• Nuclear Engineering and Technology
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• 제34권2호
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• pp.117-131
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• 2002

Core disruptive accidents have been investigated at Korea Atomic Energy Research Institute(KAERI) as part of the work to demonstrate the inherent and ultimate safety of conceptual design of the Korea Advanced Liquid Metal Reactor(KALIMER), a 150 MWe pool- type sodium cooled prototype fast reactor that uses U-Pu-Zr metallic fuel. In this study, a simple method and associated computer program, SCHAMBETA, was developed using a modified Bethe-Tait method to simulate the kinetics and thermodynamic behavior of a homogeneous spherical core over the period of the super-prompt critical power excursion induced by the ramp reactivity insertion. Calculations of the energy release during excursions in the sodium-voided core of the KALIMER were subsequently performed using the SCHAMBETA code for various reactivity insertion rates up to 100 S/s, which has been widely considered to be the upper limit of ramp rates due to fuel compaction. Benchmark calculations were made to compare with the results of more detailed analysis for core meltdown energetics of the oxide fuelled fast reactor. A set of parametric studies were also performed to investigate the sensitivity of the results on the various thermodynamics and reactor parameters.

• 한국전산유체공학회지
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• 제21권3호
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• pp.8-14
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• 2016

A numerical analysis is conducted to estimate the core expansion and the energy behaviors induced by a core disruptive accident in a sodium-cooled fast reactor. The numerical formulation based on underwater explosion theory is carried out to simulate the core explosion inside the reactor vessel. The transient pressure, temperature and expansion of the core are examined by solving the equation of state and nonlinear governing equation of momentum conservation in one-dimensional spherical coordinates. The energy balance inside the computation domain is examined during the core expansion process. Heat transfer between the core and the sodium coolant, and the bubble rise during the expansion process are briefly investigated.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2004년도 추계학술발표회 발표논문집
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• pp.535-536
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• 2004

• Nuclear Engineering and Technology
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• 제51권1호
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• pp.73-83
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• 2019

It is important to clarify the mechanisms of molten-fuel-pool sloshing behavior that might be encountered during a core disruptive accident of sodium-cooled fast reactors. In this study, motivated by acquiring some evidence for understanding the characteristics of this behavior at more realistic conditions, a number of experiments are newly performed by injecting nitrogen gas into a water pool with the accumulation of solid particles. To achieve comprehensive understanding, various parameters including particle bed height, particle size, density, shape, gas pressure along with the gas-injection duration, were employed. It is found that due to the different interaction mechanisms between solid particles and the gas bubble injected, three kinds of regimes, termed respectively as the bubble-impulsion dominant regime, the transitional regime and the bed-inertia dominant regime, could be identified. The performed analyses also suggest that under present conditions, all our experimental parameters employed can have noticeable impact on the regime transition and resultant sloshing intensity (e.g. maximum elevation of water level at pool peripheries). Knowledge and fundamental data from this work will be used for the future verifications of fast reactor severe accident codes in China.

• Nuclear Engineering and Technology
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• 제43권1호
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• pp.63-74
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• 2011

The 600 MWe, pool-type, sodium-cooled, metallic fuel loaded KALIMER-600 (Korea Advanced LiquId MEtal Reactor, 600 MWe) has been conceptually designed with an emphasis on safety by self-regulating (inherent/intrinsic) negative reactivity feedback in the core. Its inherent safety under the ATWS (Anticipated Transient Without Scram) events was demonstrated in an earlier study. Initiating events of an HCDA (Hypothetical Core Disruptive Accident), however, also need to be analyzed for assessment of the margins in the current design. In this study, a hypothetical triple-fault accident, ULOF (Unprotected Loss Of Flow) with a reduced pump halving time, is investigated as an initiator of a core disruptive accident. A ULOF with insufficient primary pump inertia may cause core sodium boiling due to a power-to-flow mismatch. If the positive sodium reactivity resulting from this boiling is not compensated for by other intrinsic negative reactivity feedbacks, the resulting core power burst would challenge the fuel integrity. The present study focuses on determination of the limit of the pump inertia for assuring inherent reactivity feedback and behavior of the core after sodium boiling as well. Transient analyses are performed with the safety analysis code SSC-K, which now incorporates a new sodium boiling model. The results show that a halving time of more than 6.0 s does not allow sodium boiling even with very conservative assumptions. Boiling takes place for a halving time of 1.8 s, and its behavior can be predicted reasonably by the SSC-K.

