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[en] Highlights: • Adopting thermal-hydraulic phenomena (list of 116 phenomena – published). • Exploiting capabilities of existing numerical tools & existing experimental database. • Procedure to plan new-prioritized research in nuclear thermal-hydraulics. - Abstract: The difficulty in predicting locally and globally the transient evolution of two-phase or multiphase flows in complex systems is well recognized in nuclear thermal-hydraulics. Large efforts involving the expenditure of huge resources during the last three decades in previous century brought to the creation of giant databases (e.g. including experimental data and results of computer code calculations) and to the perception that the safety of nuclear reactors is guaranteed notwithstanding residual areas of unawareness. Nowadays, thousands of scientists continued to generate progress in the area having available much lower resources: more and more dead-ends for established research outcomes are experienced; the progress in knowledge resembles the slow expansion of a swamp rather than the fast moving of a river. In this paper a procedure is proposed to identify directions for research in nuclear thermal-hydraulics which are consistent with the needs in nuclear reactor safety. Two pillars for the procedure are constituted by the characterization of phenomena and by the application of qualified computational tools. Decision makers and scientists may prioritize research in areas where large impacts upon design and safety issues are identified in advance.
[en] Highlights: • Thermal responses of a scaled down suppression pool are studied numerically. • Steam injection is simulated using direct contact condensation method and effective source method. • Comparing with experiment, direct contact condensation method produces better results. • Reasons for different thermal status are analyzed by flow field patterns. - Abstract: Thermal status or temperature distribution in terms of mixing or stratification inside a suppression pool (SP) is decisive for containment pressure and hence reactor safety. Numerical models were developed in this study to predict the temperature distribution inside a scaled-down SP with steam injected through a horizontal pipe. Regarding the modeling of injected steam, two methods were compared. Effective source (ES) method, which considers the effective heat and momentum source instead of calculating condensing steam directly, can predict the stratified thermal status but failed to reproduce the mixing thermal status. The reason may be attributed to the lack of an appropriate momentum source model. Direct contact condensation (DCC) method, which calculates the condensation process directly, can give agreeable results with experiments and successfully distinguish between mixing or stratification thermal status. In addition, the DCC method can also predict the transition between mixing and stratification. However, moment of the transition was found to deviate from measurement. It was assumed the measurement and proper modeling of the gas-liquid interface structures is necessary for improvement of the simulation. Besides, reasons for the above thermal statuses were analyzed by fluid velocity fields, and it was shown that stratification can be avoided if the injection location is moved downside. The numerical methods proposed can be used for design and improvement of thermal features of SP.
[en] Highlights: • Conjugating: • ALARA. • BEPU. • Extended safety margin. • Independent assessment. - Abstract: The paper aims at fixing bases for possible strengthening of current Nuclear Reactor Safety (NRS) and safety analysis: this is done by combining the logical frameworks connected with the terms As-Low-As-Reasonably-Achievable (ALARA), Best-Estimate-Plus-Uncertainty (BEPU), Extended-Safety-Margin (E-SM) and Independent-Assessment (IA). ALARA is an early principle in Nuclear Reactor Safety: designers and operators must do their best to minimize doses to the humans. BEPU is an approach in Accident Analysis, part of NRS: one may state that BEPU implies the best use of computational tools to determine the safety of nuclear installations. Then, ALARA may be seen at the origin of BEPU. Furthermore, BEPU (and BEPU elements like V & V, Scaling, procedures of code application and code coupling, etc.) can be extended to all analytical parts of the Final Safety Analysis Report (FSAR). This brings to BEPU-FSAR. Safety Margins constitute an established concept in NRS: a few dozen SM values must be calculated in current safety analyses and demonstrated to be acceptable. The concept can be extended to everything part of the design, the operation and the environment for a Nuclear Power Plant (NPP) Unit, thus forming the E-SM. Here ‘the environment’ includes the personnel in charge of activities connected with the NPP. The E-SM implies the formulation of some ten-thousands safety margins values, which shall correspond to a similar number of monitored variables. IA is an early requirement in NRS: data ownership and system complexity prevented so far a comprehensive application of the requirement. IA analyses conflict with industry policies to keep proprietary data. IA based BEPU-FSAR analyses are essential to finalize the E-SM design. The implementation of the idea in the paper brings to an additional safety barrier for existing and future nuclear reactors which may reduce the probability of core melt, presumably at an affordable cost for the industry.
