• Energy Research
• 7. Clean energy close
• 11. Sustainability close
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• Annals of Nuclear Energy close

Abstract In case of a core melt-down accident in a light water nuclear reactor, hydrogen is produced during reactor core degradation and released into the reactor building. This subsequently creates a combustion hazard. A local ignition of the combustible mixture may generate standing flames or initially slow propagating flames. Depending on geometry, mixture composition and turbulence level, the flame can accelerate or be quenched after a certain distance. The loads generated by the combustion process (increase of the containment atmosphere pressure and temperature) may threaten the integrity of the containment building and of internal walls and equipment. Turbulent deflagration flames may generate high pressure pulses, temperature peaks, shock waves and large pressure gradients which could severely damage specific containment components, internal walls and/or safety equipment. The evaluation of such loads requires validated codes which can be used with a high level of confidence. Currently, turbulence and steam effect on flame acceleration, flame deceleration and flame quenching mechanisms are not well reproduced by combustion models usually implemented in safety tools and further model enhancement and validation are still needed. For this purpose, two hydrogen deflagration benchmark exercises have been organised in the framework of the SARNET network. The first benchmark was focused on turbulence effect on flame propagation. For this purpose, three tests performed in the ENACCEF facility were considered. They concern vertical flame propagation in an initially homogenous mixture with 13 vol.% hydrogen content and different geometrical configurations. Three blockage ratios of 0, 0.33 and 0.6 were considered to generate different levels of turbulence. The second benchmark objective was the investigation of the diluting effect on flame propagation. Thus, three tests performed in the ENACCEF facility using the same blockage ratio of 0.63 and three different initial gas compositions (with 10, 20 and 30 vol.% diluents) have been considered. Since ENACCEF runs at ambient temperature, a surrogate to steam was used consisting of a mixture of 0.6He + 0.4CO 2 on molar basis. This paper aims to present the benchmarks conclusions regarding the ability of LP and CFD combustion models to predict the effect of turbulence and diluent on flame propagation.

This study explores heat transfer behaviour in mixed convection of a fluid with internal heat generation, a situation found in chemical and nuclear engineering contexts. Computational fluid dynamics is used to simulate laminar ascending mixed convection flow of a heat-generating fluid in a vertical cylinder with uniformly cooled wall, based on a liquid nuclear fuel concept. A new non-dimensional parameter, the IHG-flux number Ω, is developed to express the balance of axial convection versus radial conduction heat transfer. It was found that heat transfer behaviour depends on this parameter, and it can be used as the transition criterion to categorise the simulated results into three distinct heat transfer regimes. A heat transfer correlation using Ω was also developed for Regime I with small values of Ω, where convection and conduction effects are balanced in a stable flow. In Regime II at intermediate Ω, stronger convection gives rise to flow instability. In Regime III with large Ω, convection dominates and the temperature profile inverts so that the maximum temperature occurs at the wall, while instability remains likely.

Abstract A CANDU-6 reactor has 380 fuel channels of a pressure tube type, which provides an independent flow passage, and each pressure tube contains twelve fuel bundles horizontally. The CHF of a CANDU fuel bundle in a horizontal fuel channel is one of the important parameters determining the thermalhydraulic safety margin as well as the trip set point of the Regional Overpower Protection (ROP) system. Hence, the CHF enhancement of a CANDU fuel bundle has been an issue for a long time and can be affected by the geometric configuration of the fuel elements as well as several appendages such as the end-plates, bearing pads, and spacers attached to the fuel elements. It is known that the first CHFs of a standard 37-element fuel bundle preferably occur at the peripheral subchannels of the center rod, owing to the relative small flow area or high flow resistance under high flow conditions or the normal operating conditions of a CANDU reactor. An increase of the inner ring radius of a standard 37-element fuel bundle without any changes in the fuel bundle or element sizes is introduced to enhance the CHF in consideration of the CHF characteristics of a standard 37-element fuel bundle for a CANDU reactor. Subchannel analysis techniques using the ASSERT-PV code were applied to investigate the local CHF characteristics according to the inner ring radius variation for the original diameter of the pressure tube. It was found that the modification of the inner ring radius is very effective in enhancing the dryout power of the fuel bundle through an enthalpy re-distribution of the subchannels and change in the local locations of the first CHF occurrences.