• Nuclear Engineering and Technology
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• 제54권6호
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• pp.2120-2134
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• 2022

A hypothetical core destructive accident (HCDA) has received widespread attention as one of the most serious accidents in sodium-cooled fast reactors. This study combined recent advantages in numerical methods to realize realistic modeling of the complex fluid-structure interactions during HCDAs in a full-scale sodium-cooled fast reactor. The multi-material arbitrary Lagrangian-Eulerian method is used to describe the fluid-structure interactions inside the container. Both the structural deformations and plug rises occurring during HCDAs are evaluated. Two levels of expansion energy are considered with two different reactor models. The simulation results show that the container remains intact during an accident with small deformations. The plug on the top of the container rises to an acceptable level after the sealing between the it and its support is destroyed. The methodology established in this study provides a reliable approach for evaluating the safety feature of a container design.

• Nuclear Engineering and Technology
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• 제45권3호
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• pp.323-334
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• 2013

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.

• Nuclear Engineering and Technology
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• 제50권1호
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• pp.43-53
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• 2018

It is important to clarify the characteristics of flow regimes underlying the debris bed formation behavior that might be encountered in core disruptive accidents of sodium-cooled fast reactors. Although in our previous publications, by applying dimensional analysis technique, an empirical model, with its reasonability confirmed over a variety of parametric conditions, has been successfully developed to predict the regime transition and final bed geometry formed, so far this model is restricted to predictions of debris mixtures composed of spherical particles. Focusing on this aspect, in this study a new series of experiments using nonspherical particles have been conducted. Based on the knowledge and data obtained, an extension scheme is suggested with the purpose of extending the base model to cover the particle-shape influence. Through detailed analyses and given our current range of experimental conditions, it is found that, by coupling the base model with this scheme, respectable agreement between experiments and model predictions for the regime transition can be achieved for both spherical and nonspherical particles. Knowledge and evidence from our work might be utilized for the future improvement of design of an in-vessel core catcher as well as the development and verification of sodium-cooled fast reactor severe accident analysis codes in China.

• Nuclear Engineering and Technology
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• 제54권2호
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• pp.589-600
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• 2022

Investigations on the molten-pool sloshing behavior are of essential value for improving nuclear safety evaluation of Core Disruptive Accidents (CDA) that would be possibly encountered for Sodium-cooled Fast Reactors (SFR). This paper is aimed at synthesizing the knowledge from our recent studies on molten-pool sloshing behavior with solid particles conducted at the Sun Yat-sen University. To better visualize and clarify the mechanism and characteristics of sloshing induced by local Fuel-Coolant Interaction (FCI), experiments were performed with various parameters by injecting nitrogen gas into a 2-dimensional liquid pool with accumulated solid particles. It was confirmed that under different particle-bed conditions, three representative flow regimes (i.e. the bubble-impulsion dominant, transitional and bed-inertia dominant regimes) are identifiable. Aimed at predicting the regime transitions during sloshing process, a predictive empirical model along with a regime map was proposed on the basis of experiments using single-sized spherical solid particles, and then was extended for covering more complex particle conditions (e.g. non-spherical, mixed-sized and mixed-density spherical particle conditions). To obtain more comprehensive understandings and verify the applicability and reliability of the predictive model under more realistic conditions (e.g. large-scale 3-dimensional condition), further experimental and modeling studies are also being prepared under other more complicated actual conditions.

• Nuclear Engineering and Technology
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• 제53권12호
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• pp.3930-3943
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• 2021

To improve the capability of the SAS4A code, which simulates the initiating phase of core disruptive accidents for MOX-fueled Sodium-cooled Fast Reactors (SFRs), the authors have investigated in detail the physical phenomena under unprotected loss-of-flow (ULOF) conditions in a previous paper (Kawada and Suzuki, 2020) [1]. As the conclusion of the last article, fuel stub motion, in which the residual fuel pellets would move toward the core central region after fuel pin disruption, was identified as one of the key phenomena to be appropriately simulated for the initiating phase of ULOF. In the present paper, based on the analysis of the experimental data, the behaviors related to the stub motion were evaluated and quantified by the author from scratch. A simple model describing fuel stub motion, which was not modeled in the previous SAS4A code, was newly proposed. The applicability of the proposed model was validated through a series of analyses for the CABRI experiments, by which the stub motion would be represented with reasonable conservativeness for the reactivity evaluation of disrupted core.