HTGR包覆燃料颗粒碳化硅层细晶化研究

高温气冷堆(HTGR)是能适应未来能源市场的第四代先进核反应堆堆型之一,其固有安全性的第一道保障是使用的TRISO型包覆燃料颗粒。在TRISO型燃料颗粒4层包覆结构中,SiC包覆层是承受包覆燃料颗粒内压和阻挡裂变产物释放的关键层,制备高质量SiC包覆层是核燃料领域中的重大问题和关键技术之一。本文介绍高温气冷堆燃料颗粒的基本结构,详述制备SiC包覆层的流化床-化学气相沉积过程,提出SiC层细晶化这一研究方向,并系统阐述包覆燃料颗粒SiC包覆层细晶化的优势。在细晶化SiC材料制备方法方面,系统分析SiC粉体、陶瓷、薄膜和厚膜材料的研究现状,并结合本实验室前期研究成果提出制备细晶SiC包覆层的可行制备策略。     

Experimental Study on Base-triggered Explosions in Molten-tin/Water Systems

The mechanism of explosions triggered at the bottom of a pool of liquid (i.e., base-triggered explosions) was experimentally investigated by dropping molten tin into water. We studied the effect of the tin temperature, the water temperature, the water depth, and the distance between the walls at the bottom of the pool on the tendency for explosions to occur. The experimental results showed that base-triggered explosions occurred at the bottom surface of the tank when the water temperature was near its saturation temperature. The occurrence of base-triggered explosions was also affected by the water depth. According to the interaction at the base, the observed explosions could be classified into two types: instantaneous and delayed explosions. The difference between the triggering mechanisms was explained by considering the homogeneous nucleation temperature of the water and the instantaneous interfacial temperature. The occurrence of delayed explosions was also affected by the distance between the side walls at the bottom of the pool. As the bottom surface area decreased, the tendency for base-triggered explosion increased.     

Sodium Vapor Deposition,on Simulated Model of LMFBR Rotating Shield Plug Annulus, (I)

Experimental study was made to investigate the controlling factors on the vapor deposition rate on reactor operational shield plug annulus, which is exposed to the vapor entrained cover gas during reactor operation. Two simulated test assemblies having annuli were made for this purpose and were installed into a small test vessel. In the experiment, the average deposition rates on the annular walls of the test assemblies were measured for various pool temperatures, and their dependents upon such parameters as pool temperature, T_s (or the saturated vapor pressure P_s at pool surface), cover gas pressure P_g , and temperature drop ΔT_a across cover gas, were studied.

The results revealed that the dominant controlling factor was the vapor pressure P_s at pool surface. Dependent of the average deposition rate φbar_d. upon the above parameters was simply expressed by: φbar_d=Bσp_sD_sΔT_G , where, σ_s is the saturated vapor concentration at pool surface, D_s , the vapor diffusion coefficient, and B the proportional constant.

To these experimental results, the previously published evaporation rate data and the theoretical evaporation rate equation based on Epstein & Rosner's theory were reviewed. Then correlation between the deposition and the evaporation rates was discussed.     

Effects of molten material temperatures and coolant temperatures on vapor explosion

An observable experiment facility for low-temperature molten materials to be dropped into water was set up in this study to investigate the mechanism of the vapor explosion. The effect of the fuel and coolant interaction (FCI) on the vapor explosion during the severe accidents of a fission nuclear reactor has been studied. The experiment results showed that the molten material temperature has an important effect on the vapor explosion behavior and pressure. The increase of the coolant temperature would decrease the pressure of the vapor explosion.     

Fuel Fragmentation and Mechanical Energy Conversion Ratio at Rapid Deposition of High Energy in LWR Fuels

The fuel fragmentation is one of the important subjects in the field of molten fuel-coolant interaction (MFCI) since it is one of basic processes of the MFCI, and it has not yet been made clear enough. Accordingly, U02 fuel fragmentation was studied in a postulated reactivity initiated accident (RIA) condition by the Nuclear Safety Research Reactor (NSRR). The distribution of the size of fuel fragments was obtained through the experiments and the mechanism of fuel fragmentation was studied. Also, the relation between the conversion ratio of the mechanical energy to the thermal and the degree of fuel fragmentation was obtained experimentally.

It was revealed that the distribution of fuel fragments was well described in the form of logarithmic Rosin-Rammler's distribution law. The fuel fragmentation was found to be explained by the Weber-type hydraulic instability model and the internal pressurization model. It was also shown that the mechanical energy conversion ratio was inversely proportional to the volume-surface mean diameter defined as the ratio of the total volume of fragments to the total surface, and furthermore that it was influenced by the coolant subcooling and the volumetric ratio of fuel to water.     

