Sedimentation behavior of mixed solid particles

During the material relocation phase of core-disruptive accidents in sodium-cooled fast reactors, the sedimentation behavior of fragmented debris discharged from the reactor core into the lower plenum region leading to a debris-bed formation is crucial in regard to in-vessel retention and safety concerns. The height of the beds formed may influence both the cooling of the bed from the decay heat in the fuel and the neutronic characteristics. To develop an experimental database of bed formation behavior, a series of experiments using simulant materials, namely, Al₂O₃, ZrO₂, and stainless steel, were performed under gravity-driven discharge of solid particles from a nozzle into a quiescent cylindrical water pool. The bed height was measured for particles of different size, density, and sphericity, and an injection nozzle with varying diameter, injection velocity, and injection height. From these experiments, an empirical correlation was established to predict the bed height for both homogeneous and mixed particles for the different properties. This correlation reproduces reasonably well the experimental trend in bed height with critical factors, which were identified in this and previous experiments.     

An investigation on debris bed self-leveling behavior with non-spherical particles

Studies on debris bed self-leveling behavior with non-spherical particles are crucial in the assessment of actual leveling behavior that could occur in core disruptive accident of sodium-cooled fast reactors. Although in our previous publications, a simple empirical model (based model), with its wide applicability confirmed over various experimental conditions, has been successfully advanced to predict the transient leveling behavior, up until now this model is restricted to calculations of debris bed of spherical particles. Focusing on this aspect, in this study a series of experiments using non-spherical particles was performed within a recently developed comparatively larger scale experimental facility. Based on the knowledge and data obtained, an extension scheme was suggested with the intention to extend the base model to cover the particle-shape influence. The proposed scheme principally consists of two parts – with one part for correcting the terminal velocity of a single non-spherical particle, which is the key parameter in our base model, and the other for representing the additional particle–particle interactions caused by the shape-related parameters. Through detailed analyses, it is found that by coupling this scheme, good agreement between experimental and predicted results can be achieved for both spherical and non-spherical particles given current range of experimental conditions.     

Development of the evaluation methodology for the material relocation behavior in the core disruptive accident of sodium-cooled fast reactors

The in-vessel retention (IVR) of core disruptive accident (CDA) is of prime importance in enhancing safety characteristics of sodium-cooled fast reactors (SFRs). In the CDA of SFRs, molten core material relocates to the lower plenum of reactor vessel and may impose significant thermal load on the structures, resulting in the melt-through of the reactor vessel. In order to enable the assessment of this relocation process and prove that IVR of core material is the most probable consequence of the CDA in SFRs, a research program to develop the evaluation methodology for the material relocation behavior in the CDA of SFRs has been conducted. This program consists of three developmental studies, namely the development of the analysis method of molten material discharge from the core region, the development of evaluation methodology of molten material penetration into sodium pool, and the development of the simulation tool of debris bed behavior. The analysis method of molten material discharge was developed based on the computer code SIMMER-III since this code is designed to simulate the multi-phase, multi-component fluid dynamics with phase changes involved in the discharge process. Several experiments simulating the molten material discharge through duct using simulant materials were utilized as the basis of validation study of the physical models in this code. It was shown that SIMMER-III with improved physical models could simulate the molten material discharge behavior, including the momentum exchange with duct wall and thermal interaction with coolant. In order to develop an evaluation methodology of molten material penetration into sodium pool, a series of experiments simulating jet penetration behavior into sodium pool in SFR thermal condition were performed. These experiments revealed that the molten jet was fragmented in significantly shorter penetration length than the prediction by existing correlation for light water reactor conditions, due to the direct contact and thermal interaction of molten materials with coolant. The fragmented core materials form a sediment debris bed in the lower plenum. It is necessary to remove decay heat safely from this debris bed to achieve IVR. A simulation code to analyze the behavior of debris bed with decay heat was developed based on SIMMER-III code by implementing physical models, which simulate the interaction among solid particles in the bed. The code was validated by several experiments on the fluidization of particle bed by two-phase flow. These evaluation methodologies will serve as a basis for advanced safety assessment technology of SFRs in the future.     

