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[en] Highlights: • Analysis of SPERT-III Experiments. • Reactivity Initiated Accident Analysis. • Nuclear Data Uncertainty Propagation in CASMO down to 3-D core transient simulation. • Transient Analysis of RIA using SIMULATE-3K. • Uncertainty in Dynamical parameters. - Abstract: This research aims at validating the SIMULATE-3K code and complementing the results with the quantification of nuclear data uncertainties against the Special Power Excursion Reactor Test III (SPERT-III) experiments. To that aim, the SHARK-X methodology, under development at PSI, for the propagation of nuclear data uncertainties in CASMO5 2-D lattice calculations to 3-D core transient simulations is applied for the analysis of a SPERT-III super-prompt critical test conducted at cold startup conditions. Concerning transient results, both total power and reactivity show a good agreement with the measurements at the initial phase of power excursion, while a slight discrepancy is obtained at the final phase of the transient. The estimated uncertainties regarding both steady-state parameters such as k-eff and static reactivity worth, as well as dynamical quantities such as power pulse width and enthalpies are presented. Results show non-negligible sensitivity upon the employed nuclear data library. The uncertainty quantification results show relatively small biases for k-eff and reactivity. The uncertainty in peak power is around 3%, while it is negligible for the time to peak power and the pulse width. The time evolution of the standard deviation and skewness of the total power showed special shapes with relatively high maximum values. In addition, the uncertainty due to nuclear data in the two important safety parameters, i.e. maximum nodal fuel temperature and enthalpy reaches maximum value around 2% and 10%, respectively.
[en] The supercritical water reactor CSR1000 is selected for the study. A parallel channel flow transient flow distribution module is developed, which is used for solving unsteady nonlinear equations. The incorporated programs of SCAC-CSR1000 are executed on normal and abnormal operating conditions. The analysis shows that: 1. Transient flow distribution can incorporate parallel channel flow calculation, with an error less than 0.1%; 2. After a total loss of coolant flow, the flow of each channel shows a downward trend; 3. In the event of introducing a traffic accident, the first coolant flow shows an increasing trend.
[en] Highlights: • Pressure losses and interfacial drag in coarse-particle beds are investigated. • Pressure losses in coarse-particle beds show a down-up trend with gas velocity. • Decrease of pressure losses in coarse-particle beds results from interfacial drag. • Proportions of interfacial drag in pressure losses rise with particle size. • Interfacial drag should be considered in analysis models for coarse-particle beds. - Abstract: Motivated by reducing the uncertainties in coolability assessment of debris bed with coarse particles formed during the severe accident of core melting in light water reactors, this paper reports the experimental investigations on the pressure losses for two phase flow through packed porous beds with coarse particles as well as the effects of the interfacial drag between fluid phases on debris coolability analysis. The experiments are carried out on the test facility of DEBECO-LT (DEbris BEd COolability-Low Temperature), which was designed to investigate single/two-phase flow in porous beds. The spherical particles of 4 mm, 6 mm and 8 mm in diameter are packed in the cylindrical test section with the inner diameter of 120 mm and the height of 600 mm. The pressure losses are measured when fluids flow through the porous packed beds. The experimental data with predictable results by models are analyzed comparatively, and the influences of interfacial drag on two-phase flow pressure losses are discussed. The results show that, for the packed beds with coarse particles, the pressure losses curves show a descent-ascent tendency along with the fluid velocity, and this tendency will become more significant in packed bed with bigger size particles or under lower liquid velocity. The decrease for pressure losses in beds with coarse particles resulted from the effects of interfacial drag, and the rates of the interfacial drag to two-phase pressure losses are rising significantly in beds with greater size of particles, which indicates the interfacial drag plays an important and nonnegligible role in total pressure losses. Therefore, for the packed beds with coarse particles, the interfacial drag should be considered specifically and nonnegligible in debris coolability analysis models.
[en] Highlights: • Characteristics of pool sloshing behavior investigated experimentally. • Increasing gas-injection pressure leads to limited sloshing intensity. • Significant influence of initial water depth, gas-injection duration and nozzle size on sloshing behavior observed. • Both initial water depth and nozzle size can have impact on critical gas pressure. - Abstract: Studies on the pool sloshing behavior are important for the improved evaluation of energetic potential of a large whole-core-scale molten fuel pool that might be formed during a Core Disruptive Accident (CDA) of Sodium-cooled Fast Reactors (SFR). Motivated to understand the characteristics of this behavior, in this study a series of simulated experiments was conducted by injecting nitrogen gas into a Two-Dimensional (2D) rectangular water pool through a nozzle positioned at the center of pool bottom. To achieve a comprehensive understanding, experimental parameters, including nitrogen gas pressure (∼4.2 bar), initial water depth (∼60 cm), gas injection duration (0.06–0.1 s) along with the nozzle size (10–50 mm), were varied. Through detailed analyses, it is found that under current range of conditions, all the experimental parameters employed are confirmable to have remarkable impact on the sloshing characteristics (e.g. maximum elevation of water level at the pool center and peripheries). The performed analyses also suggest that possibly due to a diminished residence time of the injected-gas in the pool, a limited sloshing intensity is observable as the gas-injection pressure increases. Evidence and fundamental data from our work will be utilized for the empirical-model development as well as the analyses and verifications of future SFR severe accident analysis codes in China.
