Mikio Isozaki

Experimental study on fuel-discharge behaviour through in-core coolant channels

In core-disruptive accidents of sodium-cooled fast reactors, fuel discharge from the core region reduces the possibility of severe re-criticality events. In-core coolant channels with large hydraulic diameters, such as the control-rod guide tube and a concept of the Fuel Assembly with Inner Duct Structure have a potential to provide effective fuel-discharge paths if effects of sodium in these paths on molten fuel discharge are limited. Two series of experiments were conducted to investigate fuel-discharge behaviour through the sodium-filled channels. In the first series of experiments, an alloy with low melting temperature was ejected into a water channel to clarify dominant phenomena for melt discharge through the coolant-filled channel and to develop methodologies for evaluating the effects of coolant on melt discharge. In the second series of experiments, a molten alumina was discharged through the sodium-filled channel in order to verify the applicability of the knowledge and evaluation methodologies obtained in the first series of experiments to the sodium-filled channel. These series of experiments showed that the discharge path can be entirely voided by the vaporisation of a part of the coolant at the initial melt discharge phase that this is followed by coolant vapour expansion and that melt penetrates significantly into the voided channel. Preliminary extrapolation of the present results to the in-core coolant channel suggests that the effects of the sodium on fuel discharge are limited and, therefore, in-core coolant channels will provide effective fuel-discharge paths for reducing neutronic activity.

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.

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.

SIMMER-III Analyses of Local Fuel-Coolant Interactions in a Simulated Molten Fuel Pool: Effect of Coolant Quantity

Studies on local fuel-coolant interactions (FCI) in a molten pool are important for the analyses of severe accidents that could occur for sodium-cooled fast reactors (SFRs). To clarify the mechanisms underlying this interaction, in recent years, several experimental tests, with comparatively larger difference in coolant volumes, were conducted at the Japan Atomic Energy Agency by delivering a given quantity of water into a molten pool formed with a low-melting-point alloy. In this study, to further understand this interaction, interaction characteristics including the pressure buildup as well as mechanical energy release and its conversion efficiency are investigated using the SIMMER-III, an advanced fast reactor safety analysis code. It is found that the SIMMER-III code not only reasonably simulates the transient pressure and temperature variations during local FCIs, but also supports the limited tendency of pressurization and resultant mechanical energy release as observed from experiments when the volume of water delivered into the pool increases. The performed analyses also suggest that the most probable reason leading to such limited tendency should be primarily due to an isolation effect of vapor bubbles generated at the water-melt interface.

Characteristics of Pressure Buildup From Local Fuel-Coolant Interactions in a Simulated Molten Fuel Pool

Studies on local fuel-coolant interactions (FCI) in a molten pool are important for severe accident analyses of sodium-cooled fast reactors (SFRs). To clarify the mechanisms underlying this interaction, in this study a series of experiments was conducted by delivering a given quantity of water into a simulated molten fuel pool (formed with a low-melting-point alloy). Based on the experimental data obtained from a variety of conditions, including difference in water volume, melt temperature and water subcooling, the characteristics of pressure-buildup during local FCIs was investigated. It is found that under our experimental conditions the water volume and melt temperature have remarkable impact on the interaction, while the role of water subcooling seems to be less prominent. The performed analyses also suggest that the pressurization from local FCIs should be intrinsically limited, due to a suppressing role caused by the increasing of coolant volume entrapped within the pool as well as the transition of boiling mode. Current work, which gives a palette of favorable data for a better understanding and an improved estimation of severe accidents in SFRs, is expected to benefit future analyses and verifications of computer models developed in advanced fast reactor safety analysis codes.© 2014 ASME

Mechanism of Upward Fuel Discharge During Core Disruptive Accident in Sodium-Cooled Fast Reactors

The elimination of severe power excursion with significant mechanical-energy release during Core Disruptive Accidents (CDAs) is a key issue for the enhanced safety of Sodium-cooled Fast Reactors (SFRs). In order to prevent the formation of a large-scale molten fuel pool within a reactor core, which is one of factors leading to the severe power excursion during CDAs, Japan Atomic Energy Agency (JAEA) is studying the introduction of Fuel Assembly with Inner Duct Structure (FAIDUS). In the current reference design for FAIDUS, the top end of the inner duct is open whereas the bottom end of the inner duct is closed, and therefore it is expected that the molten fuel will be discharged from a reactor core towards an upper sodium plenum through the inner duct. The objective of the present study is to clarify the fundamental mechanism for upward fuel discharge through the inner duct structure in FAIDUS, and thereby to confirm the effectiveness of FAIDUS.In the previous paper, the possibility of upward discharge of a high-density melt driven by coolant vapor was confirmed by visual observation in the JAEA’s out-of-pile experiment, in which molten Wood’s metal (density at the room temperature: 8700 kg/m3, melting point: 352 K) simulating the molten fuel was injected into a coolant channel (equivalent inner diameter: 30 mm, total height: 2 m, fluid content: water) simulating the inner duct structure. In this paper, the mechanism of upward discharge of a high-density melt driven by coolant vapor pressure and/or flow in this experiment is discussed in terms of the application to reactor conditions. Through this discussion, the following mechanisms were clarified.1) Coolant vapor pressure is built up within the coolant channel after the melt injection. The magnitude of the pressure buildup becomes larger with increase of melt-enthalpy-injection rate which is defined by the product of melt-mass-injection rate into the coolant channel and melt specific enthalpy.2) Following the pressure buildup, the melt is discharged upward being driven by the coolant vapor flow directing towards the top opening end of the coolant channel. The upward discharge mass rate becomes higher with the increase of the magnitude of the pressure buildup and therefore melt-enthalpy-injection rate.From these experimental knowledge, it was suggested that the coolant pressure buildup could act as one of the driving force for the upward discharge of a high-density melt through the inner duct structure in FAIDUS under reactor conditions with higher melt-enthalpy-injection rate than the current experimental condition.Copyright