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[en] Highlights: • A flux reconstruction method is presented that uses a 3D transport theory form factor. • 3D form factor is a 2D xy-plane component times an approximate 1D z-axis component. • Method is used to simulate travelling flux detector scan (TFD scan) readings. - Abstract: Even with current computing capabilities, detailed full core three-dimensional (3-D) transport calculations are still not practical. However, if we are satisfied with knowing only the average values of spatial flux distributions, the 3-D diffusion solution will constitute the final solution. On the other hand, in reactor design and safety analysis, direct information about the local flux distribution for the heterogeneous assemblies is required to assess the design and determine the safety margins. For this reason, after having solved the full-reactor-core problem, we have to look into the possibilities of recovering in a second step the information on local properties of single heterogeneous assemblies. In particular, the detector readings at detector locations are derived using these global homogenized parameters by applying appropriate numerical methods such as advanced interpolations. In this paper, we propose a method based on flux reconstruction to calculate the simulated detector readings in three-dimensions with high fidelity. Data from detector readings are very important in ensuring optimal reactor operations as well as in detecting any deviations from normal operations. Thus, calculating the detector readings with high fidelity will allow improvements to operating and safety margins. To validate this method, comparisons between detector reading simulation results and measurements from an operating CANDU reactor will be conducted and results will be presented.
[en] Among the new failure modes introduced by computer into safety systems, the process interaction error is the most unpredictable and complicated failure mode, which may cause disastrous consequences. This paper presents safety analysis and constraint detection techniques for process interaction errors among hardware, software, and human processes. Among interaction errors, the most dreadful ones are those that involve run-time misinterpretation from a logic process. We call them the 'semantic interaction errors'. Such abnormal interaction is not adequately emphasized in current research. In our static analysis, we provide a fault tree template focusing on semantic interaction errors by checking conflicting pre-conditions and post-conditions among interacting processes. Thus, far-fetched, but highly risky, interaction scenarios involve interpretation errors can be identified. For run-time monitoring, a range of constraint types is proposed for checking abnormal signs at run time. We extend current constraints to a broader relational level and a global level, considering process/device dependencies and physical conservation rules in order to detect process interaction errors. The proposed techniques can reduce abnormal interactions; they can also be used to assist in safety-case construction.
[en] Highlights: • The characteristics of density wave oscillation in the natural circulation system were analyzed. • The effects of heaving and rolling motions on flow instability in the natural circulation system were analyzed. • The resonance phenomenon between density wave oscillation and ocean motions was found and analyzed. - Abstract: The study of natural circulation is attracting more and more attention of researchers and engineers for the improvement of the system safety. A lot of natural circulation systems are applied to new generation reactors which make reactor systems safer and more compact because there are fewer components and no circulation pump in the natural circulation systems. The residual heat can be removed passively from the reactor by the natural circulation system which prolongs the response time for operators and reduces the potential accident because of human error. With the application of nuclear reactors under the ocean conditions, the study on flow oscillation characteristics in natural circulation system under ocean condition becomes very significant. Usually, flow oscillation is harmful to the operation of natural circulation systems which can affect the cooling capability of natural circulation systems in some extreme situations. In this paper, the classification of flow instability is introduced for systems under ocean conditions. Characteristics of density wave oscillations (DWO) under natural circulation are studied, including oscillation characteristics, stability boundary and subcooling effect. The characteristics of DWO under rolling and heaving motions are compared with those under the normal condition. The effects of rolling and heaving characteristics are analyzed, including maximum rolling angle, rolling period, rolling center location, maximum acceleration and heaving period. The resonance phenomenon is found, and resonant behaviors are analyzed.
