Yoshiharu Tobita

Safety Strategy of JSFR Eliminating Severe Recriticality Events and Establishing In-Vessel Retention in the Core Disruptive Accident

In the Japan Sodium Cooled Fast Reactor (JSFR) design, elimination of severe power burst events in the Core Disruptive Accident (CDA) is intended as an effective measure to ensure retention of the core materials within the reactor vessel. The design strategy is to control the potential of excessive void reactivity insertion in the initiating phase by selecting appropriate design parameters such as maximum void reactivity on one hand, and to exclude core-wide molten-fuel-pool formation, which has been the main issue of CDA, by introducing an inner duct on the other hand. The effectiveness of these measures is evaluated based on existing experimental data and computer simulation with validated analytical tools. It is judged that the present JSFR design can exclude severe power burst events. Phenomenological consideration of general characteristics and preliminary evaluations for the long-term material relocation and cooling phases gave the perspective that in-vessel retention would be attained with appropriate design measures.

Self-Leveling Onset Criteria in Debris Beds

In a core-disruptive accident of a sodium-cooled fast breeder reactor, core debris may settle on the coresupport structure and/or in the lower inlet plenum of the reactor vessel because of rapid quenching and fragmentation of molten core materials in the subcooled sodium plenum. Coolant boiling is the mechanism driving the self-leveling of a debris bed that causes significant changes in the heat-removal capability of the beds. In the present study, we develop criteria establishing the onset of this self-leveling behavior that we base on a force balance model assuming a debris bed with a single-sized spherical particle. The model considers drag, buoyancy, and gravity acting on each particle. A series of experiments with simulant materials verified the applicability of this description of self-leveling. Particle size (between 0.5–6 mm), shape (spherical and nonspherical), density (namely of alumina, zirconia, lead, and stainless steel), along with boiling intensity, bed volume, and even experimental methods were taken into consideration to obtain general characteristics of the self-leveling process. We decided to use depressurization boiling to simulate an axially increasing void distribution in the debris bed, although bottom heating was also used to validate the use of the depressurization method. On the self-leveling onset issues, we obtained good agreement between model predictions and experimental results. Extrapolation of our model to actual reactor conditions is discussed.

A scenario of core disruptive accident for Japan sodium-cooled fast reactor to achieve in-vessel retention

As the most promising concept of sodium-cooled fast reactors, the Japan Atomic Energy Agency (JAEA) has selected the advanced loop-type fast reactor, so-called JSFR. The safety design requirements of JSFR for Design Extension Condition (DEC) are the prevention of severe accidents and the mitigation of severe-accident consequences. For the mitigation of severe-accident consequences, in particular, the In-Vessel Retention (IVR) against postulated Core Disruptive Accidents (CDAs) is required. In order to investigate the sufficiency of these safety requirements, a CDA scenario should be constructed, in which the elimination of power excursion and the in-vessel cooling of degraded core materials are evaluated so as to achieve IVR. In the present study, the factors leading to IVR failure were identified by creating phenomenological diagrams, and the effectiveness of design measures against them were evaluated based on experimental data and computer simulations. This is an unprecedented approach to the construction of a CDA scenario, and is an effective method to objectively investigate the factors leading to IVR failure and the design measures against them. It was concluded that mechanical/thermal failures of the reactor vessel due to power-excursion/thermal-load could be avoided by adequate design measures, and a clear vision for achieving IVR was obtained.

Establishment of freezing model for reactor safety analysis

A mechanistic simulation of molten core-material relocation is required to reasonably assess consequences of postulated core disruptive accidents (CDAs) in fast reactors (FRs). The dynamics of molten core-material freezing when it is driven into the channels surrounding the core region plays an important role since this affects fuel removal from the core region. Therefore, a mechanistic model for freezing behavior was developed and introduced into the FR safety analysis code, SIMMER-III, in this study. Based on the micro-physics of crystallization, two key assumptions, supercooling of melt in the vicinity of the wall and melt-wall contact resistance due to imperfect contact, were introduced. As a result, encouraging agreement both with measured melt-penetration lengths and freezing modes of UO2 and metals was obtained. Furthermore, in order to reinforce the developed model, a semi-empirical correlation to predict the supercooling temperature was found. The developed model with the new correlation reproduced both stainless steel freezing and alumina freezing.

