Progress in Nuclear Energy最新文献

Release rate estimation is crucial for the consequence assessment and emergency decision-making in nuclear accidents. However, inevitable model biases can lead to significant deviations. This study proposes an Adaptive Center Constraint for joint release rate estimation and model correction (ACC joint) for improved robustness and accuracy. It uses a tailored cost function to determine the optimal center constraint, which can automatically adapt to different cases. It was validated against four wind tunnel experiments, which simulated complex dispersion scenarios with densely built-up and highly heterogeneous terrains. The ACC joint method was compared with the Tikhonov and joint correction. The results indicate that the proposed method significantly improves the accuracy of release rate estimation. Compared to the joint correction method, the mean relative error is reduced by 38.6% and 31.4% in the all-measurement and independent validation, respectively. Furthermore, sensitivity analysis reveals that the ACC joint method provides lower mean relative error with different numbers of measurements and shows ultimate stability in all scenarios. It also suggests that measurement sites should be positioned in downwind high-concentration areas and the foot of mountainous areas for reliable estimation. The results from different cost functions verify the scalability of the proposed method, providing potential applications to other complex scenarios.

Because of their superior radiation damage tolerance, Ferritic/Martensitic (F/M) steels are candidate materials for fuel cladding tubes in nuclear reactors. This study applies extreme gradient boosting (Xgboost) and artificial neural network (ANN) algorithms to predict the outer-diametral strains of several types of F/M steel cladding tubes when subjected to the synergistic effects of neutron irradiation and creep. Key input variables were identified using the SHapley Additive exPlanations (SHAP) algorithm and a Genetic Algorithm-based feature selection method. After deploying these machine learning (ML) algorithms, they were trained and tested across their respective hyperparameter ranges. Thereafter, the ML models’ performance was assessed on an unseen/validation dataset. The outcomes revealed that the fine-tuned Xgboost model exhibited superior predictive capabilities for the validation dataset compared to the ANN model. Moreover, the trained Xgboost model surpassed the predictions made by the empirical model that couples irradiation creep with void swelling. Additionally, a more in-depth analysis of the Xgboost model was conducted to determine its ability to capture intricate relationships between input variables like irradiation dose and hoop stress and the output variable—namely, outer-diametral strain. Through synthetic experiments, it was revealed that the Xgboost model could indeed comprehend these complex interconnections effectively. Despite the challenges, such as smaller and sparser datasets with uncertainties, ML models successfully overcame the complexities and effectively predicted the outer-diametral strains in F/M steel tubes during neutron-irradiation creep.

Small Modular Reactors (SMRs) have gained significant attention as the next generation of nuclear reactors due to their unique features, such as improved economic viability and safety. This study focuses on the SMART reactor, the first certified design of an SMR, and investigates various Thermo-Neutronics aspects using an integrated coupling scheme of nuclear codes. The Neutronics cell and core calculations are performed using the DRAGON/DONJON codes, while the COBRA code is employed for thermal-hydraulic calculations. The analysis considers parameters including axial and radial power peaking factors (PPFs), critical heat flux (CHF), minimum departure from nucleate boiling ratio (MDNBR), axial average coolant temperature in hot channel and core, and maximum fuel temperature. These parameters are evaluated using the developed coupling system. Code-to-code verification is conducted to ensure the accuracy of the results, demonstrating good agreement. The coupling approach reveals significant effects compared to stand-alone code modeling, emphasizing the importance of considering the coupling between Neutronics and thermal-hydraulics, especially during the cycle length. This study contributes to the understanding of SMRs, particularly the SMART reactor, by providing insights into key parameters and their interdependencies. The integrated coupling scheme enhances the accuracy of the analysis and highlights the significance of considering coupling effects in the modeling process during the steady state and the cycle length.

