Small modular lead-based reactors with features of compact design and environments of high neutron fluence and high temperature urgently require a structural-functional integrated neutron shielding material that combines good high-temperature mechanical properties and neutron shielding performance. As boron- and gadolinium-containing steels exhibit insufficiency in mechanical properties at high temperature and process-ability, this study selected GdAl3C3 cermet as the reinforcing phase and neutron absorber and 316L stainless steel as the matrix in order to develop a GdAl3C3-316L shielding material. The GdAl3C3-316L shielding material samples with GdAl3C3 additions of 0.5%, 1.5%, 3.0% and 5.0% (in weight percentage) were prepared by powder metallurgy. Their micro-structure, tensile properties at room- and high-temperature and neutron shielding performance were systematically analyzed. The results show that chromium-rich phases agglomerate in the material as the GdAl3C3 content increases. The formation of a dual-phase structure of ferrite and austenite is promoted in 316L matrix with the solid solution of Al. Furthermore, the GdAl3C3-316L samples exhibit a bimodal grain size distribution and the addition of GdAl3C3 refined the matrix grains. With GdAl3C3 addition among 1.5% to 3.0%, the material simultaneously possesses a high proportion of fine-grained regions (＞82%), a relatively high proportion of high-angle grain boundaries (＞92%), and an appropriate proportion of twin boundaries, achieving a better balance among fine-grain strengthening, ductility coordination, and crack deflection capability. Various in situ formed Gd-Al-O ternary oxides are observed in 1.5%GdAl3C3-316L sample, including nanoscale monoclinic Gd4Al2O9 precipitates, submicron simple orthorhombic GdAlO3 precipitates, and body-centered cubic Gd3Al5O12 precipitates, which are dispersed in 316L matrix. The tensile properties at room- and high-temperatures of the samples initially increase and then decrease with the increasing of GdAl3C3 content. The 1.5%GdAl3C3-316L sample demonstrates good mechanical properties with a yield strength of 696.2 MPa and 613.2 MPa, and an elongation of 22.4% and 14.9% at room temperature and 550 ℃, respectively. The SEM image of the tensile fracture surface of the sample indicates that the material as a whole exhibits ductile fracture characteristic. The neutron shielding experiments show that 3.0%GdAl3C3-316L and 5.0%GdAl3C3-316L samples achieve a thermal neutron shielding rate of 81.25% and 95.84% at a thickness of 0.9 mm, respectively. The neutron shielding performance of the material is improved with the increase of GdAl3C3 content. Compared to Gd-316L alloys with similar gadolinium content, the 1.5%GdAl3C3-316L alloy in this study exhibits a better synergy of tensile strength and plastic deformation capability, which indicates its potential for use as a structural-functional integrated neutron shielding material in lead-based reactor.
During the in-reactor operation of pressurized water reactor (PWR) fuel rods, hydrogen from the coolant environment diffuses into the zirconium alloy cladding, leading to the precipitation of hydrides within the zirconium matrix. This phenomenon represents a critical issue in nuclear materials because hydride formation embrittles the cladding material, which can initiate crack propagation and ultimately result in material failure. Conventional methods for quantifying the hydride content in damaged fuel rods rely mainly on destructive analysis techniques, such as post-sectioning metallographic examination or high temperature hydrogen extraction. However, these approaches suffer from significant limitations, including the complexity of preparing irradiated samples and the inability to examine multiple locations on a single specimen. Such constraints hinder a comprehensive assessment of cladding integrity under operating conditions. To address the technical challenges in hydrogen analysis, this study employed neutron imaging as an advanced non-destructive testing method. The fundamental principle of neutron imaging is based on the difference in neutron attenuation between hydrogen and zirconium atoms. Through dedicated data-processing algorithms, the neutron attenuation data were converted into hydrogen distribution maps, with corrections applied for factors such as material thickness and the neutron energy spectrum. The results clearly demonstrate that neutron imaging can effectively visualize the distribution patterns of hydrides within the cladding, revealing pronounced hydride enrichment adjacent to cracks and other structural defects. Semi-quantitative analysis indicates that the average hydrogen concentration on the surface of the irradiated fuel rod is about 100 ppm, while localized hydrogen concentrations in hydride enriched regions near cracks reach approximately 5 000-10 000 ppm. This work validates the feasibility of neutron imaging to characterize the hydrogen content in damaged fuel rods. The findings contribute to optimizing the in-service performance of zirconium alloy cladding materials and provide a basis for future design improvements aimed at enhancing reactor safety and longevity.
