Boiling and dryout behind local blockages in sodium cooled rod bundles

Much attention has been given in LMFBR safety analysis to cooling disturbances caused by local blockages within a fuel subassembly. Such blockages are generally considered to be more probable in gridded fuel pin clusters which present the possibility for solid particles in the coolant to be trapped at grids to form a radially extending flow obstruction. The temperature distribution produced in the region of impaired cooling has been studied in water and sodium experiments in pin bundles of various sizes. The experimental work at KfK on local cooling disturbances culminated in two local blockage experiments in the KNS sodium loop simulating LMFBR fuel elements with a 49% central and a 21% corner blockage. In the frame of this work pin cooling in the wake of the blockage was investigated in single-phase conditions, in boiling conditions up to dryout and in conditions simulating gas release from failed pins. The general aims of the studies were to demonstrate that the consequences of a local blockage do not lead to rapid propagation of damage within a pin bundle and to obtain data for validation of theoretical models.     

Computational fluid dynamic investigations of partial blockage detection by core-temperature monitoring system of a sodium cooled fast reactor

Thermal hydraulic studies have been carried out to understand temperature dilution suffered by core-temperature monitoring system of a sodium cooled fast reactor. The three-dimensional computational model is validated against experimental results of a water model. Jet mixing phenomenon as predicted by different turbulence models is compared and RNG k–? model is found to be better than other models. A comprehensive parametric study considering: (i) effects of construction/manufacturing tolerances on thermocouple positions with respect to subassembly positions, (ii) thermal/irradiation bowing of subassemblies, and (iii) changes in core power profile during reactor operation cycles has been carried out. The studies indicate the maximum possible dilution in fuel and blanket subassemblies to be 2.63 K and 46.84 K, respectively. Shifting of thermocouple positions radially outward by 20 mm with respect to subassembly centers leads to an overall improvement in accuracy of thermocouple readings. It is also seen that subassembly blockage that leads to 7% flow reduction in fuel subassembly and 12% flow reduction in blanket subassembly can be detected effectively by the core-temperature monitoring system.     

Sodium boiling experiments in a 7-pin bundle: Investigations into some safety aspects of the KNK II driver zone

A series of sodium boiling experiments, in which the thermohydraulic characteristics of KNK II driver subassemblies were simulated, has been carried out for the purpose of studying the effects of two types of cooling disturbances: rapid flow interruption and those flow reductions which develop rather gradually. Information about the spatial and temporal development of boiling, voidage and dryout was obtained. Furthermore the feasibility of individual subassembly temperature monitoring has been investigated as well as that of two integral boiling detection methods based on acoustic noise and reactivity measurements.     

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.     

钠沸腾数学模型及其在BE+1上的验证

快堆在发生单个燃料组件瞬间完全堵流事故时,会引起堆芯内的冷却剂沸腾.钠沸腾所形成的压力和物质分布对后期事故的发展有重要影响.为了对事故进行总体性分析,本文选择两流体六方程模型,用子通道的方法进行网格划分,用D. R. Liles等人开发的半隐数值方法进行求解;在法国的BE 1实验中进行了模型验证;根据计算结果,对事故下钠的两相热工水力行为进行了解释.     

Heat transfer from a totally blocked fuel subassembly of a liquid metal fast breeder reactor: Part I. Experimental investigation

The safety issues of liquid metal fast breeder reactors (LMFBR) are crucial due to the fact that a highly reactive and hazardous fluid like liquid sodium is used as coolant. One of the extreme cases, which can occur in a fuel subassembly of an LMFBR, is a total blockage of liquid inside the subassembly, which may lead to boiling of sodium. The present study addresses this problem by conducting experiments on a 19-rod bundle assembly enclosed inside a tall hexagonal enclosure. Liquid sodium is used as the heat transfer fluid. The natural convection mode of heat transfer is the main focus of investigation with a co-flowing air through an annular packed bed to simulate the neighbouring subassemblies. The maximum temperature achieved under different rates of power generations and air flow conditions are observed. Also the radial temperature distributions at different planes under different operating conditions of power and air flow rates have been observed. The results are of significant importance for validating analysis for the purpose of prediction of boiling incipience in an LMFBR subassembly under totally blocked condition.     