[en] Highlights: • Evaluation +98 of WWER-1000 containment behavior against LBLOCA accident. • Simulation of WWER-1000 containment by CONTAIN 2.0 code. • Modeling of WWER-1000 containment by a single model. • Validation of results with Bushehr Nuclear Power Plant’s FSAR. - Abstract: The consequences of sever reactor accident depend greatly on containment safety features and containment performance in retaining radioactive material. The specific type of large LOCA is DECL (Double Ended Cold Leg) break which means a total guillotine type of break in cold leg pipe and is one of the most dangerous accidents in the reactor containment. In this paper, thermal–hydraulic parameters (temperature and pressure) of WWER-1000 (Bushehr Nuclear Power Plant) containment in a DECL accident have been simulated by CONTAIN 2.0 code and a single cell model. The containment has been divided to 23 cells in CONTAIN code but for simplicity only one cell has been considered in modeling. The model has been programmed by MATLAB. The accident has been simulated for a short time (initial 200 s) and all of the results have been compared with Bushehr’s Nuclear Power Plant FSAR
[en] In CANDU (registered) reactor design, the regional overpower protection (ROP) systems protect the reactor against overpower in the fuel which could reduce the safety margin-to-dryout. The increase in fuel power could be caused by a localized power peaking within the core (for example, as a result of a certain reactivity device configuration) or a general increase in the core power level during a slow-loss-of-regulation (SLOR) event. This overpower could lead to fuel sheath dryout. In the CANDU (registered) 600 MW (CANDU 6) design, there are two ROP systems in the core, one for each fast-acting shutdown system. Each ROP system includes a number of fast-responding, self-powered flux detectors suitably distributed throughout the core within vertical and horizontal assemblies. A new methodology for designing the detector layout for the ROP system, called the DETPLASA algorithm, has been developed recently. This method utilizes the simulated annealing (SA) technique to optimize the placement of the detectors in the core. The evaluation of the trip setpoint (TSP) corresponding to each detector layout configuration (i.e., each history in the SA algorithm) is performed probabilistically using the ROVER-F code. In this evaluation, there are uncertainties related to both the detector components (i.e., related to the margin-to-trip) and to the fuel channel components (i.e., related to the margin-to-dryout). In this paper, the importance of these uncertainties on the outcome of the detector layout optimization process is evaluated. Some parametric studies have been performed to quantify the effect of uncertainties on the resulting detector layout. Two types of investigations have been performed. First, a given detector layout will be used to explicitly determine the effect of changing the uncertainty values. In this study, 343 sets of uncertainty values are used to produce the corresponding TSP values. The variation in the TSP values is analyzed. Second, three sets of uncertainty values (a subset of uncertainties from the first study) are used in independent DETPLASA executions. The resulting detector layout configurations will be examined to observe the effect of these uncertainties on the final design. Results from these investigations are presented in this paper.
[en] Highlights: • This article is consistent with the theme of the thermal–hydraulic. • It provides a new mathematical method for the analysis of the factors affecting the CHF (Critical Heat Flux). • Also it has a great potential for research related to CHF. - Abstract: CHF (Critical Heat Flux) is an important nuclear thermal–hydraulic parameter, and it is closely related to the reactor safety. Based on Rough Set Decision Model, this article proposed a new method to seek for the main factors which may affect CHF at low pressure, low flow conditions in narrow channels. Nine condition attributes are selected to build the CHF fault diagnosis model in narrow rectangular channels, and the rules of CHF occurring are obtained. These attributes include pipe diameter, circulating mode and flow stability, etc. The results show that flow instability could cause the CHF reach a minimum and CHF could be improved with the increase of pipe size. Diagnostic results in this method could predict the main influencing factors for CHF very well, and provide a new theoretical approach for the safety analysis of nuclear plants
[en] Highlights: • A transient analysis of an ATLAS 6-in. cold-leg break was carried out with MARS. • The calculated major sequence of events showed good agreement with the measured data. • To investigate ECC bypass, variation of the boron concentration in SI was adopted. • 20–50% of the ECC bypass fraction was calculated. • Boron tracking approach is considered a feasible methodology to quantify ECC bypass. - Abstract: A transient analysis of an ATLAS (Advanced Thermal–Hydraulic Test Loop for Accident Simulation) 6-in. cold-leg break was carried out with the MARS (Multi-dimensional Analysis for Reactor Safety) safety analysis code in the frame of the Domestic Standard Problem exercise in Korea. The calculated major sequence of events of the 6-in. cold-leg break simulation showed good agreement with the measured data. The calculated break mass flowrate was predicted well, whereas