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A new empirical model is proposed for estimating the amount of volatile iodine in an aqueous phase. The volatile iodine concentration is estimated for highly irradiated CsI solutions in which the p...

An analysis of the responses of the containment during a station blackout accident is performed for the APR1400 nuclear power plant using MELCOR 2.1. The analysis results show that the containment failure occurs at about 84.14 h. Prior to the failure of the reactor vessel, the containment pressure increases slowly. Then, a rapid increase of the containment pressure occurs when a large amount of hot molten corium is discharged from the reactor pressure vessel to the cavity. The molten corium concrete interaction (MCCI) is arrested when water is flooded over a molten corium in the cavity. The boiling of water in the cavity causes a fast increase in the containment pressure. During the early phase of the accident, a large amount of steam is condensed inside the containment due to the presence of the heat structures. This results in a mitigation of a containment pressure increase. During the late phase, the containment pressure increases gradually due to the addition of steam and gases from an MCCI an...

Some advanced reactors adapt the in-vessel corium retention concept by cooing the outside wall of the reactor vessel in severe accidents. If a reactor vessel failure happens in this case, the molte...

International audience; The Great East Japan earthquake occurred on March 11, 2011, at 1446, and the subsequent tsunami led Tokyo Electric Power Company's (TEPCO's) Fukushima Daiichi Nuclear Power Station (NPS) beyond a design-basis accident. After the accident, the Japanese government and TEPCO compiled a roadmap toward an early resolution to the accident including, among the main activities, the employment and improvement of existing severe accident (SA) computer codes. In the member countries of the Organisation for Economic Co-operation and Development/Nuclear Energy Agency (OECD/NEA), SA codes were developed after the accident at Three Mile Island Unit 2 and widely employed to assess NPS status in the postulated SA conditions. Therefore, working plans have been set up with the country members of the OECD/NEA to apply existing SA codes to analyze the accidents at the Fukushima Daiichi NPS Units 1, 2, and 3 and support the decommissioning, constituting an international program named Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Station (BSAF). The objectives of the BSAF project are to analyze the accident progression of Fukushima Daiichi NPS, to raise the understanding of SA phenomena, to contribute to the improvement of the methods and models of the SA codes, and to define the status of the distribution of debris in the reactor pressure vessels and primary containment vessels for decommissioning. The present technical paper summarizes the achievements obtained through a comparison of the results, emphasizing the portions of the accident where all the participants reached a common consensus and identifying still open questions where future work should be directed. Consensus exists on the current condition of Unit 1, where a large fraction of the fuel is assumed to have relocated ex-vessel. On the other hand, larger uncertainties exist for Units 2 and 3, where in-vessel and ex-vessel scenarios produce a reasonable prediction of the accident progression.

Steam explosion experiments are performed at various modes of melt water interaction configuration using prototypic corium melt. The tests are performed to simulate both melt water interaction in a...

A scaling method is proposed for the design of a reduced-scale experimental facility for testing the performance of a newly proposed filtered containment venting system (FCVS). A full-height facili...

The ablation kinetics of special concrete, which has been developed as one of the candidate protecting materials for the EU-APR1400 ex-vessel core catcher, was investigated experimentally. Metallic...

A deflagration to detonation transition (DDT) occurrence is one of the most important issues concerning safety during severe accidents in nuclear power plants because it can damage the integrity of the containment. It is possible to arrest the acceleration of a flame which can cause DDT by installing quenching meshes between the compartments. To evaluate the applicability of a quenching mesh to nuclear power plants, it requires a means to evaluate a flame arrest of a quenching mesh under a given combustion condition. The flame-quenching models developed by previous researchers were derived to fit the experimental geometry and to consider various thermal boundary conditions from a flame to the mesh wall. Flame-quenching tests were carried out at the 10% hydrogen concentration in a dry air by changing atmospheric pressure to 2.2 bar as the initial pressure. The quenching criterion of a quenching mesh with a 0.3 mm gap distance for hydrogen–air mixtures is established by using the experimental data. The flam...

