Annals of Nuclear Energy最新文献

Due to advantage of high efficiency, low carbon, and flexible operation, the integrated supercritical CO₂ reactor plant has promising prospect in energy supply. This paper aims to clarify performance of the reactor plant and establish a reactor power control strategy through numerical modeling. Firstly, performance on separated components is evaluated to provide guidance for control strategy development. Secondly, influence of control methods, controller schemes, and saturation models on control strategy are discussed. Thirdly, dynamic response of reactor plant with proposed control strategy is confirmed via load change transients. Results show that passive control method is infeasible at EOL and load demand below 50 %FP, and active control method must be considered. The proposed control strategy adopts coupled temperature-channel and power-channel scheme and soft saturation model. It ensures safe operation of the plant over full load range in whole lifetime. These results can provide helpful guidance on part-load operation for this system.

In this study, we performed Monte Carlo coupled multi-physics simulations with spatially continuous material properties via Functional Expansion Tally combined with delta-tracking. The proposed multi-physics coupling framework significantly reduces the need for spatial discretization in the problem geometry, potentially decreasing simulation time as macroscopic cross section reconstructions are performed less frequently. The proposed method has been tested on pin-cell and assembly levels at hot full power to evaluate its applicability. At both pin and assembly levels, numerical experiments have demonstrated that the proposed framework can produce solutions that asymptotically approach those from conventional cell-based discretized simulations with infinitesimal cells. Furthermore, the proposed method accelerates the simulation time by over four times compared to the cell-based discretized approach with very small cells for cases without neutron absorbers. Therefore, the proposed method has the potential to be integrated into future Monte Carlo production codes, to meet the demanding requirements for improved reactor safety.

The present study focuses on the influence of neutron scattering cross section and angular distribution when employing the Thermal Scattering Law (TSL) on criticality calculation benchmarks, in comparison with the Free Gas Model (FGM). The benchmarks sensitive to beryllium oxide (HMT-027) and graphite (HMT-026) as well as some simplified Pressurized Water Reactor (PWR) benchmarks are interpreted using the OpenMC code to assess the influence of TSL. The results indicate that the influence of TSL mainly attributes to both the total scattering cross section and the secondary angular distribution characterized by the average cosine of the scattering angle. The HMT-027 and HMT-026 series benchmarks respectively show the more significant influence of one of these two effects. For the simplified PWR benchmark, employing the TSL of H in H₂O weakens the neutron moderation effect, which results in a reduction of around 100 pcm in the calculated $k_{\text{eff}}$, while the combined influence of cross section and secondary angular distribution is negligible for the TSL of UO₂ fuel.

The rod drop experiment was performed at the VR-1 reactor to test the detection system and collect data that describe the development of neutron flux during rod drop. This experiment was calculated in Serpent code by its dynamic part. During this experiment, an additional control rod for drop was placed in the reactor with a channel directly attached to the rod guide tube containing the detection system. The detection system developed for this experiment is based on diamond detectors and allows measurement of neutron flux development during the rod drop in almost point vise location. The results of both experiment and calculation showed a delay in the count rate drop influenced by the position of the rod during the drop. The obtained experience is also used in the design process of a new experimental device dedicated to rod drop measurements that will allow the direct rod position measurement during the drop.

This study provides a thorough simulation and analysis of cross-sections for the production of ⁵¹Cr by various nuclear reactions. Our objective is to validate the production process of ⁵¹Cr, an essential radionuclide used in nuclear medicine for studying blood cells’ physiological and pathological characteristics. In order to do this, we used the nuclear level density, optical model potential, and preequilibrium model integrated into the TALYS 1.96 code for theoretical modeling. The obtained results have been compared with experimental data taken from the EXFOR database. We have also taken TENDL library data and TALYS as a whole code to enhance our evaluation. The study assesses multiple nuclear reactions: ⁵¹V(p,n)⁵¹Cr, ⁵¹V(d,2n)⁵¹Cr, ⁴⁸Ti(a,n)⁵¹Cr, ⁵²Cr(n,2n)⁵¹Cr, ⁵⁴Fe(n,a)⁵¹Cr, and ⁵⁵Mn(p,x)⁵¹Cr in order to identify the most effective routes in terms of production, relative variance analysis, presence of nuclidic impurities, and the optimum energy range. The cross-section, theoretical yield, target thickness, and activity have been calculated to optimize and help in finding the best reaction conditions, which improve the production of ⁵¹Cr inside a cyclotron for medical uses.

Continuous degradation and shocks are two vital damage mechanisms for tubes under dynamic environments in nuclear power plants (NPPs). In this work, we presented a cumulative damage model to evaluate the reliability of tubes in NPPs, which accounts for continuous degradation and random shocks. The dynamic environments were assumed to obey a homogenous Markov process. Continuous degradation was described as a general stochastic process with independent increments. A multinomial distribution was used to depict the multiple effects of Poisson-distributed shocks on tubes. The influence of dynamic environments on the cumulative damage process of tubes was also explicitly modeled. Closed-form reliability measures were derived, such as reliability functions and mean time to failure (MTTF) of tubes. Drawing upon a case study from the existing literature, the analytical solution was validated through a comparative analysis with the outcomes of Monte Carlo simulation (MCS) and other methodologies referenced in the literature. As a practical engineering application, the reliability of steam generator tubes made of Alloy 690 within pressurized water reactors (PWRs), which are susceptible to primary side stress corrosion cracking (PWSCC) and the effects of water hammer, was calculated. The findings indicate that the probabilities of steam generator tube rupture (SGTR) as estimated by the model and those inferred from operational experience have the same order of magnitude.