accumulated mass of the break outflow was underestimated due to the underestimation of the break mass flowrate in a later phase. The general trends of the collapsed water level were well predicted in the core and the downcomer region. The loop seal clearing phenomena were observed at about 400 s in the 1-A and 2-B intermediate legs in the calculation results which is identical to the experiment. To investigate the emergency core coolant (ECC) bypass phenomena, variation of the boron concentration in safety injection water was adopted. In the loop seal clearing phase of the 6-in. cold-leg break, about 40–50% of the ECC bypass fraction was calculated and after the loop seal clearance, 20–30% of ECC water bypassed, that is, not participating in core cooling directly. The boron tracking approach is considered to be a feasible methodology with which quantify the ECC bypass flow in the early phase of a small-break loss-of-coolant accident (SBLOCA)
[en] Highlights: • Detailed studies of using nanofluids in a PWR reactor are performed. • The Case-Bibi model is utilized to consider the gap conductance. • A considerable enhancement of heat transfer is accomplished. • The critical boric acid and relative power distribution are less impacted. - Abstract: In this paper we investigate the thermal–hydraulic and neutronic attributions of using nanoparticles in the primary cooling system of a VVER-1000 reactor. The coupled analysis of the nanofluid core is performed by using DRAGON, DONJON and a thermal–hydraulic model that solves the governing momentum, energy, and mass equations. The applied approach is validated by comparing the results with the final safety analysis report (FSAR) of the plant. Finally, critical boric acid, relative power distribution, pressure drop, and temperature distributions of fuel, clad and coolant are considered for water/Al2O3 nanofluid. It is observed that low volume fraction of the nanoparticles has a minimum impact in critical boric acid up to about 3% and relative power distribution about 2% at maximum while the heat transfer is enhanced in comparison to pure water
[en] Highlights: • A coupling algorithm was developed by combining DEM with a multi-fluid model. • This method was validated by perform the simulations of gas–solid fluidized beds. • Agreement was obtained between the simulation results and experimental data. - Abstract: Gas–solid fluidization is not only an essential phenomenon in many areas of industry, but is also used to understand particle behavior in a number of research fields. For the safety analysis of core disruptive accidents in liquid-metal fast reactors, a hybrid method is developed by combining the discrete element method with a fluid-dynamics model of the reactor safety analysis code SIMMER-III to reasonably simulate particle transient behavior, as well as the occurring thermal-hydraulic phenomena. As a preliminary validation procedure, the developed hybrid method is applied to simulations of gas–solid two-phase flows. In this study, numerical simulations of two typical gas–solid fluidized bed systems are performed. The particles in the beds are porous alumina of 70 μm diameter and glass of 530 μm diameter, which belong to Geldart groups A and B, respectively. The reasonable agreement between our simulation results and experimental data from the literature demonstrates the fundamental validity of the present simulation method for multiphase flows with large amounts of solid particles
[en] Highlights: • A 14 inference modules based DFLC is designed for 70th order MIMO PHWR system. • Auto tuning of DFLC for PHWR is performed using NMA. • A novel approach is presented to overcome the shortcomings of NMA in tuning the DFLC. • The optimally tuned DFLC is evaluated for robustness and reference tracking capabilities. - Abstract: A Pressurized Heavy Water Reactor (PHWR) is a highly complex and unstable system. Designing a safe, reliable and robust controller with good performance for such a large and complex system is an important control engineering problem. In this work, a Decentralized Fuzzy Logic Controller (DFLC) with 140 input and 70 output membership functions, is designed for a 70th order Multi-Input Multi-Output (MIMO) type PHWR. In order to obtain high performance of the controller, it needs to be tuned optimally, however, it is very challenging task to optimally tune the DFLC with such a large membership functions. Moreover, PHWR is a coupled system which imposes additional limitation in tuning the controller since the output of one PHWR’s zone affects the outputs of other zones. In this work, an application of Nelder–Mead Algorithm (NMA) is presented for auto tuning the DFLC. The NMA performance depends upon objective function and initial points given to the NMA at the start of the tuning process. A novel method for selecting the optimal objective function and initial points for the NMA is also proposed since their selection is another complicated process. Although several objective functions have been proposed by the researchers for use with NMA, this work focuses five common indices (IAE, ISE, ITAE, ITSE and ISTE) as objective functions, which are simple and system independent. Finally, the optimally tuned high-performance DFLC is applied to the PHWR and evaluated by simulating different scenarios. The simulation results show that the controller is efficient, fast and robust and ensures the safety and reliability of the PHWR