Annals of Nuclear Energy最新文献

In order to investigate the influence of guide vane structure on hydraulic performance of reactor coolant pump (RCP) under coastdown condition, the guide vane structure was modified to form staggered guide vane by adding splitter plates that staggered flow channel across two layers. Numerical methods were used to compare the hydraulic performance characteristics of staggered guide vane and original guide vane in RCP. From the results, the external characteristics of RCP slightly decreased by staggered guide vane under coastdown condition. As the coastdown flow rate decreased, the difference in head and distribution of turbulent kinetic energy between two models diminished. The static pressure at impeller outlet and volute tongue increased by staggered guide vane, while distribution of high pressure inside guide vane of original model changed under coastdown condition. The area of high turbulent kinetic energy expanded due to staggered guide vane, which then became smaller after reaching the half-flow point. The pressure load on impeller blade increased by staggered guide vane, and the pressure load at impeller blade outlet was always lower than that of original guide vane. The minimum pressure appeared near the inlet of impeller, while the maximum pressure appeared near the streamline with a chord length of 0.8. Furthermore, vortex rope in outlet of staggered guide vane and outlet pipe of volute is smaller than that of original guide vane scheme.

In this study, the core loading pattern of the initial core configuration of a typical Pressurized Water Reactor has been optimized through the Backtracking Search Algorithm (BSA). The multi-objective fitness function is based on a trade-off between minimization of the power peaking factor (ppf) and maximization of the cycle multiplication factor (k_eff) simultaneously. Neutronic computations are performed using the PSU-LEOPARD (Pennsylvania State University-Lifetime Evaluating Operations Pertinent to the Analysis of Reactor Design) and MCRAC (Multiple Cycle Reactor Analysis Code) codes. The PSU-LEOPARD generated assembly data have been fed to MCRAC and it calculates normalized power profiles for all fuel assemblies with a specific loading pattern. The BSA generates best loading patterns by optimizing the multi-objective function. The implementation of the BSA scheme resulted in slight enhancements in the first cycle length (∼10.1 %). The BSA demonstrates rapid convergence, high efficiency and robustness for the core loading pattern optimization problem.

The MYRRHA project aims to realise a sub-critical nuclear reactor coupled to a 600 MeV proton accelerator. Systems engineering was chosen as a paradigm to tame the multi-faceted complexity of such an endeavour across the various stages of its life-cycle. Tools, such as Polarion, are used to create and maintain a knowledge base of design assets as well as to facilitate the creation of up-to-date documents; still, discrepancies can exist between the knowledge base and the design models, only the latter of which timely capture the current envisaged design. This paper introduces the elements of an approach aiming to create a robust link between design models and technical documents. Such a robust link could be used to timely detect models conflicts; to lessen the document-centric management of life-cycle milestones; and to semi-automatically update the system knowledge base. As a proof of concepts, compliant tools and preliminary results are presented.

Numerous research endeavours are currently underway to advance supercritical water reactors (SCWR), acknowledged as a cornerstone among Generation IV nuclear reactor designs. Evaluation and managing the reactivity of the reactor is a vital issue in the reactor operation. This research seeks to identify efficacious burnable absorber (BA) materials and determine an optimal spatial distribution within the fuel assembly to regulate reactivity levels effectively. Gadolinium, erbium and Lutetium have been suggested as BA in the form of integral burnable absorber (IBA) rods ((UO₂ + Gd₂O₃), (UO₂ + Er₂O₃) and (UO₂ + Lu₂O₃)). Two SCWR assembly models, each featuring varied quantities and distributions of BA, have been examined. Various concentrations of the suggested BAs have been examined in the suggested models to verify the optimum cases. Burnup analyses have been conducted to evaluate the proposed cases. Different alloys of BAs including B₄C+Dy₂O₃ and B₄C+Sm₂O₃ have been investigated in the control rod and compared with the standard alloy B₄C.

Understanding the thermal–hydraulic safety and transient behavior of Gen-IV LBE-cooled fast reactors are crucial for advancing nuclear safety standards. The thermal–hydraulic performance of an LBE-cooled reactor, SPARK-NC, was analyzed using the subchannel analysis code LOONG-SACOS under steady-state natural circulation conditions, focusing on temperature distribution, velocity, and density in the hottest assembly. Results revealed a peak coolant velocity of 0.296 m/sec and a maximum coolant temperature of 471 °C, with the fuel centerline temperature remaining below 2000 °C safety threshold. This underscores the ability of SPARK-NC reactor design to maintain safe and efficient performance by regulating temperatures and flow rates within specified limits during steady-state natural circulation. In the subsequent phase, a transient analysis was conducted using LOONG-SARAX and DAISY-PK codes to evaluate the safety of SPARK-NC reactor under dynamic conditions, encompassing Unprotected Transient Overpower (UTOP), Unprotected Control Rod withdrawal (UCRW) and Scram-drop transient events across various core states. The study investigated UTOP transients by introducing positive external reactivity and evaluating the inherent reactor feedback behavior. The reactivity was increased incrementally to attain maximum reactivity while ensuring the integrity of both fuel and cladding. The results indicated that upon inserting external reactivity of 1.0$, there was an initial rapid power surge followed by stabilization, indicating that both fuel and cladding maintained integrity within the predefined failure thresholds. Additionally, analysis of UCRW transients enabled risk assessment during control rod maneuvers across various positions, wherein the withdrawal of control rod C₆ resulted in a total reactivity insertion of 0.94$, stabilizing at a normalized power level of 4.35. Finally, the scram-drop transient demonstrated the rapid shutdown capability of the reactor, promptly transitioning it to a secure state, ensuring effective post-insertion temperature control as feedback reactivity stabilizes at 0.24$, which highlights the robust inherent safety of the SPARK-NC reactor.