Preliminary thermal-hydraulic phenomena investigation during total instantaneous blockage accident for CEFR

The Chinese Experimental Fast Reactor (CEFR) is under installation and commissioning right now. It is essential to investigate core disruptive accidents (CDAs) for the evaluation of CEFR's safety characteristic. As part-I preliminary investigation, accident of total instantaneous blockage (TIB) in single subassembly scale is modeled and analyzed. The degradation scenario has been calculated by a fluid-dynamics analysis code for liquid–metal fast reactors (LMFRs). For further investigation of accident process and influence to the neighboring bundles, seven subassembly domain is then simulated and calculated as part-II investigation. Total instantaneous blockage is assumed to occur in the center subassembly under normal operating conditions and consequences to neighboring assemblies are studied. The result shows that the key events, such as sodium boiling, clad melting, fuel particles relocation, hexcan melt-through and melt propagation into neighboring six assemblies symmetrically are adequately simulated. From comparison and discussion of the CEFR's results with the SCARABEE tests and Superphenix (SPX1)-type reactor simulation, it is concluded that all the key events appear in the same sequence whereas the propagation is limited in neighboring six assemblies. The discrepancy is probably due to less fuel inventory and better cooling capacity in CEFR subassembly design. TIB calculations help to give a better understanding and prediction of hypothetical accident scenario in subassembly blockage accidents for CEFR.     

Detection of local boiling in an LMFBR subassembly by temperature fluctuations analysis at the outlet

The aim of the experiments is to detect boiling in a sodium cooled subassembly by measuring fluctuations behind the bundle outlet. The measurements were carried out on an electrically heated 28-rod bundle with a partially blocked section. Fast responding thermocouples were installed downstream of the bundle outlet and downstream of a flow mixing system. Statistical parameters were investigated such as root mean square (RMS) and power spectrum density (PSD). The boiling conditions were generated by reducing the system pressure or flow velocity reduction. The experiments have shown that statistical analysis of temperature fluctuations can produce significant results in the detection of boiling behavior at both the outlet of a subassembly, and behind a flow mixing system.     

Establishment of the design requirements for a flow blockage detection system through a LES analysis of the temperature fluctuation in the upper plenum

When an assembly in a core is partially blocked, the temperature in the upper plenum fluctuates at an early stage. Therefore, the temperature fluctuation in the upper plenum can provide the information about a local blockage of an assembly. For developing the detection algorithm for the partial blockage, we analyzed the temperature fluctuation in the upper plenum due to the partial blockage in an assembly. The LES turbulence model in the CFX code was used for analyzing the temperature fluctuation in the upper plenum because the LES is suitable for analyzing the time dependent turbulence variables. After analyzing the temperature fluctuations in the upper plenum, we established basic design requirements for the flow blockage detection system through a FFT analysis and some statistical analysis. We concluded that response time of a measuring device was less than 13 m s and that it should cover a high temperature range of 1000 K. In addition, the resolution of the thermocouple was less than 2 K and its location should be within 25 cm from the exit of each assembly.     

A comparative study of the MATRA-LMR-FB calculation with the SABRE result for the flow blockage accident in the sodium cooled fast reactor

A sub-channel flow blockage may be initiated by an ingression of damaged fuel debris or foreign obstacles into a core subassembly for the sodium cooled fast reactor (SFR) due to the compact design of the fuel arrangement. Since local coolant temperature could go up high enough to reach a safety limit by the blockage disturbance in the subassembly, the MATRA-LMR-FB code was developed to analyze such blockage effect. An effort has been undergoing to enhance its reliability.In this study, a code-to-code comparison analysis with another code, SABRE4, was performed to supplement a qualification of the MATRA-LMR-FB. The two codes were applied to the analysis of partial sub-channel blockage accidents in a subassembly of the KALIMER-150, which is a conceptual design of a sodium-cooled fast reactor with an electric output of 150 MW. The analyses were carried out not only for radially different blockage positions but also for different blockage sizes in the subassembly.In result, the two code results were generally agreed both in magnitude and trend within a range. Therefore, it was concluded that the comparison results could support complementarily the applicability of the MATRA-LMR-FB to the partial flow blockage accident in the subassembly of the SFR.     

Local Sodium Boiling behind Local Flow Blockage in Simulated LMFBR Fuel Subassembly

This paper deals with local sodium boiling in the downstream of a six-subchannel blockage in an electrically heated LMFBR fuel subassembly mock-up.