[en] Highlights: • Solidification effects of Bi–Tin melt jets into water coolant were investigated. • Melt-jet modes were organized in a two-dimensional mode diagram. • Melt-jet modes were separated in the jet-breakup and solidification regime. • A scaling analysis using modified Stephan number and Weber number was proposed. • Meltjet was found to completely fragment during CDAs in SFRs. - Abstract: We present a scaling analysis approach that estimates the melt-jet behavior during the core disruptive accidents (CDAs) in sodium-cooled fast reactors (SFRs). For the scaling analysis, we injected a simulant metal into water coolant and observed the solidification effects on a melt jet. Melt-jet behavior occurred in the water coolant, and the debris settled in the test section. From the experimental results, we categorized the melt-jet behaviors into four modes. We then constructed a scaling analysis based on the Weber number and the modified Stefan number, and mapped the solidification effects on the melt jet in a two-dimensional mode diagram. Using the scaling analysis, we estimated the melt-jet mode in the previous experiments conducted under the large-scaled prototypic conditions, and that under the postulated condition during CDAs in SFRs.
[en] Some comments are made about the effective worth of a control plate in a reactor which has sustained structural damage. The damaged core is modelled using a statistical description based on the source-sink method. Detailed quantitative results are presented for a plate reactor and a major conclusion is that, on average, randomness reduces the effective worth of a control element
[en] Highlights: • New experiments performed by lessening known uncertainties. • Consistent impact of melt temperature, water volume and subcooling reproduced. • Overall rationale of previous statistical results at JAEA confirmed. • Effect of water-lump shape and melt depth newly verified. • Most probable reason for limited pressurization confirmed to be isolation effect of vapor bubbles at interface. - Abstract: Studies on local fuel-coolant interaction in a molten pool are important for the improved evaluation of severe accidents that might be encountered for sodium-cooled fast reactors. To clarify the characteristics of this interaction, several years ago a series of simulated experiments was conducted at the Japan Atomic Energy Agency (JAEA) by delivering a given quantity of water into a molten pool formed with a low-melting-point alloy. In this work, motivated by acquiring more reliable database and knowledge, an improved experimental system has been developed recently at the Sun Yat-sen University. Experimental results from various conditions including much difference in water volume, melt temperature, water subcooling, initial water-lump shape along with melt depth, are discussed. It is found that the water volume, melt temperature, initial water-lump shape and melt depth are observable to have remarkable impact on the pressure-buildup characteristics while the role of water subcooling is less prominent. The observed consistent influence of water volume, melt temperature and water subcooling, as compared to previous studies at JAEA, indicates that despite the existence of some uncertainties, the statistical results recognized at JAEA might be overall valid. Owing to much enriched evidence, the present analyses suggest that the most probable reason leading to the limited pressurization as water volume increases for a given melt and water temperature, should be primarily due to the isolation effect of vapor bubbles at the melt-water interface.
[en] Highlights: • Experimental investigations on molten metal spreading and depositing behaviors on the steel plate were carried out. • The three stages of molten metal spreading behaviors with collision against the plate surface were defined by the observation results. • New scaling relation was developed by focusing on the initial spreading pause of the molten metal droplet. - Abstract: On March 11, 2011, huge earthquake and tsunami attacked Fukushima Daiichi Nuclear Power Plant. After the accident, research on plant decommissioning has become actively worldwide. Several research institutes have performed experiments to investigate methods of identifying the location and spreading/deposition behaviors of molten core debris in the bottom of Primary Containment Vessel (PCV) using Severe Accident (SA) analysis codes. Nevertheless, knowledge of spreading and deposition behaviors of corium is not sufficient, especially phenomena involving collision against the floor surface. In this study, experimental investigations on molten metal spreading and depositing behaviors on the steel plate were carried out. Zinc and copper were utilized for the molten metal samples and spreading behaviors were carefully observed using high speed video camera. Immediately after the collision between falling molten metal and steel surface, initial pause on spreading was observed. New scaling relation based on Dinh et al. (2000) was developed by focusing on the initial spreading pause of the molten metal droplet. Proposed correlation is capable to predicting the spread and deposition of falling molten metal at average error of 18.1%.
[en] Highlights: • Addition FEUMIX module to ASTEC-Na to give it a new spray fire modeling capability. • Validation of ASTEC-Na against a database of historical experiments. • Development of a new correlation for the Sih parameter. • Predictions of containment response during a hypothetical accident. - Abstract: One of the most significant containment-related safety issues during a severe accident at a sodium fast reactor is a spray fire, as the high energy ejection and combustion of sodium can cause a sudden, large spike in containment pressure, and can possibly cause the containment building’s early failure. This paper discusses the incorporation sodium spray fire models into ASTEC-Na, a simulation tool currently under development by IRSN for sodium fast reactor severe accidents. An extensive comparison to past experimental results is included, and a new empirical model that describes the combustion phenomena for large flow sodium spray fires is developed. This paper also extends ASTECNa to a hypothetical core disruptive accident, and discusses a fire confinement suppression technique.
[en] A large-break loss-of-coolant accident (LOCA) was analyzed in the course of the design study for the direct-cycle supercritical-pressure light water reactor (SCLWR). The advantages of SCLWR are a higher thermal efficiency and simpler reactor system than the current light water reactors. A computer code was prepared for the analysis of the blowdown phase from the supercritical pressure. The calculation was connected to the REFLA-TRAC code when the system pressure decreased to around atmospheric pressure. The analyzed accidents are 100, 75, 50 and 25% cold-leg and 100% hot-leg breaks. First, blowdown and heatup phases without an emergency core cooling system (ECCS) were evaluated. A low-pressure coolant injection system (LPCI) was designed to fill the core with water before the cladding (stainless-steel) temperature reached a limit of 1260oC. The LPCI consists of four units, each of which has the capacity 805 kg/s. An automatic depressurization system was designed to release the steam generated in the core in the case of cold-leg breaks and to permit operation of LPCI in the case of LOCAs of less than 100% break. For all cases analyzed, the peak cladding temperatures were lower than the limit when the designed ECCS is implemented. (author)