Transient Boiling of Sodium in a Seven-Pin Bundle under Loss-of-Flow Conditions

Transient sodium boiling experiments have been conducted in an electrically heated 7-pin bundle under loss-of-flow conditions. Each run was made by reducing or stopping the inlet flow at constant heater power.

There was no strong effect of temperature ramp rate on the incipient-boiling (IB) wall superheat and the data were scattered with values as high as ~190°C in the range of the present experimental conditions.

The observed coolant voidings, except in some high superheat runs, were initially limited in the central subchannel because of the steep temperature gradient in the bundle, and then spread slowly. In order to describe these voiding processes, a two-dimensional voiding model was required. The initial expulsion acceleration of liquid was higher with higher IB wall superheat. The measured values were lower than the theoretical prediction from the one-dimensional liquid column expulsion analysis.

The bulk pressure rises at the initial vaporization were less than the vapor pressure corresponding to the IB wall superheat. The pressure pulse occurred at the vapor collapse correlated reasonably well with the re-entrant liquid velocity, but the measured value was less than the prediction from sodium hammer analysis.     

颗粒尺寸分析仪法测量包覆颗粒疏松层密度的研究

包覆燃料颗粒的质量对于高温气冷堆安全运行起着重要作用。低密度热解炭层作为包覆的第一层非常关键,关系到包覆燃料颗粒和燃料元件的性能质量。本文介绍一种用颗粒尺寸分析仪测量疏松热解炭层密度的方法,该方法采用颗粒尺寸分析仪测量包覆前后颗粒的直径,再结合天平称得包覆前后颗粒的质量,经过计算得到包覆燃料颗粒疏松热解炭层的密度。对该方法测量包覆燃料颗粒疏松层密度的测量精度进行了验证。结果表明,该方法的测量精度满足测试要求,且该方法快速、便捷,适于工程应用。     

Simulation Test on Interactions of Zircaloy Claddings Deforming in Fuel Assembly under LOCA Conditions

In a hypothetical LOCA of LWR, it is assumed that Zircaloy claddings of fuel assemblies occur ballooning and cause thermal and mechanical interactions among themselves. To investigate the phenomena, burst tests were conducted with a single rod fuel simulator placed in the center of eight non-pressurized external heater pipes. It was found that the deformation and rupture behavior of ballooning cladding which made contact with the pipes depends on the following conditions: (1) temperature difference between cladding and heater pipes; higher temperature in the cladding than in the pipes produces an extended deformation and contact area, (2) internal pressure of cladding, and (3) temperature level at which ballooning takes place.

It was observed and analyzed that cladding hoop tensile stress is largest at the inflection points generated in the contact surface of the expanding cladding.     

Effects of Rod Pre-Pressurization on Light Water Reactor Fuel Behavior during Reactivity Initiated Accident Conditions

Previously pressurized (pre-pressurized) fuel rod tests recently performed in the Nuclear Safety Research Reactor (NSRR) investigate the effects of initial internal pressure on fuel rod behavior during reactivity initiated accident (RIA) conditions. A single PWR type fuel rod was contained within a waterfilled, ambient temperature and ambient pressure capsule. The fuel rod was then heated by the pulsing operation of the NSRR.

Results from the tests show that the effect of pre-pressurization was significant for the fuel rods with initial internal pressure of 0.8 MPa and above, and fuel rod failure occurred from rupture of the cladding with lower threshold energy deposition for failure as the initial internal pressure was increased. The cladding rupture was governed mainly by the cladding temperature rise, not by the rod internal pressure rise during the transient. The relationships between cladding burst pressure and cladding burst temperature and between cladding strain and cladding temperature at cladding rupture obtained in the present study under an RIA condition agree with the results obtained from various in- and ex-reactor experiments under a LOCA condition, although the obtained time-averaged strain rate of the Zircaloy cladding was much greater than that in a LOCA condition.     