[en] Highlights: • Numerical stability of RELAP5 is improved by additional differential terms. • Numerical accuracy of RELAP5 is enhanced by TVD flux-limiter scheme. • An alternative interphase friction relation is implemented for LOCA analysis. • Numerical examples and experiments prove the improvement measures. - Abstract: The nuclear reactor safety system code RELAP5 decomposes complex flow system of a nuclear reactor into a series of one-dimensional control volumes connected by flow junctions, and solves a set of two-phase two-fluid equations to predict the nuclear reactor system behavior. In spite of its extensive applications, there indeed exist many numerical shortcomings in RELAP5 and it is desirable to constantly improve its numerical performance. In the present work, the numerical performance improvement to RELAP5/MOD3 is carried out from the aspects of numerical stability, high-resolution and alternative constitutive relations. For the enhancement of numerical stability, the virtual mass term is replaced and an interfacial pressure term is added in the phase momentum equations of RELAP5 to make all the characteristic roots real, thus improve the model’s hyperbolicity. In addition, the second-order Minmod TVD flux-limiter scheme replaces the original first-order upwind scheme for advection terms to reduce the numerical diffusion. Furthermore, an alternative interphase friction relation is substituted for the built-in model in RELAP5 to calculate the interphase drag for bubbly/slug flow in the vertical bundle channels. The performance improvement measures work reasonably well, as indicated by the simulation of selected numerical examples and the Bethsy 6.2TC integral effect experiment which corresponds to an intermediate break Pressurized Water Reactor Loss of Coolant Accident (PWR LOCA).
[en] Highlights: • Analytical characteristic analysis of two-phase two-fluid six-equation model. • A Roe-type explicit numerical solver in solving two-phase two-fluid six-equation model. • The numerical solver is capable of simulating both discontinuous benchmark problems and practical engineering problems. • Realistic equation of state, realistic flow regimes, and closure correlations were used. - Abstract: In the nuclear industry, a method based on a staggered grid is used in two-phase flow system codes such as RELAP, TRAC, and CATHARE. Solving the two-phase two-fluid model with this method is complicated. The objective of this article is to develop a new solver, which is mathematically consistent and algebraically simpler than existing codes. The extension of existing shock-capturing upwind schemes for single-phase flows is our way. A numerical solver with a Roe-type numerical flux is formulated based on a very well-structured Jacobian matrix. We formulate the Jacobian matrix with arbitrary equation of state and simplify the Jacobian matrix to a simple and structured form with the help of a few auxiliary variables, e.g. isentropic speed of sound. Because the Jacobian matrix is very structured, the characteristic polynomial of the Jacobian matrix is simple and suitable for analytical analysis. Results from the characteristic analysis of the two-phase system are consistent with well-known facts, such as the ill-posedness of the basic two-phase two-fluid model which assumes all pressure terms are equal. An explicit numerical solver, with a Roe-type numerical flux, is constructed based on the characteristic analysis. A critical feature of the method is that the formulation does not depend on the form of equation of state and the method is applicable to realistic two-phase problems. We demonstrate solver performance based on three two-phase benchmark problems: two-phase shock-tube problem, faucet flow problem, and Christensen boiling pipe problem. The solutions are in excellent agreement with analytical solutions and numerical solutions from a system code. The new solver provides essential framework for developing a more accurate and robust solver for realistic reactor safety analysis. However, improvements on the new solver is necessary for achieving a high-order accuracy and increasing the robustness.
[en] Highlights: • A post-dryout heat transfer correlation for developing regime was validated over a wide range of flow conditions. • The dryout location is estimated using the CHF look-up table. • Good prediction results are obtained over wider pressure conditions. - Abstract: The present paper deals with the validation of a post-dryout heat transfer correlation, which has been proposed by Nguyen and Moon (2015), to show its extended applicability to developing post-dryout region. It is also shown that the well-known film-boiling look-up table (LUT) method currently having limited application to fully-developed flow can be improved to extend its applicability to the developing post-dryout region. In order to show that the correlation can be utilized for nuclear reactor safety analysis, eight different sets of post-dryout data covering flow conditions both in both large break loss-of-coolant accident (LB LOCA) and prolonged station blackout (SBO) scenarios have been assessed. It turns out that the correlation predicts well the measured wall temperature with a total average error of 3.81% and a root-mean-square error of 13.46% in case of given CHF (dryout) condition. The prediction accuracy of the correlation is strongly influenced by how well the local vapor temperature and the CHF locations can be predicted.
