfer mechanisms, have been implemented and compared with the original closure models.A first assessment of the extended TRACE version has been carried out, by using the code to model experiments that simulate a loss-of-flow (LOF) accident in a SFR. One- and two-dimensional representations of the test section have been considered. Comparison of the 1D model predictions, with both experiment and SIMMER-III code predictions, confirm the ability of the extended TRACE code to predict the principal sodium boiling phenomena. Two-dimensional representation of the test section, however, has been found necessary for providing more detailed comparisons with the experimental data and thereby studying, in greater detail, the influence of the physical models on the calculated results.The paper thus presents a first-of-its-kind application of TRACE to two-phase sodium flow. It shows the capability of the extended code to predict sodium boiling onset, flow regimes, pressure evolution, dryout, etc. Although the numerical results are in good agreement with the experimental data, the physical models should be further improved. Other integral experiments are planned to be simulated, in order to further develop and validate the two-phase sodium flow modeling.     

Retention factors for fission products from sodium tests to simulate a severe LMFBR accident

In case of an LMFBR whole-core accident, fuel and fission products may be released instantaneously from the HCDA bubble through vessel head leaks or delayed from hot or boiling sodium pools after vessel melt-through. It is necessary to investigate the radiological source terms for both scenarios. In the paper we present and discuss retention factors for simulated fission products and fuel from FAUST under-sodium rupture disk tests on the instantaneous source term and from NALA tests with hot and boiling sodium pools on the delayed source term.     

RELAP5再淹没临界后传热模型不确定性研究

在大破口失水事故最佳估算加不确定性分析中,再淹没临界后传热模型的不确定性评价研究十分关键。本文针对RELAP5再淹没临界后传热模型展开研究,选取Weisman在0.1~0.4 MPa条件下进行的圆管过渡沸腾试验数据对再淹没过渡沸腾关系式进行评价,选取爱达荷国家实验室（INEL）低压下的圆管膜态沸腾试验数据对再淹没膜态沸腾关系式进行评价,给出其概率分布类型和范围,为进行大破口失水事故不确定性分析打下基础。     

Results of acoustic and temperature noise measurements on sodium cooled electrically heated 7 rod bundle

Assembly cooling deficiency in a LMFBR is one of the most important safety problems for reactor design and operation.

Studies on early detection and diagnosis of local accident by means of noise analysis techniques have been initiated at CNEN. Acoustic and temperature noise measurements have been carried out on a 7 rod bundle during slow power transients up to boiling conditions. The test section, simulating the italian PEC reactor fuel element, was mounted on ENA-2 sodium loop located at the CSN Casaccia.

Acoustic noise spectral analysis up to 32 kHz shows the appearance, in presence of boiling, of power increase at certain frequencies. Power spectra and rms values are updated and recorded every 0.3 sec and show large variations going from single phase to boiling.

Temperature noise spectral analysis shows that the power, between 1 and 50 Hz, increases, in presence of boiling, by a factor bigger than 30. It has been tested the sensitivity of other indicators of the temperature fluctuations, like skewness and flatness, to reveal boiling.     

Sabena: Subassembly boiling evolution numerical analysis

This paper describes the computer code SABENA that has been used in subassembly sodium boiling evolution numerical analysis as a contribution to fast breeder reactor safety analysis. SABENA is a two-fluid model subchannel code system to calculate coolant boiling and two-phase flow in a rod bundle together with external loop characteristics which affects the overall boiling behavior in the bundle section. With the use of relatively simple but reasonable constitutive models, the SABENA code has been applied to and validated against many multi-pin sodium boiling problems. The results have shown excellent agreement with the experiments. The numerical methods and models employed in the code have proven to be robust and efficient in light of the extreme severity of the conditions characterizing low-pressure sodium boiling.     

Investigations on Sodium Boiling Accidents in a Large Fast Breeder Reactor, (I)

The radial propagation of sodium boiling by thermal process in a 1,500 MWe FBR is analyzed with the BOIP-T code. Since the model used embodies some uncertainty, several modes of molten fuel behavior are considered. The time required for boiling to propagate from one channel to its neighbor is calculated. Even on the most conservative basis, the boiling propagation requires at least 10 sec, which indicates that the radial propagation of sodium boiling could be prevented by a suitable core protection system.