The NACIE-UP (NAtural Circulation Experiment-Upgrade) facility is located at the ENEA Brasimone Research Centre (Italy) and is designed to investigate the thermal–hydraulic properties of Lead-Bismuth Eutectic (LBE) within a wire-spaced assembly under uniform and non-uniform power distribution conditions. This benchmark exercise proposed by ENEA is based on previous experiment results, intending for development and validation of CFD/TH system code, stand-alone CFD simulation, etc. In this study, the thermal–hydraulic characteristics of Fuel Pin Simulator (FPS) are analyzed, which contains inlet bend and outlet section. ANSYS Fluent solver is applied with the COUPLE algorithm of RANS $\text{SST}k-\omega$ model. Cheng&Tak correlation and computational conditions are implanted into Fluent in the form of UDFs. Moreover, calculations were performed for ADP10, ADP06 and ADP07. The results show that the RMSEs of ADP10SS1 and ADP10SS2 are 2.52 K and 3.69 K, 7.05 K and 8.48 K for ADP06SS1 and ADP06SS2, respectively, which indicate that the adopted models have good prediction ability for the temperature of the experimental conditions. Meanwhile, the heat transfer characteristics of different sub-channels and sections are investigated and compared with the correlations of Ushakov, Mikityuk, Kazimi and Carelli. The experimental and simulation values of the uniformly heated ADP10 are in good agreement with the correlations, while there is a large deviation for the non-uniformly heated ADP06. Therefore, it is necessary to develop new correlations for the heat transfer characteristics of LBE in wire-wrapped bundle. Above all, this study provides quantitative and qualitative analysis for verification and validation of CFD/TH system code and stand-alone CFD simulation.

The critical systems in industries such as nuclear power plants rely on various preventive methods to minimize failures or risks through efficient strategies and equipment. However, in many businesses, maintenance tasks are carried out infrequently, improperly, and without consideration for the overall state of the plant or its equipment. To choose the appropriate risk preventive approach, a thorough examination of each component’s risks in a sequential manner becomes imperative. This paper introduces the Fuzzy Analytic Hierarchy Process (AHP)-VIKOR technique, a Multicriteria Decision Making Approach employed to rank the various risks prevalent in the nuclear power industry. By identifying the proper sequence of risks, this approach aims to reduce the occurrence of unfortunate mishaps, along with minimizing recovery time and costs. Five experienced researchers and experts assessed the impact of risk based on three risk criteria: Severity, Occurrence, and Detection. Utilizing the opinions and judgments of these experts, the Fuzzy AHP-VIKOR approach was employed to calculate the weight of each performance criterion and the ranking of each risk or hazard. The suggested method is designed to assist supervisors in resolving discrete problems characterized by incommensurable and conflicting criteria. This study contributes to the industry by providing a higher risk analysis incorporating various performance parameters. The paper concludes by presenting a result with the priorities of all risks in the industry using the fuzzy AHP-VIKOR method.

This study introduces an explainable artificial intelligence (XAI) approach designed to estimate background spectra in unknown spectral measurements. The approach combines kernel-modeled Gaussian processes (GP) for naturally occurring radioactive material (NORM) estimation with fuzzy logic inference for isotopic photopeak identification. Recognizing the diverse interpretations of background radiation, the paper’s objective is to propose a multi-mode driven approach, with each mode implementing a distinct set of fuzzy rules, thus modeling different backgrounds. Importantly, each mode includes rules associated with nuclides expected to be present in specific locations, such as medical isotopes in a hospital setting. A key innovation of the method is the additional step of providing explanations for the estimated contributions that accompany the estimated background spectrum. Results obtained from a range of gamma-ray spectra representing different locations demonstrate the framework’s potential in estimating background radiation and aiding decisions in the nuclear security domain, particularly for identifying potential nuclear threats in unknown measurements.

The desire to improve the economic competitiveness and deployment pace of nuclear energy through modularization, manufacturing, and series production had led to the development of smaller size reactors. As the standard 17x17 fuel technology is mainly maintained in the pressurized water reactors (PWRs) category, this translates into a lower number of fuel assemblies in the core and sometimes a reduced fuel height. To assess the impact of such scale change in core design on fuel cycle cost and spent fuel volume, a scoping analysis tool is developed based on infinite lattice calculations, leakage, fuel management reduced models, and levelized unit cost of electricity (LCOE) estimate. As such, cost dynamics driven by fuel specific power, burnup, core leakage, feed, cycle length, fuel assembly height as well as uranium market data are captured with consistent set of assumptions and analysis methods. A selection of 5 reactor designs representative of leading PWR developers is assessed and compared. Pursuing higher specific powers and optimal burnups are highlighted as the main fuel cost reduction drivers, nevertheless, practical limitations and opportunities must be evaluated to establish the feasibility of such enhanced fuel operation. In consequence, a detailed core design is performed using SIMULATE3 code for 5 PWR variations including natural and forced coolant circulation modes, two reactor scales, power densities of 73, 112, and 123 kW/l and higher discharge burnups. Design and optimization are performed at the lattice level, for the reflector, and at the core loading level. Satisfactory steady-state operation including power distribution, coolant operating limits, and reactivity requirements are analyzed and reported in this paper. The fuel economics of the detailed core designs confirm the scoping analysis findings. Despite the unlocked power uprates in small PWRs, the achievable burnup for a given fuel specific power requires more enrichment and shorter fuel height results in higher fabrication costs per mass of fuel, which makes scaling down core size a more expensive endeavor on the fuel cycle front. Spent fuel volumes are reported for the PWRs designed in this paper. These volumes are driven by the core average discharge burnup regardless of the scale in consideration. Additional cost and core performance aspects related to heavy reflector gains, fuel-reflector substitution, and disposal cost policy in the U.S. are examined.