Flooded Cavity Research Articles

DAVINCI-SP tests for debris bed formation and spreading in a pool with center conical structure under two-phase condition

In case of severe accidents in light water reactors (LWRs), the coolability of relocated corium from the reactor vessel is a significant safety issue. The failure in cooling and stabilizing the molten core in the containment vessel can threaten the integrity of the containment boundary. In a flooded cavity, a porous corium debris bed is expected to develop on the bottom of the cavity pool due to the melt jet breakup and fragmentation. The geometric configuration of the porous debris bed is important for accurate coolability assessment under accident conditions. Earlier, Kim et al. (2016a, 2016b) presented an analytical model to estimate the spreading of the debris bed in terms of two-phase flow and the debris injection parameters. Here we are testing the effect of conical bottom surface in passively dispersing the debris particle to obtain a levelled bed formation. For this, the previous model for debris bed development is modified to simulate the development of debris bed over conical bottom surface. The existing ‘The Debris Bed Research Apparatus for Validation of the Bubble-Induced Natural Convection Effect Issue (DAVINCI) - Spreading (SP) experimental facility was modified with a center conical structure. Five kilograms of stainless steel simulant debris were injected stepwise from the top of the water level, while air bubbles simulating the two-phase vapor flow were injected from the bottom of the particle catcher plate. The test results from the current study to describe the spreading of the debris bed in terms of two-phase flow, debris injection parameters, and conical bottom surface geometrical parameters are discussed. Subsequently, the model for characteristics length and side slope angle of debris bed were modified using the empirical parameters from the tests. Comparison between the tests results and modified model prediction are presented. Lastly, the effectiveness of the conical bottom surface on particle dispersion is discussed.

Experimental investigation of quenching phenomena in high temperature conical debris bed – Part 1: MONET tests

In case of severe accidents in light water reactors (LWRs), the coolability of relocated corium from the reactor vessel is a significant safety issue. In a flooded cavity, a porous corium debris bed is expected to develop on the bottom of the cavity pool due to the melt jet breakup and fragmentation. The debris bed needs to be quenched by water either flooding from the top or flooding from the bottom until continuous cooling is established. If it is not quenched, this mass will begin to re-melt due to decay heat and leads to Molten Corium Concreate Interaction (MCCI). Most of the studies reported have conservative assumptions of one-dimensional homogeneous particulate beds with top or bottom flooding condition of coolant. Understanding of debris beds quenching with multi-dimensional characteristics are limited. To investigate the quenching behavior of the conical debris bed by means of experiments, a test facility MCCI-mitigatiON through passive cooling Effect Test (MONET) apparatus is established. Specifically, the objective is to understand the mechanism of heat transfer and quench front movement inside the conical heated debris bed during quenching by cooling water supplied from the bottom of the bed. Two different particle bed bottom wall boundary conditions are studied; closed bottom wall and open bottom wall to understand the effect of water ingression from the bottom wall. The particulate debris bed is formed with two different types of particles; Ø 3 mm SS 304 and Ø 3 mm alumina. In addition, the effect on the quench front movement inside the particle bed has been reported by using temperature measurements together with direct visualizations.

Coupled MELCOR/COCOMO analysis on quench of ex-vessel debris beds

The cornerstone of severe accident strategy of Nordic BWRs is to flood the reactor cavity for the long-term coolability of an ex-vessel debris bed. As a prerequisite of the long-term coolability, the hot debris bed formed from fuel coolant interactions (FCI) should be quenched. In the present study, coupling of the MELCOR and COCOMO codes is realized with the aim to analyze the quench process of an ex-vessel debris bed under prototypical condition of a Nordic BWR. In this coupled simulation, MELCOR performs an integral analysis of accident progression, and COCOMO performs the thermal–hydraulic analysis of the debris bed in the flooded cavity. The effective diameter of the particles is investigated. The discussion on the bed’s shape shows a significant effect on the propagation of the quench front, due to different flow patterns. Compared with MELCOR standalone simulation, the coupled simulation predicts earlier cavity pool saturation and containment venting.

Development of a downward-facing nucleate boiling correlation for thermal hydraulics analysis

The concept of in-vessel retention (IVR) by passive external reactor vessel cooling (ERVC) in a flooded cavity during a severe accident has now been recognized as a viable approach to retain the radioactive core melt (i.e., corium) within the reactor vessel. During the process of long-term in-vessel cooling of core melt, the heat flux must remain below the critical heat flux (CHF) level to maintain a regime of nucleate boiling such that the integrity of the reactor pressure vessel (RPV) is not compromised. In this study, steady-state data for downward facing boiling (DFB) obtained under simulated IVR-ERVC conditions in the Subscale Boundary Layer Boiling (SBLB) test facility is critically analyzed based upon which a new DFB correlation is derived mathematically from theoretical considerations with the coefficients resulting from the scaling analysis being determined from the DFB data. The new correlation which adequately predicts the local variation of the nucleate boiling heat flux along the outer surface of the RPV under IVR-ERVC conditions can be used to describe the long-term cooling behavior of the corium within the RPV of advanced nuclear power plants.

