Progress in Nuclear Energy最新文献

Natural circulation is the important way for taking away decay heat in pool-type sodium-cooled fast reactor (SFR) when all primary pumps trip. Three-dimensional (3D) unrestricted flow and thermal stratification phenomena exist in sodium pool during natural circulation. 3D modeling of whole reactor system with complex secondary or third circuit may generate huge numbers of grid, it may cause the difficulty of performing simulation. 3D computational fluid dynamics (CFD) method is mainly used for the thermal hydraulic analysis in primary system. The power variation of heat exchanger is unclear under natural circulation condition, which is key boundary condition in 3D CFD simulation. It is necessary to develop the numerical method for realizing the boundary more closed to the actual condition. In this paper, a “1D system code + 3D CFD” simulation method combining the simulation of 3D CFD and system code SAC-IRACS was proposed. The steady and transient simulations with 24000 s of PHENIX natural circulation test were performed. Flow paths in PHENIX primary system during the test were identified and mass flow rate were obtained. 3D thermal stratification phenomena in sodium pool during natural circulation test were captured. Variation trends of key thermal hydraulic parameters are basically consistent with the test. There is the reverse flow in RVCS under natural circulation conditions. At 24000 s, mass flow rate of reactor core, primary pumps and RVCS are 63.13 kg/s, 58.04 kg/s, −5.11 kg/s respectively. The decay heat in primary system can be taken away by intermediate reactor auxiliary cooling system (IRACS). Obvious thermal stratification phenomena appear again in sodium pool with the efficient cooling. This “1D system code + 3D CFD” simulation can provide reasonable results, and can be applied to natural circulation analysis in other pool-type SFRs.

This study comprehensively evaluates thorium-based fuel utilisation in the Molten Salt Reactor (MSR) with a once-through fuel cycle. It examines the whether the use of thorium is beneficial in extending the fuel cycle length of an MSR, fuel utilisation efficiency, reactor safety, and reduction of long-lived radioactive waste. To observe those parameters, this research compares the use of thorium and uranium as fertile fuel in a single-fluid, once through MSR using Uranium-233 (U-233), Uranium-235 (U-235), and Reactor Grade Plutonium (RGPu) as well as the fissile driver. MCNP radiation transport code with ENDF/B-VII.0 continuous neutron library was used to analyse the neutronic parameters and fuel burnup in a multi-refuelling scheme for a total burnup time of 8 years. The findings indicate that the utilisation of thorium in conjunction with U-233 and U-235 significantly enhances fuel consumption and improve reactor safety. Such benefits, however, did not appear in RGPu. Moreover, thorium effectively reduces the production of long-lived radioactive waste, a crucial issue in nuclear waste management. Considering technical and environmental aspects, this study offers novel insights into using thorium as a more sustainable nuclear alternative under a once through fuel cycle, within a condition of multi-refuelling cycle which is only possible in an MSR.

The Main Steam Line Break Accident (MSLB) threatens the safe operation of nuclear power plants. The transient safety parameters of the Passive Containment Cooling System (PCCS) in the MSLB accident are predicted by the time series deep learning model based on LSTM or RNN. In the dataset preprocessing, the linear normalization and two of feature label segmentation methods are used. The neural network models and the other models, namely the GBRT, RF and SVR, are used to construct the multi-parameter time series deep learning models to predict the transient safety parameters. The performance of the different models and the impact of two segmentation methods on models are compared, and it is obtained that the prediction accuracy of LSTM is higher than that of other models.

During the conceptual design of the Mars surface nuclear reactor coupled with supercritical carbon dioxide Brayton cycle, the heat rejection system occupies most of the weight, and reducing the weight of heat rejection system can effectively cut down the transportation cost from Earth to Mars. This paper explored the feasibility of adopting a combination of heat pipe cooling device and convective heat exchanger to design the heat rejection system for the Mars surface nuclear power station, established flow and heat transfer model and weight model, developed a thermal-hydraulic design program for the heat rejection system, and verified its accuracy; then, the NSGA-II multi-objective optimization method is used to optimize the weight of heat rejection system. Pareto front shows that with higher heat rejection proportion of convective heat exchanger, corresponding inlet velocity of Mars' atmosphere should also be increased. As the optimization result of heat rejection system, the ratio of weight to heat rejection is between 0.94 kg/kW and 2.92 kg/kW, and the maximum power consumption of convective heat exchanger accounts for 3.29% of the total power generation.

Addition of Zn²⁺ and Mg²⁺ ions to the coolant water is followed in boiling and pressurized water reactors to control the radiation field. These cation additions modify the composition of double-layered spinel nickel chromium ferrite oxide films formed on stainless steel. The primary objective of this paper is to offer a systematic perspective for comprehending the dissolution behaviour of non-stoichiometric oxides upon Zn²⁺ and Mg²⁺ addition. In this regard single-phased nickel chromium ferrites were synthesized and characterized by various techniques. The crystal morphology and band gap changed with an increase in lattice dimensions upon incremental substitution of Ni. The oxidative dissolution rates increased with increasing Mg²⁺ or Zn²⁺ in the Ni–Cr–Fe–O lattice with a maximum at x = 0.6. Magnesium-substituted nickel ferrites showed a higher dissolution kinetics compared to their zinc-substituted counterparts. The dissolution behavior and kinetics are explained on the basis of cation redistribution in the octahedral or tetrahedral sites of the spinel lattice. The Laser Raman Spectroscopic (LRS) analysis corroborated the X-ray diffraction (XRD) and X-ray photoelectron spectroscopy (XPS) observations. These results will provide an insight into the mechanism of radioactivity removal kinetics for developing effective dose rate reduction practices for nuclear plants.