A Validation of ATR LOCA Thermal-hydraulic Code with a Statistical Approach

When cladding temperatures are measured for a blowdown experiment, cladding temperatures at the same elevation in the fuel bundle have usually some differences due to eccentricity of the fuel bundle and other reasons such as biased two-phase flow. In the present paper, manufacturing tolerances and uncertainties of thermal-hydraulics are incorporated into a LOCA code that is applied with the statistical method. The present method was validated with the results of different blowdown experiments conducted using the 6 MW blowdown facility simulating the Advanced Thermal Reactor (ATR). In the present statistical method, the code was modified to run fast in order to calculate the blowdown thermal-hydraulics a lot of times with the code using different sets of input data. These input data for sizes and empirical correlations are prepared by the effective Monte-Carlo method based on the distribution functions deduced by the measured manufacturing errors and the uncertainties of thermal hydraulics. The calculated curves express uncertainties due to the different input deck. The uncertainty band and tendency of the cladding temperature were dependent on the beak sizes in the experiment. The measured results were traced by the present method.     

Effects of UO2/Zircaloy-4 Mole Ratios on Reaction of UO2 Fuel with Molten Zircaloy-4

Dissolution of UO₂ crucibles by molten Zircaloy-4 (Zry) was investigated in the temperature range of 2,223-2,373 K and for specimens having UO₂/Zry mole ratios between 7 and 18.2. The uranium concentration in the Zry melt rapidly increased during initial reaction time and approached saturated values, depending on reaction temperature and UO₂/Zry mole ratio. Kinetics of uranium concentration increase in the melt was analyzed based on a natural convection mass transfer model that takes into account the change of contact surface area/melt volume ratio with reaction time. The saturated uranium concentration in the Zry melt was inversely proportional to the U0₂/Zry mole ratio. An empirical correlation of saturated uranium concentration in the Zry melt was obtained as a function of UO₂/Zry mole ratios and reaction temperature. This study of the empirical correlation was intended to estimate maximum UO₂ fuel dissolution by molten Zry cladding during severe fuel damage accidents for three different reactor type fuels.     

Fragmentation of a Single Molten Metal Droplet Penetrating Sodium Pool I Copper Droplet and the Relationship with Copper Jet

The progression of hypothetical core disruptive accidents in metallic fuel fast breeder reactors is strongly affected by the fragmentation of molten metallic fuels due to the molten fuel-coolant interaction (FCI). As a basic study of FCI, the present paper focuses on the fragmentation of a single molten copper droplet with mass from 1 to 5 g, whichpenetrated a sodium pool at instantaneous contact interface temperatures (T_i) from 995 to 1,342°C. Intensive fragmentation of a single molten copper droplet was clearly observed even if T_i values are below the melTsing point (1,083°C) of copper besides the higher T_i range. The intensive fragmentation shows that the mass median diameters (D_m) of copper droplets with a fivefold difference in mass or the same mass have little difference, i.e., they are nearly the same. Under the lower T_i condition, the D_m data of droplet fragments of both the same and different masses scatter widely. It is found that the present D_m/D₀ data of mass median diameter normalized by the diameter before touching sodium (D0) give a distribution with larger values than those of molten copper jets with large mass from 20 to 300 g under the lower T_i condition, which were previously reported by the authors, because of the limited amount of heat of droplets. The present Dm=D0 data under the higher T_i condition are found to show an effecTive fragmentation compared with those of molten copper jets with a large mass of 4 kg.     

A review on the clad failure studies

In this paper, an attempt has been made to systematically organize the research investigations conducted on clad tube failure, so far. Before presenting the review on the clad failure studies, an introduction to different clad materials has been added, in which the effect of alloying elements on the material properties have been presented. The literature on clad failure has been broadly categorized under the headings LOCA and RIA. The failure mechanisms like creep, corrosion and pellet-clad interaction have been discussed in details. Each subsection of the review has been provided with summary table, in which the studies are arranged in the chronological order. A small section on acceptance criteria for ECCS has also been included. The last section of the review has been dedicated to the core-degradation phenomena.     

Assessment of fuel rod performance codes under ramp scenarios investigated within the SCIP project

The behavior of advanced cladding materials under challenging conditions needs to be fully characterized, understood and modeled. This paper assesses the current predictability of fuel performance codes under loadings expected from pellet-clad mechanical interactions. A set of scenarios experimentally characterized within the SCIP project, were chosen so that a variety of materials and ramp power sequences could be examined.Four codes have been used in this study: ALCYONE V1.1, FALCON-PSI, FRAPCON-3 v3.3 and STAV7.3. Their predictions have been compared to data in terms of cladding oxidation, diameters and elongation. Predictability of clad oxidation was certainly scattered and while some codes showed reasonable accuracy, other results were notably deviated. As for diameters, most of the codes were capable of qualitatively capturing the axial profile, and showed consistency between diameters and hoop stress and strain predictions. Elongation estimates were generally poor, and were rather far from measurements in most cases (even the trends observed just vaguely followed by the codes).The results reported have been discussed in the light of the set of individual hypotheses and approximations made by modelers and codes regarding both boundary conditions (i.e., power histories, inlet coolant temperature, refabrication, etc.) and fuel and clad characterization (i.e., densification, rim porosity, materials properties, etc.). Additionally, code-to-code comparisons of some key variables (i.e., fuel temperature, contact pressure, hoop and axial stresses, etc.) highlighted systematic tendencies of the codes and supported the observations made.     