Sensitivity and uncertainty analyses of ex-vessel molten core cooling in a flooded cavity during a severe accident

Sensitivity and uncertainty analyses of molten core cooling under the ex-vessel phase of a severe accident of a light water reactor was performed with the COOLAP-I (COOLability Analysis Program-I) model, a parametric model considering one-dimensional heat transfer of a porous debris particle bed, covering a broad range of phenomena from the melt jet release to the long-term cooling process. COOLAP-I improved the previous version by including particle generation by the fuel-coolant interaction (FCI), and internal heat generation by the decay heat. With nine representative input parameters, an uncertainty analysis using Latin hypercube sampling (LHS) method with 300 samples were conducted, and the cooling characteristics such as total enthalpy, maximum temperature, decay heat ratio, and cake (a lump of connected particles) fraction, were examined for the elapsed time of up to 50 h. This analysis demonstrates the impacts of the water pool depth, the jet breakup-related parameters, and the accumulation area of the debris particles on the cake formation by particle agglomeration and analyzes the long-term coolability of debris particle bed in plant scale conditions.

Development of an ex-vessel corium debris bed with two-phase natural convection in a flooded cavity

During severe accidents of light water reactors (LWRs), the coolability of relocated corium from the reactor vessel is a significant safety issue and a threat to the integrity of containment. With a flooded cavity, a porous debris bed is expected to develop on the bottom of the pool due to breakup and fragmentation of the melt jet. As part of the coolability assessment under accident conditions, the geometrical configuration of the debris bed is important. The Debris Bed Research Apparatus for Validation of the Bubble-Induced Natural Convection Effect Issue (DAVINCI) experimental apparatus facility was constructed to investigate the formation of debris beds under the influence of a two-phase flow induced by steam generation due to the decay heat of the debris bed. Using this system, five kilograms of stainless steel simulant debris were injected from the top of the water level, while air bubbles simulating the vapor flow were injected from the bottom of the particle catcher plate. The airflow rate was determined based on the quantity of settled debris, which will form a heat source due to the decay of corium. The radial distribution of the settled debris was examined using a ‘gap–tooth’ approach. Based on the experimental results of this study, an analytical model was established to describe the spreading of the debris bed in terms of two-phase flow and the debris injection parameters. This model was then used to analyze the formation of debris beds at the reactor scale. A sensitivity analysis was carried out based on key accident parameters, including the quantity of corium melt, cavity flooding level, volumetric decay heat rate, and the size of the melt jet.

Ice‐melt rates by steam condensation during explosive subglacial eruptions

AbstractSubglacial volcanism melts cavities in the overlying ice. These cavities may be flooded with meltwater or they may be fully or partially drained. We quantify, for the first time, heat transfer rates by condensation of steam on the walls and roof of a fully or partially drained subglacial eruption cavity. Our calculations indicate that heat fluxes of up to 1 MW m−2may be obtained when the bulk vapor in the cavity is in free convection. This is considerably smaller than heat fluxes inferred from ice penetration rates in recent subglacial eruptions. Forcing of the convection by momentum transfer from an eruption jet may allow heat fluxes of up to 2 MW m−2, consistent with values inferred for the Gjálp 1996 subglacial eruption. Vapor‐dominated cavities in which vapor‐liquid equilibrium is maintained have thermal dynamic responses that are an order of magnitude faster than the equivalent flooded cavities.

Fluidized soil–cement mixes for backfilling of flooded cavities

This paper deals with the development of flowable soil–cement mixes to be used for filling underground cavities flooded by sulphate contaminated water. The research started from the need to backfill a former gypsum mine which was excavated with the chamber-pillar technique. The dissolution of gypsum and anhydrite, consequent to the flow of water inside the abandoned tunnels, and the possible collapse of the pillars puts at serious risk of collapse the area above the mine. The paper describes the steps which led to the definition of a cement–soil mixture characterized by rheological properties adequate to be introduced in the tunnels from the surface and substitute the sulphate contaminated water. The relationships between the composition of the flowable soil–cement mixture and its properties in the fresh and hardened state are discussed.

