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[en] Developing automated methods for data collection and analysis that can facilitate nuclear nonproliferation assessment is an important research area with significant consequences for the effective global deployment of nuclear energy. Facility modeling that can integrate and interpret observations collected from monitored facilities in order to ascertain their functional details will be a critical element of these methods. Although improvements are continually sought, existing facility modeling tools can characterize all aspects of reactor operations and the majority of nuclear fuel cycle processing steps, and include algorithms for data processing and interpretation. Assessing nonproliferation status is challenging because observations can come from many sources, including local and remote sensors that monitor facility operations, as well as open sources that provide specific business information about the monitored facilities, and can be of many different types. Although many current facility models are capable of analyzing large amounts of information, they have not been integrated in an analyst-friendly manner. This paper addresses some of these facility modeling capabilities and illustrates how they could be integrated and utilized for nonproliferation analysis. The inverse problem of inferring facility conditions based on collected observations is described, along with a proposed architecture and computer framework for utilizing facility modeling tools. After considering a representative sampling of key facility modeling capabilities, the proposed integration framework is illustrated with several examples.
[en] Highlights: • VLE data of TBP-NPH system at different pressures were measured experimentally. • Wilson and NRTL models were used to correlate the VLE data. • Thermodynamic consistency of the VLE data was carried out using Herington method. • Mathematical model of distillation column was developed and solved numerically. - Abstract: The presented work details the modeling and simulation of distillation column for the separation of tributyl phosphate (TBP) and normal paraffin hydrocarbon (NPH) which is an integral unit in the PUREX solvent regeneration system. The vapor liquid equilibrium data of TBP + NPH system at 5.5 torr and 7.5 torr were experimentally generated using an ebulliometer and the data were subjected to thermodynamic consistency test using Herington's method. The experimental VLE data were correlated using Wilson and non random two-liquid (NRTL) models and the binary interaction parameters regressed from the data are reported. The distillation column simulation was carried out using an algorithm involving separate solver/corrector loops for the MESH variables and the NRTL model was used for predicting the vapor liquid equilibrium of TBP-NPH system. Validity of the developed code was tested by comparing the simulation outputs with the experimental results from PUREX solvent regeneration plant.
[en] Highlights: • Large eddy simulation of the LWR unit cell was made. • LBM and FVM are used as numerical solvers. • LES and VLES turbulence models implemented to LBM are validated. • The effect of mesh size is studied for each turbulence model. • The results of LWR unit cell simulations are compared for each turbulence model. - Abstract: In this paper, turbulent flow simulations of a typical LWR unit-cell are presented. Three subgrid scale turbulence models (WALE, VLES () and VLES ()) have been implemented in open source lattice-Boltzmann code Palabos. The Smagorinsky-Lilly model already exists in the code. The simulations also include Smagorinsky-Lilly model results. The standard wall function implementation is made for all turbulence model simulations. The simulation results are compared with the Hooper's experimental data. Additionally, fine grid ANSYS Fluent simulations are performed using WALE turbulence model. The results show that VLES () and VLES () models give better agreement with experimental results even if the grid is intermediate in size.
[en] Highlights: • Experimental study on blowdown transients of a natural circulation PWR-type SMR in a scaled integral test facility. • The ECCS design was validated by the experiment results that it can keep the water level above the core. • Self-sustained flow rate oscillation was observed, which was induced by the decreasing of the water level. • Experimental results were used to verify the RELAP5 prediction. - Abstract: Experimental study on natural circulation flow instabilities is of great importance for the safety analysis in a PWR-type SMR, especially for accident scenarios such as loss of coolant accident (LOCA) and loss of heat sink accident (LOHS). In this study, an experimental natural circulation facility was built by scaling down from a typical PWR-type SMR. The scaling ratios were derived from non-dimensional field and constitutive equations of the drift flux model. The test facility has a height of 3.44 m with an operating pressure limit of 1.0 MPa. Two kinds of tests, the blowdown test and cold-blowdown test were performed. The blowdown test was designed to simulate the low pressure phase (<1 MPa) of a LOCA event and evaluate the performance of the emergency core cooling system (ECCS). The test results indicate that the ECCS can keep the RPV water level above the core, and the RPV pressure keeps decreasing during the entire accident transients. The cold-blowdown test was designed to simulate the whole transients of the system blowdown at low pressure conditions. The oscillations of the pressure drop, natural circulation rate, void fraction and inlet temperature were observed and analyzed. The cold-blowdown test results were compared with the RELAP5 simulation. Although some discrepancies existed at the initial blowdown phase, the code calculation agreed well with the experiment data.
[en] For the purpose of figuring out the thermal-hydraulic behaviors during bottom reflooding in the narrow rectangular channel of plate-type fuel reactor, experimental apparatus ‘THERMAL’ was established to simulate the bottom reflooding process, with different inlet velocities and initial surface temperatures. The narrow channel was formed between a heating plate made of stainless steel and a heat-proof glass. Thermocouples (TCs) were fixed to measure the solid temperature in vertical and horizontal directions, and a high-speed camera was used to record flow regime near the quench front. Based on experimental results, conclusions can be drawn that quench velocity increases with increasing inlet velocity and decreasing initial solid temperature. The quench temperature, which is strongly affected by the initial wall temperature, is almost independent of the inlet velocity. A new method is utilized to determine rewetting front and quench front from ‘temperature variation speed’ curve. Moreover, an analytical model of quench velocity in narrow rectangular channel is discussed and proposed, which can predict quench velocity more precisely in ‘THERMAL’ facility.