Two types of a passive safety containment for a near future BWR with active and passive safety systems

The paper presents two types of a passive safety containment for a near future BWR. They are named Mark S and Mark X containment. One of their common merits is very low peak pressure at severe accidents without venting the containment atmosphere to the environment. The PCV pressure can be moderated within the design pressure. Another merit is the capability to submerge the PCV and the RPV above the core level. The third merit is robustness against external events such as a large commercial airplane crash. Both the containments have a passive cooling core catcher that has radial cooling channels. The Mark S containment is made of reinforced concrete and applicable to a large power BWR up to 1830 MWe. The Mark X containment has the steel secondary containment and can be cooled by natural circulation of outside air. It can accommodate a medium power BWR up to 1380 MWe. In both cases the plants have active and passive safety systems constituting in-depth hybrid safety (IDHS). The IDHS provides not only hardware diversity between active and passive safety systems but also more importantly diversity of the ultimate heat sinks between the atmosphere and the sea water. Although the plant concept discussed in the paper uses well-established technology, plant performance including economy is innovatively and evolutionally improved. Nothing is new in the hardware but everything is new in the performance.     

Variations of a passive safety containment for a BWR with active and passive safety systems

The paper presents variations of a certain passive safety containment for a near future BWR. It is tentatively named Mark S containment in the paper. It uses the operating dome as the upper secondary containment vessel (USCV) to where the pressure of the primary containment vessel (PCV) can be released through the upper vent pipes. One of the merits of the Mark S containment is very low peak pressure at severe accidents without venting the containment atmosphere to the environment. Another merit is the capability to submerge the PCV and the reactor pressure vessel (RPV) above the core level by flooding water from the gravity-driven cooling system (GDCS) pool and the upper pool. The third merit is robustness against external events such as a large commercial airplane crash owing to the reinforced concrete USCV. The Mark S containment is applicable to a large reactor that generates 1830 MW electric power. The paper presents several examples of BWRs that use the Mark S containment. In those examples active safety systems and passive safety systems function independently and constitute in-depth hybrid safety (IDHS). The concept of the IDHS is also presented in the paper.     

Integrated software safety analysis method for digital I&C systems

The digitalized Instrumentation and Control (I&C) system of Nuclear power plants can provide more powerful overall operation capability, and user friendly man-machine interface. The operator can obtain more information through digital I&C system. However, while I&C system being digitalized, three issues are encountered: (1) software common-cause failure, (2) the interaction failure between operator and digital instrumentation and control system interface, and (3) the non-detectability of software failure. These failures might defeat defense echelons, and make the Diversity and Defense-in-Depth (D3) analysis be more difficult. This work developed an integrated methodology to evaluate nuclear power plant safety effect by interactions between operator and digital I&C system, and then propose improvement recommendations. This integrated methodology includes component-level software fault tree, system-level sequence-tree method and nuclear power plant computer simulation analysis. Software fault tree can clarify the software failure structure in digital I&C systems. Sequence-tree method can identify the interaction process and relationship among operator and I&C systems in each D3 echelon in a design basis event. Nuclear power plant computer simulation analysis method can further analyze the available backup facilities and allowable manual action duration for the operator when the digital I&C fail to function. Applying this methodology to evaluate the performance of digital nuclear power plant D3 design, could promote the nuclear power plant operation safety. The operator can then trust the nuclear power plant than before, when operating the highly automatic digital I&C facilities.     

An overview of instrumentation and control systems of a Korea standard nuclear power plant: A signal interface standpoint

This paper presents an overview of instrumentation and control (I&C) systems of a pressurized water reactor (PWR) type nuclear power plant (NPP) in Korea. Yonggwang unit 3, which was constructed as a basis model for a Korea standard nuclear power plant (KSNP), is selected as an example for the presentation. This overview is derived from analyzing the I&C systems based on a top-down approach. The I&C systems consist of 30 systems. The 183 I&C cabinets are also analyzed and mapped to the systems. The overview is focused on an interface between the systems and the cabinets. This information will be used to understand the implementation of the I&C systems and to group the systems for an upgrade.