Transient Pool Boiling Critical Heat Flux of FC-72 Under Saturated Conditions

In this study, the steady-state and transient critical heat fluxes (CHFs) in pool boiling were measured on 1.0 mm diameter horizontal cylinders of gold and platinum heaters under saturated conditions due to transient heat inputs, Q{sub 0}exp(t/t), in a pool of Fluorinert FC-72. Heaters were heated by electric current with the periods, t, ranged from 10 ms to 20 s, and the pressures ranged from atmospheric up to around 1.2 MPa. The steady-state CHFs measured are dependent on pressure and almost agree with the values obtained by Kutateladze's correlation based on hydrodynamic instability (HI) model. It was considered that the boiling inception and the direct transition during the steady-state period occur by the pre-pressure of {approx}1.2 MPa. The trend of typical transient CHFs were clearly divided into the first, second, and third groups for long, short, and intermediate periods, respectively. The direct transition processes to film boiling without nucleate boiling for the short periods obtained from both heaters were confirmed due to the heterogeneous spontaneous nucleation (HSN) in flooded cavities on the cylinder surface. The empirical correlations to express each of corresponding CHFs measured on both heaters for the short periods are presented in this paper. (authors)

Limiting Factors for External Reactor Vessel Cooling

The method of external reactor vessel cooling (ERVC) that involves flooding of the reactor cavity during a severe accident has been considered a viable means for in-vessel retention (IVR). For high-power reactors, however, there are some limiting factors that might adversely affect the feasibility of using ERVC as a means for IVR. In this paper, the key limiting factors for ERVC have been identified and critically discussed. These factors include the choking limit for steam venting (CLSV) through the bottleneck of the vessel/insulation structure, the critical heat flux (CHF) for downward-facing boiling on the vessel outer surface, and the two-phase flow instabilities in the natural circulation loop within the flooded cavity. To enhance ERVC, it is necessary to eliminate or relax these limiting factors. Accordingly, methods to enhance ERVC and thus improve margins for IVR have been proposed and demonstrated, using the APR1400 as an example. The strategy is based on using two distinctly different methods to enhance ERVC. One involves the use of an enhanced vessel/insulation design to facilitate steam venting through the bottleneck of the annular channel. The other involves the use of an appropriate vessel coating to promote downward-facing boiling. It is found that the use of an enhanced vessel/insulation design with bottleneck enlargement could greatly facilitate the process of steam venting through the bottleneck region as well as streamline the resulting two-phase motions in the annular channel. By selecting a suitable enhanced vessel/insulation design, not only the CLSV but also the CHF limits could be significantly increased. In addition, the problem associated with two-phase flow instabilities and flow-induced mechanical vibration could be minimized. It is also found that the use of vessel coatings made of microporous metallic layers could greatly facilitate downward-facing boiling on the vessel outer surface. With vessel coatings, the local CHF limits at different angular locations of the vessel outer surface could be enhanced by ~1.2 to 2 times the CHF compared with a plain vessel without coatings. The CHF enhancement could be attributed to the structure of the porous coating itself and the capillary action it induced. The matrix of cavities and voids within the coating effectively trap vapor, which serve as active nucleation sites. These sites in turn are fed with liquid flowing through the interconnected channels. The pores on the surface of the porous coating serve as flow inlets for liquid supply to the heating surface, leading to appreciable enhancement in downward-facing boiling heat transfer and the local CHF limits. Results of the present study suggest that by utilizing an enhanced vessel/insulation design with vessel coating, it is possible to significantly enhance the CLSV and the CHF limits as well as minimize the two-phase flow instability problems, thus substantially increasing the margin for IVR.

兵庫県旧生野鉱山中央竪坑水位の潮汐変化の定量的解釈

Excavated areas such as the ore veins, tunnels and shafts of the Old Ikuno Mine, an abandoned mine in western Japan, can be regarded as comprising a huge reservoir of groundwater extending over several square kilometers to a depth of about 1km. In 1988, tidal fluctuations with an amplitude on the order of 1mm were observed in the water level of the Chuo shaft in this mine. The vertical length of the Chuo shaft is 370m, about two-third of which is flooded. Observations were curtailed until 1997, when this well-tide was again observed. The persistence of this phenomenon over a nine-year period implies that the well-tide in the Chuo shaft is an intrinsic feature of this huge underground reservoir. To understand the well-tide in the Chuo shaft, we compared the tidal fluctuations of its water level with crustal movement measured in the same area. We found that (1) the tidal response of the water level to crustal strain is frequency dependent, and (2) the tidal fluctuation exhibits a phase advance. These are quantitatively explained by a finite-cavity model with a drained condition. The estimated volume of the flooded cavity is on the order of 106m3, which is consistent with the excavated volume of the Old Ikuno Mine.