[en] Highlights: • PSA analyzes vulnerabilities of complex systems and quantify risk. • We present a Condition-Based PSA for augmenting PSA capabilities. • CB-PSA makes use of the information made available by sensors and/or inspections. • We demonstrate the CB-PSA on a spontaneous Steam Generator Tube Rupture accident. • The novel approach overcomes the results of a conventional PSA. - Abstract: Condition-Based Probabilistic Safety Assessment (CB-PSA) makes use of the information made available during operation by sensors and/or inspections on the state of components and systems. This allows specializing the PSA to the conditions of the components and systems, reducing the uncertainty on the risk measures quantified. In this paper, we demonstrate the CB-PSA with reference to a spontaneous Steam Generator Tube Rupture (SGTR) accident scenario in a Pressurized Water Reactor (PWR). Results show that the updated risk measures are capable of reflecting the actual state of the SG in the tailored risk evaluation.
[en] Highlights: • A model for spray cooling using an Euler-Euler two-fluid approach is presented. • Different simulations show the importance to observe condensation/evaporation. • All simulations are compared to experimental data of THAI HD-31-SE. • The monodisperse spray droplet approach is extended to a polydisperse spray. • Plots show the heating process and the condensation area. - Abstract: CFD (Computational Fluid Dynamics) simulation of spray is a challenging task in the field of nuclear reactor safety. In the current publication a CFD model for spray cooling is presented, which is able to predict heat and mass transfer between cold droplets and a hot humid air gas atmosphere. The model, which is implemented via user defined functions in the commercial CFD code ANSYS CFX 16.1, enables the simulation of spray cooling physics with an Euler-Euler two-fluid approach. A comparison of simulations with mono- and polydisperse spray configurations shows the relevance to consider droplet size distributions within spray. For validation, the pressure and temperature transients of the experiment THAI HD-31-SE (Thermal-hydraulics, Hydrogen, Aerosols and Iodine) are used. During the experiment, a cold spray is injected into a hot humid air gas atmosphere, which leads to a cooling effect inside of the model containment. A full three dimensional geometrical mesh of THAI is used for all simulations. Simulation results indicate a good agreement with experimental data for the polydisperse spray configuration.
[en] Highlight• Reflood experiments have been performed for small and medium break LOCAs. • Intact and deformed rod bundles were used for the experiments. • Parametric effects of various parameters were analyzed. • The coolability was enhanced in the deformed rod bundle. • MARS-KS code showed good prediction of the experimental results. Reflood experiments at elevated pressures for small and medium break loss-of-coolant accidents were performed to investigate the influence of fuel deformation on the peak cladding temperature and local quenching behaviors. The typical experimental parameters such as initial water level, power, reflood rate, system pressure, initial wall temperature, and fluid temperature were varied to cover a wide range of flow conditions under a medium and high pressure. The experimental results showed remarkable differences in the peak cladding temperature and local quenching behaviors between the intact and deformed rod bundles due to the flow blockage effect. The present results implied that the fuel rod deformation can enhance the coolability of fuel rods by the flow blockage in the present fully blocked channel. Any bypass in a partially blocked channel may change those results due to a crossflow between intact and deformed rods. The experimental results were then used to validate the system analysis code, Multi-dimensional Analysis of Reactor Safety-KINS Standard (MARS-KS). The MARS-KS code showed good prediction of the present experimental data for most of the flow conditions.
[en] Highlights: • The wall-to-vapor convective heat transfer in partially blocked rod bundles is investigated. • The COBRA-TF code underestimated local heat transfer downstream of flow blockage. • A new correlation is developed to accurately describe the effect of flow blockage. • Determination of reattachment point and maximum Nusselt number are critical for the new correlation. - Abstract: Coolability of the partially blocked core in a large break loss-of-coolant accident (LB LOCA) is one of the most important thermal-hydraulic concerns for nuclear reactor safety analysis. During blowdown phase and early stage of reflooding phase in the LB LOCA, the prevailing wall-to-vapor convective heat transfer plays an important role on the decay heat removal process. Experiments on single-phase convective heat transfer to vapor were conducted in 5 × 5 heater rod bundles containing 3 × 3 ballooned rods of 90% flow blockage ratio with consideration of fuel relocation phenomenon. The obtained experimental data were used to assess the single-phase heat transfer enhancement models of the COBRA-TF code. The assessment results showed underprediction of local heat transfer downstream of the flow blockage. Therefore, a new correlation has been proposed to improve the prediction capability of the conventional models by more accurately describing the flow blockage effect. The new correlation predicted the local heat transfer satisfactorily within a 20% discrepancy of the experimental data for various kinds of flow blockage configurations.