To enhance the transport efficiency of spent nuclear fuel, a new free-standing storage basket has been designed, which can store spent fuel assemblies and transport them simultaneously. The seismic performance of these baskets is a critical component of nuclear safety engineering, related to the containment of radioactive products. The basket is tall and thin in comparison to traditional racks, making it easier to rock or overturn during an earthquake. In a previous study, we carried out a series of shaking table experiments on a single basket, and we found that the single basket would not overturn in seismic conditions. However, the interaction between multiple baskets and fluid may also affect their seismic response, which requires further investigation. To obtain the dynamic response of the multiple baskets during earthquakes, a series of 1/2 scale-down shaking table tests are conducted. Beat tests are employed to study the impact of fluids on spent fuel storage baskets. Seismic tests are carried out in both velocity similarity laws and acceleration similarity laws to obtain the response of baskets. The results reveal that the baskets remained stable during the earthquake and did not overturn or collide with the baskets and wall. Furthermore, the combination of multiple baskets reduces the sliding and rocking amplitudes.

The paper considers the mathematical formulation of the problem of transverse oscillations of a vertical rod under gravity, friction and external pulse effect, leading to thermal expansion of the rod. The dynamics of the system under consideration corresponds to the behavior of a fuel element (FE) in a pulsed reactor and is related to the dynamic stability of the processes occurring in it.

The FE dynamics is described by inhomogeneous linear differential equation of the fourth order with non-constant coefficients. Initial boundary conditions are not smooth since they correspond to the instant heating of the part of the FE surface exposed to neutron pulse radiation. The general solution of the homogeneous linear equation can be found using the concept of generalized functions expressed as a series expansion in terms of coordinate eigenfunctions dependent on $p$ parameter. The $p$ parameter is related with the FE mass and at some certain values results in bifurcation points of the boundary value problem for eigenfunctions. The partial solution can be found in several ways: using the Fourier transform, the method of Green's function, and in terms of series expansion by eigenfunctions.

Eigenfunction expansion coefficients in an explicit form have been obtained and the numerical solution accuracy has been estimated for some important particular cases. The results of the obtained exact solutions appear to be very close to the results of the ANSYS numerical estimations. In the future the obtained exact solutions will be used as input data to simulate self-consistent system dynamics of more than 400 FE. Therefore, the advantage of the exact solution in practical use is that its calculation is much faster as compared with the finite element method done with ANSYS.

The supercritical water cooled reactor (SCWR) is stands out among the Generation IV reactors with high safety and economy. Understanding the fluid flow and heat transfer characteristics within rod bundles under near-critical conditions is crucial for ensuring the safety of the SCWR core. In this study, computational models utilizing the SST k-ω turbulence model and boundary-resolved grids were validated against heat transfer data for near-critical water within a rod bundle. Subsequently, the validated model was utilized to examine the impacts of system pressure, fluid temperature at inlet, mass flux, and heat flux on the heat transfer characteristics of rod bundle with spacer grids under subcritical and supercritical conditions. The study revealed that the heat-transfer coefficient under supercritical conditions is significantly larger than that under subcritical conditions. Additionally, the heat transfer coefficient increases with mass flux, while independent from pressure and heat flux.

In small break LOCA sequences with failure of the high-pressure safety injection system, the reactor coolant system pressure can stagnate at a high value making the medium and low-pressure safety injection systems unable to inject water into the core before its peak cladding temperature exceeds the safety limit. In this work, a review and comparison of different strategies presented in the Emergency Operating Procedures for managing these sequences in VVER-1000/V320 and Westinghouse PWR has been carried out. For this purpose, the Integrated Safety Assessment methodology, developed by the Spanish Nuclear Safety Council has been applied. The results show that the strategy related to the controlled SGs depressurization at a primary side cooling rate of 60 K/h in VVER-1000/V320 reactors and 55 K/h in Westinghouse PWR provides a wide safety margin. In cases where the Inadequate Core Cooling temperature is reached, the fast SGs depressurization strategy is also effective to avoid the core damage.