The first series of experiments were conducted to measure temperature distributions in the downstream of the blockage under non-boiling conditions. The measured temperature rise due to the blockage agreed fairly well with the calculation by the LOCK code.

The second series of experiments were performed to investigate local boiling phenomena. In the local boiling region, no flow instability was observed since the sub-channels near the wrapper wall were still filled with sub-cooled liquid. In the nearly bulk boiling region, however, considerable upstream voiding occurred and then the inlet flow decreased, leading to final dryout.

The boiling caused a considerable increase in acoustic noise intensity. The root-mean-square (RMS) noise level of approximately 20 mbar obtained in the present local boiling experiments with sodium was much higher than that (approximately 0.5 mbar) in the ordinary nucleate boiling experiments with water. The peak observed in the hertz ranges was due to the repetition of bubble formation and collapse. In the kilohertz ranges, however, resonance peaks were superposed on a smooth curve with a broad peak at approximately 7 kHz.

The frequency (2.9 and 20.2 sec^?1) of bubble formation decreased with the increase of the bubble size at its point of maximum development. The product of the bubble frequency and the equivalent diameter was found to be constant.     

Analysis of temperature noise propagation in a large fuel subassembly of FBR

We have developed a method to calculate the three-dimensional distribution of root-mean-square (RMS) values of temperature noise in the single phase flow in a fast reactor fuel subassembly with a local flow blockage. Employed are the subchannel method in a pin bundle region and the finite difference method in the region downstream of the bundle. We have compared the calculated RMS values of temperature noise with experimental data for a sodium loop test using a wire-spacered 91-pin-bundle fuel sub-assembly with a local blockage. We have investigated the possibility of detection of the blockage by temperature noise by taking into account the influence of structures in the upper part of the subassembly.     

快堆严重事故工况下燃料组件盒破损机理的研究

在钠冷快增殖堆安全性分析中,六角形不锈钢燃料组件盒的破损时间和位置是一个重要的问题.对于严重的局部事故而言,钢盒破损的可能性基本上等同于事故向邻近燃料组件蔓延的可能性.本文以SCARABEE-N系列实验和SIMBATH系列实验为基础,对快堆严重事故工况下六角形钢组件盒的破损机理进行了研究.对于冷却状况良好的组件盒,提出了一种新的熔穿机理:局部热侵蚀进而诱发钠侧局部烧干,随后发生熔穿.在此基础上,对中国实验快堆(CEFR)在单个燃料组件瞬间完全堵流事故工况下组件盒破损的时间进行了预测.预测结果为,相邻燃料组件的六角形钢盒应该在堵流后7.2～8 s发生熔穿,随后事故开始向相邻的燃料组件蔓延.     

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.     

Hydraulic simulation of local blockage in a LMFBR fuel subassembly

Experimental studies of local flow blockage in a LMFBR fuel subassembly have been carried out using a simulating model in a water test loop. The studies verified the numerical results of an analytical code. The experiments were conducted in a 61-pin bundle containing a planar blockage without leakage flow. Central and edge blockage were used. The wake flow behind the blockage was visualized with dye or air bubble injection to grasp the flow characteristics. The velocity, static pressure and residence time were measured in the wake flow region.     

Analysis of the central blockage wake in an LMFBR-subassembly

A simple model is presented to estimate the wake size and the average velocity of the circulating flow downstream of a central blockage in a subassembly. A computer code, newly developed for the purpose was used to examine the model results. It is shown that both the model results and the code closely fit experimental results.It is seen that the size of a turbulent wake is nearly independent of inlet coolant velocity, whereas the velocity of the circulating glow in the wake is nearly proportional to the inlet velocity. Both the length of the wake and the velocity in it increase as the blockage increases. The maximum temperature is reached with a blockage of more than 0.6 of the subassembly cross section, and varies inversely with inlet velocity.     