Ex-vessel boiling experiments: laboratory- and reactor-scale testing of the flooded cavity concept for in-vessel core retention Part II: Reactor-scale boiling experiments of the flooded cavity concept for in-vessel core retention

This paper summarizes the results of a reactor-scale ex-vessel boiling experiment for assessing the flooded cavity design of the heavy water new production reactor. The simulated reactor vessel has a cylindrical diameter of 3.7 m and a torispherical bottom head. Boiling outside the reactor vessel was found to be subcooled nucleate boiling. The subcooling mainly results from the gravity head, which in turn results from flooding the side of the reactor vessel. The boiling process exhibits a cyclic pattern with four distinct phases: direct liquid-solid contact, bubble nucleation and growth, coalescence, and vapor mass dispersion. The results show that, under prototypic heat load and heat flux distributions, the flooded cavity will be effective for in-vessel core retention in the heavy water new production reactor. The results also demonstrate that the heat dissipation requirement for in-vessel core retention, for the central region of the lower head of an AP-600 advanced light water reactor, can be met with the flooded cavity design.

Ex-vessel boiling experiments: laboratory- and reactor-scale testing of the flooded cavity concept for in-vessel core retention Part I: Observation of quenching of downward-facing surfaces

This paper presents the results of a series of quenching experiments to examine the boiling curve of relative large downward-facing surfaces. The test masses are 61 cm in diameter and the downward-facing surface is either flat or curved. The work is motivated by the need to assess the ex-vessel boiling process for in-vessel core retention. The critical heat flux is found to be approximately 50 W cm −2. The nucleate boiling regime and the critical heat flux regime are found to be characterized by cyclic two-phase flow patterns.

In-vessel retention of corium at the Loviisa plant

In-vessel retention of corium has been approved to be part of the severe accident management strategy for IVO's Loviisa plant. The approach selected takes advantage of the unique features of the plant such as a low power density, a reactor pressure vessel (RPV) without penetrations at the bottom, and ice-condenser containment which ensures a flooded cavity in all risk significant sequences. The thermal analyses, which are supported by an experimental program, demonstrate that, in Loviisa, the molten corium on the lower head of the RPV is externally coolable with wide margins. This paper summarizes the approach, the thermal analyses and the plant modifications being implemented.

Boiling phenomenon due to quasi-steadily and rapidly increasing heat inputs in LN 2 and LHe I

Dynamic boiling processes, including the transition from a single-phase non-boiling regime to film boiling caused by exponentially increasing heat inputs, Q 0e t τ for a wide range of periods and pressures on horizontal wires in LN 2 and LHe I were investigated. The main problem is that there are no active cavities on the wire surfaces for initial boiling in the liquids. The heat transfer processes due to increasing heat inputs with increasing rates ranging from quasi-steady to rapidly increasing ones in LN 2 were classified into three types for the pressures. The dynamic boiling processes in LHe I due to rapidly increasing heat inputs at the pressures tested here correspond to Type 3 processes including semi-direct transitions in LN 2 at pressures higher than about 1 MPa. The lower limit temperatures of boiling initiation on the wire surfaces for initial boiling in liquids at pressures due to quasi-steadily increasing heat inputs are clearly lower than the homogeneous spontaneous nucleation temperatures corresponding to these pressures. Liquid superheat close to the solid surface in LHe I was evaluated from the value of the wire surface temperature, taking off the temperature drop due to Kapitza resistance. The initial boiling temperatures due to quasi-steady heat inputs at pressures in saturated LN 2 and LHe I agreed with the values derived from the theoretical model based on the heterogeneous spontaneous nucleation in flooded cavities on the solid surface.

Predicting the onset of nucleate boiling in wavy free-falling turbulent liquid films

Experiments are performed to develop a fundamental understanding of boiling incipience in wavy free-falling turbulent liquid films. Incipience conditions are measured and correlated for water and a fluorocarbon (FC-72) liquid. Incipience in water films is influenced by turbulent eddies and, to a larger extent, by interfacial waves. A new approach to predicting incipience in water and other non-wetting fluids is presented. This approach utilizes physical parameters of commonly accepted incipience models and provides a means of correcting these models for the effects of turbulent eddies and roll waves. This study also demonstrates some unique incipience characteristics of fluorocarbon films. The weak surface tension forces of FC-72 allow droplets and liquid streams to break off the crests of incoming roll waves prior to, and during nucleate boiling. The low contact angle of FC-72 allows the liquid to penetrate deep inside wall cavities. Thus incipience from these flooded cavities requires much higher wall superheat than predicted from incipience models.