Weaknesses of the current Severe Accident Management Guideline (SAMG) in handling the cooling of a molten core are discussed, and three improvements for the SAMG are presented. It is suggested that instrumentation to detect either a breach of the reactor vessel or a discharge of corium into the reactor cavity is essential to effectively perform the SAMG. A detailed analysis for a specific plant is necessary to make a decision as to whether preflooding or postflooding should be initiated for effective molten core cooling. Also, an optimal choice of depressurization capacity not only would significantly delay failure of the reactor vessel but also would increase the coolability margin of the molten corium in a reactor cavity. Analyses using the MELCOR computer code were performed for the Ulchin Units 1 and 2 and Kori Unit 1 nuclear power plants to illustrate the effectiveness of the proposed improvements in cooling of the molten core in the reactor cavity, where in-vessel retention of molten corium by prefl...

The SuperCritical Water-cooled Reactor (SCWR) is one of the candidates for the fourth-generation nuclear power plant, and it uses light water as a coolant. Heat transfer between a fuel assembly and...

To investigate the fundamental mechanism behind the recent experimental observation that the composition of a material considerably affects the strength of a steam explosion, physical and chemical analyses for the fast-quenched particles of a prototypic corium were performed. Six cases including fully oxidized and partially oxidized corium were selected for the study, in which the melt composition was changed, while the other initial and boundary conditions of the molten fuel and water interaction tests, such as the melt temperature, amount of water, and free fall height, were maintained the same. The proposition that the inner structure and solidification behavior of particles due to the existence of a mushy phase are responsible for the strong influence of the material composition on the strength of a steam explosion was examined from the results of physical and chemical analysis. The results of the present analysis are supportive of the proposed argument.

The TROI tests were analyzed in view of the particle size responses for various types of fuel-coolant interactions. This can provide an understanding about the relationship among the initial conditions, mixing, and explosion. First, several findings from the TROI experiments were considered. The results of the fuel-coolant interactions (FCIs) were dependent on the composition of the corium, the water depth, and the water area in the TROI experiments. Then, the difference between the explosive FCI and nonexplosive FCI was defined by comparing the final particle size. This analysis indicates that the explosive FCI resulted in a large amount of fine particles and a small amount of big particles. With this, the mixing size of the particles to participate in the steam explosion and the fine particle size produced from the steam explosion could be defined in the TROI test. And then, the parametric effects on the particle size were analyzed using the nonexplosive TROI tests. We note that the explosive test results cannot provide information on the mixing process. This analysis on the particle size response indicates that the explosive system includes large-sized particles to participate in the steam explosion, but the nonexplosive system includes less large-sized particles and more fine-sized particles. These particle size responses should be considered during a reactor safety analysis because the particle size will be an important parameter for explaining a steam explosion occurrence or steam explosion strength.

To simulate a fuel and coolant interaction phenomenon during a postulated severe accident in a nuclear reactor, a series of experiments were performed using a partially oxidized corium, which is a mixture of UO 2 , ZrO 2, Zr, and stainless steel. The composition of the melt was chosen such that a separation of the oxidic liquid from the metallic liquid occurred due to the existence of a miscibility gap. A melting and solidifying experiment and two fuel and coolant interaction experiments to explore the possibility of an energetic steam explosion were performed in the TROI facility. The placement of a metal-rich layer consisting of U, Fe, and ZrO 2 beneath the oxidic corium layer due to the existence of a miscibility gap was observed in the melting and solidifying experiment. An energetic steam explosion with a propagation of the dynamic pressure wave was observed in one test out of the two tests. The physical and chemical analyses were performed or the corium particles collected after the experiments. It is shown that U, Zr, and Fe formed a heterogeneous mixture and the morphology was in irregular shape with many pores at nonuniform sizes. In the case of nonenergetic interaction, where the melt temperature was lower than the energetic case, the mean particle size was bigger than that of the energetic case, and the melt-water interaction resulted in a substantial amount of hydrogen gas generation, while the amount of hydrogen gas generation was negligible in the case with an energetic steam explosion.

Triggered steam explosion experiments have been carried out in the TROI facilities to investigate the energetics of the steam explosions. Two types of corium melt were used as a melt. One was eutec...