Granular bentonite is commonly regarded as a preferred anti-seepage and plugging material for the joints in high-level radioactive waste (HLW) repository. This study investigated the swelling, compression, rebound and permeability properties of MX80 granular bentonite with alkaline solution (0–1.0 mol/L). Results showed that compared with the alkaline concentration, the final swelling strain, compression and rebound indexes were negligibly affected by the particle size distribution. The preconsolidation pressure and hydraulic conductivity increased with increasing the concentration. When concentration c < 0.3 mol/L, the hydraulic conductivity was less affected by alkaline solution. When c > 0.3 mol/L, large aggregates were dissolved and resulted in a higher hydraulic conductivity. Results of scanning electron microscope (SEM) and processed images showed that the complexity, roughness and the size of the pores increased as the concentration increased. The dissolution of montmorillonite affected the pore size distribution and the hydraulic and mechanical properties of granular bentonite.

This work follows the AGR-like FHR ongoing research that proposes to adopt the British Advanced Gas cooled Reactor (AGR) geometry combined with the molten salt Fluoride-Cooled High Temperature Reactor (FHR) concept.

This work presents the new models and material properties implemented in the TRANSURANUS code for Cr-doped UO${}_2$ fuel, which is one of the advanced technology fuels considered for light water reactors. For this purpose, we update the mechanistic model for fission gas behaviour in the code by means of a dedicated fission gas diffusion coefficient recently proposed by Cooper et al. on the basis of atomistic scale simulations to take into consideration the impact of the dopant on the point defects that control the fission gas diffusivity in various temperature regions of interest. In a consistent manner, we propose also a modified creep correlation based on the mechanistic model for standard oxide fuels. Furthermore, we analyse the effect of cracking observed in doped fuels subjected to power ramps and take into consideration the limited densification of the high-density fuel reported in the open literature. A subsequent parametric study pointed out the main factors affecting the integral fission gas release rate, which was shown to be a limiting factor in the AGR-like FHR under consideration. Finally, the improved performance of the advanced technology fuel is shown by means of the reduced inner gas pressure at end of life, as well as the reduced pellet cladding mechanical interaction during the postulated operational transient from the open literature for AGRs. As a result, the doped fuel is shown to be able to sustain higher power levels in the AGR-like FHR.

Serpent 2 particle transport code is used to develop the pre-conceptual neutronic design of the Supercritical Water Cooled SMR. After initial criticality and burnup calculations, the starting core design of (Schulenberg and Otic, 2021) is improved using predetermined criteria, such as burnup cycle length and power distribution, while also considering operational safety. In order to achieve higher reserve reactivity, several modifications are considered, including the introduction of alternative structural materials and fuel assembly wall type, moderation improvement by adjustment of moderator temperature and fuel assembly gap width, and selection of a suitable enrichment map.

As a result of the introduced modifications, the burnup cycle length is increased to 26 months and an acceptable core power distribution is achieved. The improved core design can be used for further investigations, such as coupled calculations using neutronic and thermal–hydraulic codes and examinations targeting reactivity control during burnup.

Timely and accurate detection of defective fuel rods is critical as the release of radioactive fission products from defective fuels can lead to primary circuit contamination and radiation exposure. Due to the complexity of the physical phenomena, models for fault diagnosis can be difficult to construct and recently data driven surrogate models have being increasingly used to detect and characterize defective fuel rods: they make use of a computational database to learn from and make predictions about new unknown data. In this paper, we present a method for the elaboration of an anomaly detector based on neural networks, taking into account the fact that physical computation can be CPU intensive and thus overcome this issue. A physical model for fission products release and coolant activity calculation was built and used to generate a surrogate activity model that enables the generation of a bigger database in small amount of CPU times. Then using this bigger computational database, a recurrent autoencoder was trained for anomaly detection. The network classifies the defect status with 100% accuracy and a good time precision. A sensitivity analysis with lower activity increase at defect onset and addition of noise was conducted in order to better understand the limits of this method. Such methods can be useful for operators of the existing as well as future reactors to make timely predictions of defective fuel rods and avoid operational and economic setbacks for power plants. The work described in this paper was carried out within the R2CA (Reduction of Radiological Consequences of design basis and extension Accidents) project, funded in HORIZON 2020 and coordinated by IRSN (France).

In the present article, the applicability and accuracy of different boundary conditions for the simulation of turbulent, single-phase flows with heat transfer were assessed within the context of coarse-mesh CFD simulations for engineering applications. Standard wall functions for relevant turbulent quantities were extended to include geometry-dependent effects and implemented as boundary conditions for existing OpenFOAM solvers, along with a set of coarse-mesh wall models based on empirical correlations. The different models were tested in the simulation of numerical experiments where high-fidelity simulations can be provided and, in general, results show that the application of the new set of boundary conditions produces a satisfactory prediction of the streamwise velocity and temperature in the evaluated conditions, even when the first cell center is far from the wall. The analyzed extensions and corrections produce a better balance between accuracy and computational speed for coarse discretization, compared to traditional wall treatments.