One-phonon correction method has been implemented into the thermal scattering law (TSL) calculation module sab_calc in nuclear data processing code NECP-Atlas, to enhance the accuracy of TSL data for inelastic scattering. This method relaxes the incoherent approximation used in the conventional TSL calculation code. The updated sab_calc module is used to calculate the TSL data of graphite, α-SiO₂, and α-Al₂O₃. The microscopic cross sections of these materials are calculated with the obtained TSL data, and compared with the available experimental and theoretical data. The results show that the sab_calc module are capable of providing accurate TSL data for inelastic scattering.

This study explores the effectiveness of the mode-K+ strategy in managing a Very-Low Soluble Boron (VLSB) APR1400 core, loaded with LEU + fuel, and utilizing the Centrally-Shielded Burnable Absorber (CSBA) and erbia for excess reactivity control. Initially, a 24-month two-batch fuel management scheme was developed, targeting an equilibrium cycle Burnup (BU) of 26 GWD/MTU. The scheme required a maximum of ∼550 ppm Critical Boron Concentration (CBC) in the early cycle BU. Analysis of radial and axial power profiles revealed that the power peaking factors remained within acceptable limits, with values not exceeding 1.4 and 1.6 throughout the cycle, respectively. Subsequently, various Daily Load-Follow Operation (DLFO) scenarios were examined at different BU levels. These investigations demonstrated a consistent alignment between the demanded power and the reactor's power output. Furthermore, critical safety parameters, including the Axial Shape Index (ASI) and inlet coolant temperature, consistently remained within prescribed design specifications. The study's findings underscore the feasibility and reliability of implementing a VLSB APR1400 core with LEU + fuel, using mode-K+ and advanced fuel management strategies for effective power generation and safety compliance. To perform the analysis, an in-house diffusion code KANT was utilized, KANT is based on NEM-CMFD accelerated simulation tool, while the cross-section was generated via Serpent 2.2.0 assessed with ENDF VII.B data library.

The analytical approach that was proposed in our recent paper has been applied to simulate effects of pH and redox potential (E) on the sorption of uranium onto potentially redox active bioorganic model particles in saline or other aquatic environments. Specifically herein, it is applied to the mono- and poly-hydroxy-aromatic (polyphenolic) sites which account for approximately 30% of bioorganic site capacity. The derived expression is aimed to avoid use of the classical approach of sorption, which requires experimental data and empirical models. The expression provides a distribution coefficient (K_d e.g. mL g⁻¹) as function of pH, E and soluble ligand concentration by considering a surface complexation model on mono- or multi-dentate complexation surface sites > Su(OH)_c. The application of the model uses correlations between the surface complexation constants and hydrolysis constants, for all potential species and all form of sorption sites. The model was used to quantify the uranium sorption onto hydroxy-benzene, dihydroxy-benzene, and dihydroxy-naphthalene sites with or without carbonates in solution. The latter is the primary interfering reagent in waters that decreases Log K_d. The calculated distribution coefficients were found sensitive to both pH and E and very sensitive to the presence of carbonates. The reduction of uranium U(VI), and its carbonate complexes, to U(IV) during sorption was simulated by decreasing the redox potential. It was found that the transition phase between U(VI) and U(IV) was generally below the redox stability limits of water. However, the reduction of U(VI) to U(IV) was found to be potentially associated with their reaction with the polyphenols, decreasing the redox potential subsequently. The calculated sorption coefficient values were validated using the values reported in literature for the sorption of uranium onto specific adsorbents. The methodology of the simulation is also applicable to the sorption of other redox sensitive elements, and with the addition of a scaling factor, it would allow the predictions of co-complexation phenomena by employing relevant site formulations. The oxidation of mono-hydroxy- benzene in di-hydroxy-benzene enhances the sorption of uranium by a factor 10⁶ which may be applied to its extraction from seawater. XX.