Convergence study of two-dimensional discrete ordinates (SN) method has been carried out by Fourier analysis for single-group neutron transport k-eigenvalue problems. However, conclusions and improved coarse mesh finite difference (CMFD) acceleration method derived from single-group Fourier analysis are not fully applicable to the realistic multigroup problems. The convergence characteristic of two-dimensional SN for multigroup problems has not been systematically investigated, which is an important work that complements the existing studies. In this study, a Fourier analysis for solving multigroup neutron transport k-eigenvalue problems is performed. Firstly, the influence of multigroup structure is analyzed and results show that when neglecting the intergroup scattering, the spectral radius of multigroup case is the maximum value of all the single-group cases. Then, the effects of scattering ratio on the convergence behavior are presented. Fourier analysis results show that when increasing the within-group scattering ratio, the spectral radius of the whole iteration decreases. While for the intergroup scattering ratio, the phenomenon is totally opposite. When increasing the intergroup scattering ratio, the spectral radius increases. Lastly, the diffusive coefficient of CMFD is guided based on the Fourier analysis results, which considering the influences of the intergroup scattering. Numerical results show that improved CMFD achieves better convergence performance for 2D C5G7 benchmark and especially for high intergroup scattering problems.

The most important characteristic of main Steam Line Break (SLB) accident is the asymmetry cooldown between loops due to secondary break spurting, which results non-uniform reactivity feedback in different reactor core sections. The fuel rod DNBR limit would be more challenged in the core section with peak power. The SLB safety analysis methodologies in Administration License Application for typical PWR in China are investigated in this paper. Based on the conservative universal methodology, the best-estimated code RELAP5/MOD3 is used to perform the SLB thermal-hydraulic calculation for a three-loop PWR. In this analysis, the critical conservative models submitted for Licensing process are implemented in RELAP5/MOD3 specified modeling, which includes the Three-Channel Core Model, core inlet mixture array, core outlet mixture, vessel upper head flow and the more realistic SG model. The “second pressurizer” phenomenon in vessel upper head flashing is studied, which affects the system pressure response and mitigation. The effect of SG spouting during the transient is also studied in this paper, which is different with lumped parameter SG model. The combined approach for Deterministic safety analysis is validated in this paper, which is composed with best-estimated code, Conservative assumptions and Conservative initial & boundary conditions. Because of the flexibility and adequate conservatism, it can be spread for design optimization and independent safety assessment from the third party.

Monitoring radioactive releases during the normal operation of power plants is crucial to ensure compliance with safety limits set by regulatory bodies for environmental safety. In the case of the Barakah nuclear power plant in the UAE, comprehensive multiunit dispersion modelling and radiological safety analysis have been conducted. Utilizing the HotSpot Health Physics Code and GENII (Second-generation environmental dosimetry), assessments were made for both 37 gaseous and 51 liquid source terms generated by GALE. Release rates for source terms were determined using GALE based on APR 1400 specifications. HotSpot employed the Gaussian Plume model to simulate the dispersion of gaseous source terms up to 80 km surrounding the plant, calculating Total Effective Dose Equivalent (TEDE) and Committed Effective Dose Equivalent (CEDE) in rural and urban areas. Findings indicated that neither scenario exceeded the 1 mSv threshold for the general public nor the 20 mSv limit for operational workers. Notably, skin, thyroid, and surface bones exhibited the highest CEDE, primarily influenced by iodide radionuclides. GENII's Surface Water module modelled the effects of liquid source terms, accounting for various contamination and exposure pathways such as external exposure and ingestion across different age groups. The calculated doses remained well below FANR's annual limits, with negligible cancer incidences and fatalities predicted for one year of exposure.

A refined coating degradation model that comprehensively considers the phenomena of coating oxidation, diffusion, reduction, and volatilization is used to evaluate the performance of accident tolerant Cr-coated zirconium alloy cladding under accident conditions. The typical accident types of Station Blackout (SBO) and Large Break Loss of Coolant Accident (LBLOCA) are analyzed, with the consideration of three modes of coating failure: Failure in coating oxidation stage, Failure in oxide layer reduction stage, Failure caused by eutectic reaction. For the SBO accidents with relatively long time, the initial coating thickness of 9∼10 μm is recommended, which could provide good protection for zirconium alloy throughout the entire process and trigger the reduction of Cr₂O₃ before reaching Cr–Zr eutectic temperature to inhibit the occurrence of eutectic reaction. For the LBLOCA accidents where the cladding is rapidly heated, eutectic reaction is the dominant factor leading to coating failure, so how to avoid eutectic reaction failure remains to be further studied.