Visual study of ex-pin phenomena for SFR with metal fuel under initial phase of severe accidents by using simulants

In the present Korean sodium-cooled fast reactor (SFR) program, early dispersion of the molten metal fuel within a subchannel is suggested as an inherent safety strategy in the initiating phase of a hypothetical core disruptive accident (HCDA). This safety strategy provides a negative reactivity driven by the melt dispersion; therefore, it could reduce the possibility of occurrence of a severe recriticality event. In the initiating phase, the melt could be injected into the subchannel horizontally by the internal pressure of the fuel pin. Complex phenomena occur during intermixing of the melt with the coolant after the horizontal injection of the melt. It is rather difficult to understand the several combined mechanisms that occur that are related to the dispersion and fragmentation of the melt. Thus, it seems worthwhile to study the horizontal injection of melt at lower temperatures, which could help to observe the dispersion phenomenon and understand the fragmentation mechanism. In this work, for a parametric study, tests were performed under structural conditions, coolant void conditions, and boiling conditions. As a result, in some cases, the injected molten materials were stuck around the injection hole. On the other hand, the molten materials were dispersed upward sufficiently well under the boiling condition when R123 was used as the coolant. The built-up vapor pressure was found to be one of the driving forces for the upward dispersion of the molten materials.     

Empirical correlations for predicting the self-leveling behavior of debris bed

Studies on the self-leveling behavior of debris bed are crucial for the assessment of core-disruptive accident (CDA) occurred in sodium-cooled fast reactors (SFR). To clarify this behavior over a comparatively wider range of gas velocities, a series of experiments were performed by injecting nitrogen gas uniformly from a pool bottom. Current experiments were conducted in a cylindrical tank, in which water, nitrogen gas and different kinds of solid particles, simulate the coolant, vapor (generated by coolant boiling) and fuel debris, respectively. Based on the quantitative data obtained (mainly the time variation of bed inclination angle), with the help of dimensional analysis technique, a set of empirical correlations to predict the self-leveling development depending on particle size, particle density and gas injection velocity was proposed and discussed. It was seen that good agreement could be obtained between the calculated and experimental values. Rationality of the correlations was further confirmed through detailed analyses of the effects of experimental parameters such as particle size, particle density, gas flow rate and boiling mode. In order to facilitate future analyses and simulations of CDAs in SFRs, the obtained results in this work will be utilized for the validations of an advanced fast reactor safety analysis code.     

A new empirical model for self-leveling behavior of cylindrical particle beds

ABSTRACT

During the material relocation phase of core disruptive accidents in sodium-cooled fast reactors, the rapid quenching and fragmentation of molten materials discharged from the reactor core into the lower plenum region can lead to the formation of debris beds. Coolant boiling may lead to leveling of the mound-shaped beds, which changes both the beds' coolability with decay heat in the fuel and the neutronic characteristics. In this study, a series of experiments using simulant materials were performed to develop an experimental database of self-leveling processes of particle beds in a cylindrical system. To simulate the coolant boiling in the beds in the experiments, a gas injection method was used to percolate nitrogen gas uniformly through the base of a bed with a conical-shaped mound. Time variations in bed height during the self-leveling process were measured for different particle sizes, densities and sphericities, and gas injection velocities. Using a dimensional analysis approach, a new model was proposed. This model correlates the experimental data on transient bed height with an empirical equation using a characteristic time for self-leveling development and an equilibrium bed height. The proposed model reasonably predicts the self-leveling development of particle beds.     

Melting behavior and thermal conductivity of solid sodium-concrete reaction product

This study revealed melting behavior and thermal conductivity of four samples generated by sodium-concrete reaction (SCR). We prepared the samples using two methods such as firing mixtures of sodium (Na) and grinded concrete powder, and sampling depositions after the SCR experiments. In the former, the mixing ratios were determined from the past experiment. The latter simulated the more realistic conditions such as the temperature history and the distribution of Na and concrete. The thermogravimetry-differential thermal analyzer (TG-DTA) measurement showed the temperatures of the onset of the melting (solidus temperatures) were 865–942°C, but those of the samples containing metallic Na could not be clarified. In the two more realistic samples, the compression moldings in a furnace were observed. The observation revealed the softening temperature was 800–840°C and the solidus temperature was 840–850°C, which was 10–20°C lower than the TG-DTA results. The thermodynamics calculation of FactSage 7.2 revealed the solidus temperature was caused by melting of the some components such as Na₂SiO₃ and/or Na₄SiO₄ and NaAlO₂. Moreover, the thermal conductivity was λ ~ 1–3 W/m-K, which was comparable to xNa₂O - (1 - x)SiO₂ (x = 0.5, 0.33, and 0.25), and that at 700°C was explained by NBO/T of Equation (1).     