Experimental Studies and Empirical Models for the Transient Self-Leveling Behavior in Debris Bed

Studies on the self-leveling behavior in debris beds are crucial in the assessment of core-disruptive accidents (CDAs) that could occur in sodium-cooled fast reactors (SFRs). To clarify this behavior, a series of experiments have been performed in which nitrogen gas has been percolated uniformly through a particle bed. In these experiments, solid particles and water contained in a rectangular tank simulate respectively fuel debris and coolant. Based on the data obtained, an empirical model was developed to describe the transient variation in the bed inclination angle during the self-leveling process. Good agreement has been obtained between calculated and experimental values. Verification of the model has been confirmed through detailed analysis of the effects of experimental parameters such as particle size, particle density, and gas flow rate. Its applicability to extended conditions was further discussed by performing modeling simulations and comparing results against experimental data obtained from a larger-scale experimental system that employed a conventional boiling method. With further improvements, the model will be tested under more realistic reactor conditions and is expected to benefit future analyses and simulations of CDAs in SFRs.

CHARACTERISTICS OF SELF-LEVELING BEHAVIOR OF DEBRIS BEDS IN A SERIES OF EXPERIMENTS

During a hypothetical core-disruptive accident (CDA) in a sodium-cooled fast reactor (SFR), degraded core materials can form roughly conically-shaped debris beds over the core-support structure and/or in the lower inlet plenum of the reactor vessel from rapid quenching and fragmentation of the core material pool. However, coolant boiling may ultimately lead to leveling of the debris bed, which is crucial to the relocation of the molten core and heat-removal capability of the debris bed. To clarify the mechanisms underlying this self-leveling behavior, a large number of experiments were performed within a variety of conditions in recent years, under the constructive collaboration between the Japan Atomic Energy Agency (JAEA) and Kyushu University (Japan). The present contribution synthesizes and gives detailed comparative analyses of those experiments. Effects of various experimental parameters that may have potential influence on the leveling process, such as boiling mode, particle size, particle density, particle shape, bubbling rate, water depth and column geometry, were investigated, thus giving a large palette of favorable data for the better understanding of CDAs, and improved verifications of computer models developed in advanced fast reactor safety analysis codes.

Thermophysical Properties of Lead-Bismuth Eutectic Alloy in Reactor Safety Analyses

A consistent set of thermophysical properties of a lead-bismuth eutectic (LBE) alloy was developed for use in safety analyses of lead-alloy-cooled fast reactor systems. The vapor and liquid thermodynamic states of LBE were modeled up to and above the critical point based on a van-der-Waals type of equation. We assumed that LBE vapor is composed of monatomic lead and bismuth and diatomic, bismuth components, and that liquid LBE is a non-ideal mixture of lead and bismuth. Recommended equations were also presented for the transport properties and surface tension of liquid LBE.

Transient Heat Transfer Characteristics Between Molten Fuel and Steel with Steel Boiling in the CABRI-TPA2 Test

Abstract In the TPA2 test of the CABRI-RAFT program, which is part of a fast reactor safety study, fuel-to-steel heat transfer characteristics within a molten fuel/steel mixture system have been investigated. This test was performed in the French CABRI reactor and used a test capsule that contained fresh 12.3%-enriched UO2 pellets with embedded stainless steel balls. Following a preheating phase, the capsule was subjected to a transient overpower that resulted in fuel melting and steel vaporization. The observed steel vapor pressure buildup was quite low, which suggested the presence of a mechanism that significantly reduced the fuel-to-steel heat transfer. A detailed experimental data evaluation by SIMMER-III led to one possible interpretation that the steel vaporization at the surface of the steel ball blanketed the steel from the molten fuel.

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.

Experimental study of bubble behavior in a two-dimensional particle bed with high solid holdup

In a core disruptive accident of a fast breeder reactor, the post accident heat removal is crucial to achieve in-vessel core retention. Therefore, a series of experiments on bubble behavior in a particle bed was performed to clarify three-phase flow dynamics in debris bed, which is essential in heat-removal capability, under coolant boiling conditions. Although in the past several experiments have been carried out in the gas-liquid-solid system to investigate the bubble dynamics, most of them were under lower solid holdup (≤ 0.5), where the solid-phase influence may be not so important as much as the liquid phase. While for this study, the solid holdup is much higher (> 0.55) where the particle-bubble interaction may be dominated. The current experiment was conducted in a 2D tank with the dimensions of 300 mm height, 200 mm width and 10 mm gap thickness. Water was used as liquid phase, while bubbles were generated by injecting nitrogen gas from the bottom of the tank. Various experimental parameters were taken, including different particle bed height (from 30 mm to 200 mm), various particle diameter (from 0.4 mm to 6 mm), different particle type (acrylic, glass, alumina and zirconia beads), and different nitrogen gas flow rate (around 1.75 ml/min and 2.7 ml/min). By using digital image analysis method, three kinds of bubble rise behavior were observed under current experimental conditions and confirmed by the quantitative detailed analysis of bubble rise properties including bubble departure frequency and bubble departure size. This experiment is expected in the future to provide appropriate quantitative data for the analysis and verification of SIMMER-III, an advanced fast reactor safety analysis code.Copyright