[en] Modeling of dynamic processes in nuclear reactors is carried out, mainly, on the basis of the multigroup diffusion approximation for the neutron flux. The basic model includes a multidimensional set of coupled parabolic equations and ordinary differential equations. Dynamic processes are modelled by a successive change of the reactor states, which are characterized by given coefficients of the equations. In the modal method, the approximate solution is represented as an expansion on the first eigenfunctions of some spectral problem. The numerical-analytical method is based on the use of the dominant time-eigenvalues of a multigroup diffusion model taking into account delayed neutrons. In this work, the application of the modal methodology based on calculation of the dominant eigenvalues and eigenfunctions of α-eigenvalue problem has been tested for the VVER-1000 reactor test model. The last is characterized by the fact that some eigenvalues are complex. Reactor dynamics behavior is simulated for symmetrical and non-symmetrical control rods insertion/withdrawal. The power calculation results obtained with the modal method were compared with the numerical solution of the dynamics problem. A rather good agreement was shown for the problem with single delayed neutron precursor group.
[en] In this paper, experimental investigation of the molten metal jet's colliding and spreading behaviors on the flat steel surface covered with water layer was carried out. High-frequency induction heating system was utilized to produce the molten metal sample and it was released to the wet surface from a fixed elevation. As the molten metal collides against the surface, it rapidly goes through solidification while spreading on the wet surface. High-speed thermo-camera was utilized to measure the molten metal's surface temperature during the spreading transient. Once the molten metal completely solidifies, molten metal's spread area and thickness were measured. From the obtained database, a dimensional analysis was conducted to investigate the key parameters responsible for the molten metal spreading on the wet surface. Based on the key non-dimensional parameters identified in the current analysis, the new empirical correlation was proposed. Its predictive capability was found to be 18.9% in mean absolute relative deviation.
[en] For safety management of nuclear power plants, accurate impact mass estimation of loose parts is very important. A center frequency method based on Hertz's impact theory, a frequency ratio method, and so forth are studied for mass estimation, but it is known that these approaches can hardly provide accurate information on impact response for identifying the impact source. In this paper, a finite element analysis (FEA) model to simulate the propagation behavior of the bending wave, generated by a metal ball impact, was validated by performing a series of impact tests and the corresponding finite element analyses for a flat plate as well as a curved plate (a half scale model of steam generator vessel). Various impact parameters such as amplitude, center frequency, group velocity and attenuation ratio of bending wave acceleration signal were investigated to verify the usefulness of the FEA model. Also, an FEA-based metal sphere signal map was developed, and then blind tests were performed to verify the map. This study provides an accurate simulation method for predicting the metal impact behavior and for building a metal sphere signal map, which can be used to estimate the mass of loose-parts on site in nuclear power plants.
[en] Highlights: • A novel neutron shielding composite is made by natural Szaibelyite resource. • Shielding percentage increases with the adding amounts and thickness of shielding material. • Shielding percentage of all samples is nearly 100% when the thickness is more than 2 cm. • The shielding mechanism of shielding material is analyzed. - Abstract: In this paper, different amounts of Szaibelyite (100, 300, 500, 700 and 900) were used to prepare Szaibelyite/epoxy resin composites for neutron shielding. Microstructures of fracture surfaces of the prepared composites were tested by scanning electron microscope while the neutron shielding properties were determined by Am-Be neutron source. It was found that the shielding percentage of all composites increases with the addition of Szaibelyite as well as the thickness of composites increment. M5 was found to be the best neutron shielding composite, the shielding percentage was nearly 100% when the thickness of M5 chosen to be 1.5 cm. Furthermore, the shielding mechanism of M5 for neutron shielding could be concluded as follows: Firstly, H, O, C, Fe and B-11 isotope contained in M5 slowed down the neutron to thermal neutron, which then absorbed by the B-10 isotope. According to the achieved results, the performance of the new neutron shielding material owns the value of practical application.
[en] In present study, comprehensive experimental examination is performed to investigate the characteristics of packing for the separation of light stable isotopes. This work focused on the effect of liquid redistribution on wall flow reduction. In this research work, 0.005 m Dixon ring packing was constructed and characterized. Experimental tests were conducted in a packed column with a diameter of 0.04 m and various heights up to 1 m. Structural characteristics, pressure drop, dynamic hold-up and redistribution spacing were comprehensively investigated. Dry and wet pressure drop charts were plotted for a wide range of gas superficial velocities corresponding to gas loading factor values of 0.2–1.3 Pa0.5 Height of the packed section between two consecutive redistributors was defined at various liquid superficial velocities. Our results show that compared to larger columns, liquid should be redistributed in smaller spacings when the diameter of packed columns is smaller than 0.05 m. In addition, it is found that efficient redistribution spacing is about 0.3 m in a 0.04 m column. According to our findings, this value decreases the wall flow more than 50%.