Abstract The flow accelerated corrosion (FAC) phenomenon is critical phenomenon that undermine the safety of nuclear power plant. The problems of FAC are not only induces pipe’s thinning but occurs widely in nuclear power plants. Therefore, precise diagnosis of FAC phenomenon is important but the FAC phenomenon has a complicated mechanism. As a result, accurately diagnosing the FAC phenomenon is difficult (Kain, 2014). To overcome these difficulties, we proposed a methodology utilizing vibration data and deep learning to diagnose FAC induced thinning. To minimize the effect of outliers, Cook’s distance and the Hilbert transform were used. When there was a significant difference in the pipes’ thickness, the support vector machine (SVM), convolutional neural network (CNN), and long-short term memory (LSTM) network all showed good results. However, when the difference in pipes’ thickness was subtle and thinning was non-uniform, only LSTM successfully diagnosed the pipe’s condition.

Abstract The decommissioning of nuclear facilities must be accomplished according to its structural conditions and radiological characteristics. An effective risk analysis requires basic knowledge about possible risks, characteristics of potential hazards, and comprehensive understanding of the associated cause-effect relationships within a decommissioning for nuclear facilities. The hazards associated with a decommissioning plan are important not only because they may be a direct cause of harm to workers but also because their occurrence may, indirectly, result in increased radiological and non-radiological hazards. Workers need to be protected by eliminating or reducing the radiological and non-radiological hazards that may arise during routine decommissioning activities as well as during accidents. Therefore, to prepare the safety assessment for decommissioning of nuclear facilities, the radiological and non-radiological hazards should be systematically identified and classified. With a semantic differential method of screening factor and risk perception factor, the radiological and non-radiological hazards are screened and identified.

Abstract A systematic core design method is developed to design Gd-bearing fuel assembly having two types of Gd rods, low-wt%-Gd rod and high-wt%-Gd rod. The purpose of the method is to lower the critical boron concentration (CBC) of a preliminary core loading pattern, and consequently to achieve more negative or less positive moderator temperature coefficient (MTC). The proposed core design method is a process of solving a non-linear programming problem stated with a system of equations. In this method, both the ratio of the number of low-wt%-Gd rod to the number of high-wt%-Gd rod ( r ) and the assembly average Gd wt% ( w ) are the solution variables of the system of equations. The target function is the amount of soluble boron concentration reduction, Δ CBC , which is correlated with the reactivity change, Δ k FA , per Gd-bearing fuel assembly by a quadratic reactivity equation. The coefficients of the quadratic equations are calculated prior to the determination of Gd-bearing fuel assembly pattern, using the least square method. The constraints required to determine ( r , w ) are physically realizable Gd rods pattern, Δ k i close to Δ k FA derived from Δ CBC , etc. An objective function, min f ∑ i ( Δ k FA - Δ k i ) , enables a final loading pattern to reach a target CBC. This design methodology is applied to APR 1400. Total six cases with various target CBCs are investigated to validate the proposed method. CASMO-3/MASTER calculations with new design assemblies produce lower CBCs at BOC than target CBCs keeping maximum pin power below the safety limit, and thus show more negative MTC.

Abstract This paper aims at the design review and controllability assessment of a secondary system of SMART-ITL which is an integral effect test facility for an integral type reactor of SMART. The design concepts and operating characteristics of the SMART-ITL secondary system were described, and compared with a prototype SMART plant. Then, based on an experimental investigation, the thermal-hydraulic behaviors in the secondary system were analyzed in three sequential steps: a steady-state analysis of the thermodynamic aspect, a dynamic analysis using the lumped heat transfer model, and heat-up simulations as a case study. The key parameter determining system pressure was identified in the steady-state analysis. A dynamic model was established with several physical assumptions, and validated using relevant experimental results. Then, a case study was conducted to identify the effect of recirculation water during heat-up simulations using the validated dynamic model. The results indicated that the design of the present secondary system was appropriate as a part of the integral effect test facility of SMART-ITL, and was being properly operated to match the design intent. Furthermore, the design concept of the present SMART-ITL secondary system is expected to be widely utilized for various separate and integral effects tests for pressurized water reactor designs.

Abstract Extracting features for early failure detection in rotating machinery of nuclear power plants (NPPs) is difficult because in the early stages of failure the impact on the vibration signals is weak. To improve early fault detection in rotating machinery, a fault feature extraction method based on the combination of parameter-adaptive Variational Mode Decomposition (VMD) and Teager energy operator (TEO) is proposed in this paper. Firstly, we introduce the maximum weighted kurtosis index (WKI) as the objective function, and the Artificial Bee Colony (ABC) is used to optimize the VMD parameters. Then, the optimized VMD is used to decompose the vibration signal into multiple intrinsic mode functions (IMFs). Finally, TEO is used to demodulate the sensitive mode with the maximum WKI and identify the fault frequencies. Simulation and experiment show that the early fault features in vibration signals can be effectively extracted by the proposed method, and the comparisons with other three methods highlight the advantages of the proposed method.