Analysis of three different types of blockage in a sodium flow path with the MATRA-LMR-FB code

A local blockage in a subassembly of an LMR is of particular importance because the local temperature of the coolant increases at the downstream of a blockage and the integrity of the fuel clad can be threatened when an obstacle or a blockage is formed in a flow path. To analyze a flow blockage in a nuclear reactor core, Korea Atomic Energy Research Institute developed a subchannel analysis computer code MATRA-LMR-FB. This code adopts several enhanced modeling features such as a distributed resistance model, state-of-the-art turbulent mixing models, a hybrid difference scheme, and a porous body pressure drop model, therefore, it is applicable to a flow path with a plate-type or a porous-type blockage. The effect of each model has been evaluated through an analysis for the THORS experiment, in which a plate-type blockage was located in a flow channel with wire-wrapped fuel rods. The overall capability of the code has also been evaluated for the KNS experiment with a plate-type blockage in a grid-spaced flow channel, and for the SCARLET-II experiments with a porous blockage in channels formed by wire-wrapped fuel rods. The code shows good predictions for the experiments with a wire-wrapped flow path with a plate-type or a porous type blockage. The analyses for the KNS experiments reveal that the code requires a precise blockage model related to a grid spacer model.     

单盒钠冷快堆燃料组件堵流事故的CFD分析

钠冷快堆大都采用金属绕丝来固定燃料组件,细长狭窄的流道容易积聚腐蚀沉积物,可能会引起钠的局部沸腾和包壳的传热恶化。本文利用商用计算流体动力学软件STAR-CCM+程序对中国实验快堆单盒燃料组件的堵流事故进行了数值模拟,分析了包壳内壁面温度与冷却剂在堵块附近的轴向流场分布,并与正常工况下的计算结果进行对比。计算结果表明：实心介质堵流危害比多孔介质更为严重;实心介质堵流事故的包壳峰值温度局部最高点始终位于堵块中心位置,而多孔介质堵流事故的位于堵块后方,且随堵块面积的增大而往下游偏移;堵块的孔隙率对包壳在堵块下游的最大温升有明显影响,随堵块孔隙率的增大而减小。     

Sodium boiling experiments in the THORS facility

Experiments have been performed with 19- and 61-pin test assemblies in the Thermal-Hydraulic Out-of-Reactor Safety (THORS) Facility at the Oak Ridge National Laboratory (ORNL) since 1971. The THORS Facility is a high-temperature sodium system operated for the US Liquid-Metal Fast Breeder Reactor (LMFBR) Safety Program. The facility is used primarily for testing simulated LMFBR fuel subassemblies (pin (bundles). High-performance, electrically heated fuel pin simulators (FPSs) duplicate the heat generating capabilities and the dimensional characteristics of the nuclear fuel pins. A number of test bundles have been built and operated to obtain base thermal-hydraulic data, inlet and heated zone blockage data, and transient boiling data. Five of these bundles have been operated under two-phase conditions. Sodium boiling for periods up to twelve minutes were sustained in one bundle. (The lengths of the periods were limited only by automatic data recording capability). Clad dryout occurred in several tests. Tests were run at widely varying conditions of flow and power density. Testing with nonuniform power distribution across the bundle was also a part of the program.A 19-pin bundle with 12 peripheral guard heaters and a 6-subchannel blockage around the center pin in the heated zone was tested. The test program for this bundle was designed to determine if local boiling in the wake of the blockage propagates radially or axially during quasi-steady-state conditions. Post-test inspection revealed that significant helical distortion of the FPSs occurred in the vicinity of the blockage plate. This distortion probably influenced the boiling behavior. In the more severe tests, boiling initiated at the outlet of the heated zone and propagated radially into the unblocked subchannels after it had progressed upstream to the blockage. The subchannel analysis codes, SABRE and COBRA, accurately predict the extent of the boiling region.Experimental and analytical studies of sodium boiling behavior in unblocked 19- and 61-pin bundles indicate that cooling can be maintained for a significant period of time beyond boiling inception in a flow-power transient. Quasi-steady-state boiling occurred under natural-convection conditions.Investigations of the temperature data indicate that the thermal-hydraulic behavior during boiling transients is determined by two-dimensional effects, and that one-dimensional models cannot accurately predict the important phenomena associated with sodium boiling in test bundles. The subchannel code SABRE-2P (with a simple two-phase multiplier boiling model) and the two-region equilibrium mixture code THORAX (developed at ORNL) accurately predict the two-dimensional behavior between boiling inception and dryout.Extrapolation of the data from the smaller bundle tests to full-size fuel assemblies shows that the time between boiling inception and dryout would be lower for a 217-pin bundle than for a 61-pin bundle for a comparable transient. However, the time delay would still be significant, especially in a heterogeneous reactor core.     

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.