Distance for fragmentation of a simulated molten-core material discharged into a sodium pool

A series of experiments has been carried out to obtain experimental knowledge on the distance for fragmentation of a molten core material discharged into the sodium plenum during postulated core disruptive accidents of sodium-cooled fast reactors. In the current experiments, 0.9 kg of molten aluminum (initial temperature: around 1473 K) was discharged into a sodium pool (diameter: 0.11 m, depth: 1 m, initial temperature: 673 K) through a nozzle (inner diameter: 20 mm). Visual observation of the fragmentation behavior was performed using an X-ray imaging system. The following experimental results were obtained. (1) Liquid column of molten aluminum was intensively fragmented almost simultaneously with a rapid expansion of sodium vapor in the vicinity of the column. (2) Due to the intensive fragmentation, penetration of the liquid column was limited to approximately 100 mm or so from the sodium level. (3) The molten aluminum was rapidly cooled after the intensive fragmentation. Based on these results, the distance for fragmentation of the liquid column was estimated to be 100 mm in the experiments. Through the current experiment, useful knowledge was obtained for the future development of an evaluation method of the distance for fragmentation of the molten core material.     

Self-leveling behavior of mixed solid particles in cylindrical bed using gas-injection method

ABSTRACT

The self-leveling of debris beds is a critical phenomenon to be clarified in the safety assessment of material relocation and decay heat-removal phases in postulated core-disruptive accidents in sodium-cooled fast reactors. In recent years, we have conducted several series of experiments using simulant materials to develop an experimental database of self-leveling processes in particle beds and proposed an empirical model to predict the bed height during self-leveling. The obtained experimental knowledge and proposed model applicability have been limited to homogeneous particles, although fragmented core debris could be mainly mixtures of fuel and stainless-steel particles with a size distribution. We conducted self-leveling experiments using mixtures of solid particles with different properties to understand the characteristics of self-leveling phenomena under a wider range of conditions. An improved empirical model for transient bed height was developed to correlate the experimental data under various conditions of particle mixtures. The developed model reproduces the self-leveling development of mixed particle beds and those of the homogeneous particle reasonably.     

Development of safety assessment methodology on fuel element failure propagation in SFR and its application to Monju

Probabilistic and deterministic safety assessments and experimental studies on local fault (LF) propagation in sodium-cooled fast reactors (SFRs) have been performed in many countries because LFs have been historically considered as one of the possible causes of severe accidents. Adventitious fuel pin failures have been considered to be the most dominant initiators of LFs in these probabilistic assessments because of its high frequency of occurrence during reactor operation and possibility of subsequent pin-to-pin failure propagation. Four possible mechanisms of fuel element failure propagation from adventitious fuel pin failure (FEFPA) were identified from a state-of-the-art review of open papers. All the mechanisms for FEFPA analysis including thermal, mechanical and chemical propagation are modeled into a safety assessment code which is applicable to arbitrary SFRs by developing some needed but missing methods. Furthermore, an assessment on FEFPA of Japanese prototype fast breeder reactor (Monju) was performed using this methodology. It was clarified that FEFPA was highly unlikely and limited at most within one subassembly in Monju owing to its redundant and diverse detection and shutdown systems for FEFPA even assuming the propagation. These results also suggested future possibility of run-beyond-cladding-breach operation which would enhance the economic efficiency in Monju.     