Heat transfer test facility, SPHINX (Supercritical Pressure Heat Transfer Investigation for NeXt Generation), was constructed at KAERI (Korea Atomic Energy Research Institute) for an investigation of the thermal-hydraulic behaviors of supercritical CO 2 at the various geometries of the test section. The test data will be used for the reactor core design of the SCWR (Supercritical Water-cooled Reactor). As a working fluid, CO 2 was selected to make use of the low critical pressure and temperature of CO 2 compared with water. An experimental study was carried out in the SPHINX to investigate the characteristics of heat transfer and pressure drop at a vertical single tube with an inside diameter of 4.4 mm in case of an upward flow of supercritical CO 2 . The heat and mass fluxes were varied at a given pressure. The mass flux was in the range of 400∼1,200kg/m 2 s and the heat flux was chosen up to 150kW/m 2 . The selected pressures were 7.75, 8.12, and 8.85MPa. A heat transfer deterioration occurred at the lower mass fluxes. The experimental heat transfer coefficients were compared with the ones predicted by several existing correlations. The standard deviation was about 20% for each correlation and an apparent discrepancy was not found among the correlations. The major components of the pressure drop were a gravitational pressure drop and a frictional pressure drop. The frictional pressure drop increases as the mass flux and heat flux increase.

The goals for hydrogen control in nuclear power plants are to design countermeasures that allow operators to avoid deflagration-to-detonation transition (DDT) and to ensure the survivability of equipment. These goals could be achieved by using a quenching mesh. Flame arrest tests are carried out using a quenching mesh with a 0.3-mm gap distance. When the quenching mesh is installed between compartments, the quenching mesh plays a role in flame quenching below 1.8 bars of the initial pressure and less than ~1.6 m/s of the flame velocity. Therefore, if the quenching mesh is properly installed in the containment, the flame could be arrested within the mesh boundary, resulting in the prevention of DDT and the survivability of equipment. Flame-quenching criteria are suggested using the expansion ratio, the initial air pressure, and the flame velocity.

The paper discusses the results of steam explosion experiments of TROI-13, TROI-14, and TROI-15, which were performed under the research program named “Test for Real cOrium Interaction with water (TROI).” TROI-13 and TROI-14 used corium, which is a mixture of UO2 and ZrO2 at a 70:30wt%, while TROI-15 used ZrO2. These three cases resulted in spontaneous steam explosions. It is an important observation from the aspect of the explosivity of prototypic material, as it was not observed in the previous experiments. Various aspects of the test results including cold crucible melting, hydrogen generation, melt temperature measurement, debris morphology, size distribution of the debris, dynamic pressure, dynamic force, shape of the melt jet, location of the trigger, response of the vessel pressure, and the response of the water pool temperature are discussed. The potential contributors to the explosivity of corium are suggested.

The quenching distance of hydrogen gas was experimentally investigated by considering the effects of the initial pressure and steam addition. The quenching distance decreases with the initial pressure and there is a little increase with the addition of steam. Performance tests have been carried out to check the applicability of quenching mesh for the purpose of arresting hydrogen flame propagation during a severe accident in nuclear power plants. The experimental facility for the performance test of the quenching mesh consisted of a model compartment, a visualization system and an ignition system. Dimensions of the single model compartment were 300×300×300 mm. Three-compartments are connected in parallel. The quenching mesh is located between the first and second compartments. It was observed that the flame from the first compartment where the ignition starts does not propagate to the second compartment. The quenching mesh played a role of preventing flame propagation.

An analysis is presented of the integral behavior of the external cooling of a reactor vessel by natural circulation during a severe accident to investigate the feasibility of the in-vessel retention strategy for a high-power reactor by using the RELAP5/MOD3 computer code. It is shown that two-phase flow instability phenomena, including natural-circulation oscillation and density wave oscillations, affect the local thermal margin at the reactor vessel wall. The heat load on the reactor vessel is simplified as a uniform heat flux load of 600 kW/m{sup 2} in the base case. A sensitivity study for the effect of the inlet K factor, nonuniform heat flux distribution, inlet flow area, and subcooling of the pool water is performed to evaluate the local thermal margin. The results of the analysis show that natural-circulation cooling is marginal at this level of heat flux. It also clearly indicates that a system level of analysis for two-phase natural circulation, including the sensitivity study on the design parameters, is necessary to ensure successful implementation of the external cooling.