池式钠冷快堆事故源项计算方法研究及其应用

针对传统轻水堆事故源项计算方法不适用池式钠冷快堆的问题,分析可能发生的设计基准事故和超设计基准事故的释放路径,研究建立适用于池式钠冷快堆的堆芯损伤类、泄漏类和钠火类事故源项计算方法。结合示范快堆的6种典型事故：1盒燃料组件瞬时全部堵塞事故、反应堆堆本体覆盖气体边界泄漏事故、一次氩气衰变罐破损事故、主容器泄漏事故、一回路外无保护套管的钠净化管道泄漏事故和一回路无保护套管的外辅助管断裂或泄漏合并隔离阀关不住事故,开展事故源项计算及其剂量后果评价。结果表明：6种事故的放射性后果均低于GB 6249-2011的要求。该方法还可为回路式钠冷快堆、铅铋快堆以及气冷快堆事故源项计算提供参考。     

Particle Effects on Penetration and Solidification of Flowing Mixed Melts on Metal Structures

In severe-accident analyses of liquid-metal-cooled reactors, assessing the relocation and solidification of disrupted core materials is of importance. We investigate here the fundamental characteristics of these behaviors in flowing melt mixed with solid particles under various conditions. To simulate the melts, we use a low-melting-point metal (viz., Bi-Sn-In alloy) mixed with various concentrations of copper and bronze as solid particles; the flow channels used were inclined open ones with a V-shaped cross section made of either stainless steel or brass plate. Transient melt flow was recorded, and melt penetration lengths and frozen melt distributions along the channel were measured. Results indicate that penetration length decreases for molten-metal/solid particle mixtures (mixed melts) compared with a pure molten metal (a pure melt), as well as decreases with decreasing solid particle size and increasing particle volume fraction in the melt. For the pure melt, we found only one freezing mode of all melt adhesions along the channel, whereas there were two freezing modes of melt separation with high solid particle concentrations, as well as melt adhesion, along the channel for mixed melts. The results obtained will be utilized in an experimental database to validate and improve physical models used for reactor safety analysis codes.     

Study on Calculation Method of Accidental Source Term for Pool-type Sodium-cooled Fast Reactor and Its Application

To deal with the problem that the traditional light water reactor accidental source term calculation method is not suitable for sodium-cooled fast reactor, calculation methods for accidental source term of pool-type sodium-cooled fast reactor, including core damage type, leak type and sodium fire type, were studied and derived on basis of the analysis of release path of potential design basis accidents and beyond design basis accidents. The methods were applied to six typical accidents of the demonstration fast reactor, including the total instantaneous blockage of one fuel assembly, the leakage of cover gas region of reactor main vessel, the damage of primary argon decay tank, the leakage of main vessel, the leakage of sodium purification pipeline without protective sleeve outside the primary circuit, and the leakage of external auxiliary pipeline without protective sleeve outside the primary circuit or the isolation valve tube not be closed. The calculation of accidental source terms and their radiological consequences were carried out. The results show that the radioactive dose consequences of the six accidents are lower than the requirements of GB 6249-2011. The methods proposed can provide reference to the calculations of accidental source term of loop-type sodium-cooled fast reactor, lead-cooled fast reactor and gas-cooled fast reactor.     

A study on self-terminating behavior of sodium-concrete reaction (2)

Fundamental experiments were carried out to investigate the self-termination behavior of sodium–concrete reaction (SCR). In the reference experiment, the reaction time was controlled to investigate the distribution change of Na and the reaction products in the pool and around the reaction front. The measured concentrations at the reaction front were 18–24 wt.% for Na, 22–18 wt.% for Si, and 4–3.4 wt.% for Al and Ca after the self-termination. From the thermodynamics calculations, stable materials at the reaction front comprised more than 90 wt.% solid products such as Na₂SiO₃ and no Na. Furthermore, in two sensitivity experiments with additional heating or using mortar, the concrete-ablation behavior depended strongly on the reaction at the reaction front. It was concluded that SCR termination was caused by the lack of Na at the reaction front.

The distribution of Na and reaction products could be explained by a steady-state sedimentation–diffusion model. At the early stage of SCR, the reaction products were suspended as particles in the Na pool because of the high H₂-generation rate. As the concrete ablation proceeds, they start settling down due to the decreased H₂-generation rate, thereby allowing SCR termination.     

Analysis of fuel subassembly innerduct configurational effects on the core characteristics and power distribution of a sodium-cooled fast breeder reactor

The power distribution and core characteristics in various configurations of fuel subassemblies with an innerduct structure in the Japan sodium-cooled fast reactor were evaluated using a Monte Carlo code for neutron transport and burnup calculation. The correlation between the fraction of fuel subassemblies facing outward and the degree of power increase at the core center was observed regardless of the compositions. This indicated that the spatial fissile distribution caused by innerduct configurations was the major factor of the difference in the power distribution. A power increase was also found in an off-center region, and it tended to be greater than that at the core center because of the steep gradient of neutron flux intensity. The differences in the worth of control rods caused by the innerduct configurations were confirmed.     

Updating of local blockage frequency in the reactor core of SFR and PRA on consequent severe accident in Monju

Fuel subassemblies of sodium-cooled fast reactors (SFRs) are densely arranged and have high power densities. Therefore, the local fault has been considered as one of the possible initiating events of severe accidents. In the conventional analyses for the license of Japanese prototype SFR (Monju), according to the local fault evaluation under the condition of one sub-channel flow blockage in the analyses of design basis accident (DBA), it was confirmed that the pin failures were limited locally without severe core damage. In addition, local flow blockage of 66% central planar in the subassembly was historically investigated as one of the beyond-DBAs. However, it became clear that these deterministic analyses were not based on a realistic assumption by experimental studies. Therefore, probabilistic risk assessment on local fault which was initiated from local flow blockage was performed reflecting the state-of-the-art knowledge in this study. As a result, damage propagation from local fault caused by local flow blockage in Monju can be negligible compared with the core damage due to anticipated transient without scram or protected loss of heat sink in the viewpoint of both frequency and consequence.     

二维双区球流运动唯象方法的数值模拟

为研究燃料球和慢化球呈双区分布的球床式高温气冷堆球流运动的规律,以二维双区球流运动实验台架为参照,采用离散单元法(DEM)进行数值模拟。从唯象的角度对模拟结果进行分析,研究了双区分布的形成过程、中心区、交混区与滞留区以及速度分布等问题。模拟结果表明：在当前模拟条件下,可形成稳定的中心区,中心区与环形区之间存在交混区,且存在滞留区。本体内越靠近底部卸料口,竖直方向速度分布越不均匀,水平方向的扩散越来越大。     

二维球流运动摩擦系数影响的数值模拟

为研究摩擦系数对球流运动的影响,参照二维堆芯球流运动实验台架,采用离散单元法对球流运动进行了数值模拟,并对从标记点产生的球流区域的均值流线、标准方差和平均停留时间等进行了比较和分析。结果表明：球的摩擦系数对球流流场影响不大;球的摩擦系数越大,水平方向扩散越小,流动越均匀;壁面摩擦系数对水平方向扩散影响不大;壁面摩擦系数越大,流动越不均匀。     

Updating of adventitious fuel pin failure frequency in sodium-cooled fast reactors and probabilistic risk assessment on consequent severe accident in Monju

Experimental studies, deterministic approaches and probabilistic risk assessments (PRAs) on local fault (LF) propagation in sodium-cooled fast reactors (SFRs) have been performed in many countries because LFs have been historically considered as one of the possible causes of severe accidents. Adventitious-fuel-pin-failures (AFPFs) have been considered to be the most dominant initiators of LFs in these PRAs because of their high frequency of occurrence during reactor operation and possibility of fuel-element-failure-propagation (FEFP). A PRA on FEFP from AFPF (FEFPA) in the Japanese prototype SFR (Monju) was performed in this study based on the state-of-the-art knowledge, reflecting the most recent operation procedures under off-normal conditions. Frequency of occurrence of AFPF in SFRs which was the initiating event of the event tree in this PRA was updated using a variety of methods based on the above-mentioned latest review on experiences of this phenomenon. As a result, the frequency of occurrence of, and the core damage frequency (CDF) from, AFPF in Monju was significantly reduced to a negligible magnitude compared with those in the existing PRAs. It was, therefore concluded that the CDF of FEFPA in Monju could be comprised in that of anticipated transient without scram or protected loss of heat sink events from both the viewpoint of occurrence probability and consequences.