The rapid advancement of micro-electro-mechanical system (MEMS) has imposed increasingly stringent demands on integrated energy devices, particularly in terms of miniaturization, long operational lifetime, and high structural and operational stability. In this context, radioisotope photovoltaic batteries have attracted sustained interest as long-lived power sources capable of operating in environments where conventional energy harvesting technologies are ineffective. However, despite their high radiation tolerance and technological maturity, traditional inorganic radiophotovoltaic isotope batteries remain fundamentally limited by their rigid device architectures, complex and energy-intensive fabrication processes, and insufficient spectral compatibility with the narrowband and extremely low-intensity radioluminescence emitted by tritium-based light sources. These intrinsic constraints significantly hinder their application in flexible, lightweight, and weak-light energy harvesting systems required for next-generation MEMS and self-powered microsystems. To address these challenges, this study proposed an alternative material and device strategy for radiophotovoltaic isotope batteries based on wide-bandgap organic photovoltaic (OPV) materials. Owing to their tunable absorption spectra, solution processability, low cost and compatibility with flexible and miniaturized device architectures, wide-bandgap organic semiconductors provide a promising pathway for enhancing spectral matching with radioluminescent sources while maintaining structural adaptability. In this work, particular emphasis was placed on the narrowband and ultra-weak emission characteristics of a 3H/ZnS∶Cu tritium radioluminescence source. Two representative wide-bandgap non-fullerene OPV systems, namely PM6∶Y6 and D18∶L8-BO, were fabricated and systematically investigated as active layers for radiophotovoltaic energy conversion. Device performances were evaluated under both standard AM1.5G solar illumination and radioluminescent excitation from the 3H/ZnS∶Cu source, enabling a direct comparison of their photovoltaic behaviors across vastly different photon flux regimes. The experimental results demonstrate that wide-bandgap organic systems exhibit pronounced advantages in radioluminescence-driven energy harvesting. Under an ultra-low irradiance of 8.00×10−4 mW/cm2 provided by the 3H/ZnS∶Cu tritium source, the PM6∶Y6 device achieves a power conversion efficiency of 16.03%. More notably, the D18∶L8-BO device, benefiting from superior optical absorption capability and enhanced spectral matching with the dominant radioluminescent emission peak, delivers an output power density of 1.32×10−4 mW/cm2 and achieves a conversion efficiency of 16.51%. Based on this organic photovoltaic system, the overall efficiency of the constructed radiophotovoltaic isotope battery reaches 1.52%. These findings clearly indicate that precise alignment between the absorption spectrum of the organic active layer and the emission profile of radioluminescent sources is a critical factor for improving photoelectric conversion efficiency under ultra-weak light conditions. In addition to spectral compatibility, organic semiconductors exhibit stable built-in electric field utilization under low photon flux, along with inherent advantages including low material cost, mechanical flexibility, and suitability for scalable and miniaturized fabrication. Collectively, these attributes render organic photovoltaic materials highly competitive alternatives to conventional inorganic systems for radiophotovoltaic isotope batteries. As experimentally validated in this study, the D18∶L8-BO system emerges as one of the most promising organic active layer candidates for tritium-driven radiophotovoltaic applications. Overall, the incorporation of wide-bandgap organic photovoltaic materials provides a viable and effective technological route for powering MEMS, self-sustained microsystems, and flexible energy devices, offering critical material solutions and design strategies for overcoming the spectral and structural limitations inherent to traditional inorganic radiovoltaic technologies.
In order to understand the influence of cold wall on characteristics of subcooled flow boiling in petal-shaped fuel assemblies, a numerical investigation was performed to analyze subcooled flow boiling characteristics in two petal-shaped fuel assembly configurations (with and without guide tubes) by applying the two-fluid model with the RPI wall boiling model. The simulation data of these two geometric models were compared, and the influence of guide tubes on flow and heat transfer characteristics, such as flow field, heat transfer capability, onset location of fully developed nucleate boiling (FDB), and void fraction, all of them were systematically investigated. The results indicate that the transverse flow velocity is primarily concentrated around the fuel rods, forming swirling flows in the assembly with guide tube, no significant transverse flow is observed, cross-flow velocity is zero near the guide tube. Transverse flow similarly concentrates around the fuel rods, but both central and peripheral rods exhibit pronounced swirling patterns in the assembly without guide tube, with peak transverse velocity reaching 0.15 m/s. The guide tubes reduce transverse flow velocity, leading to more uniform lateral velocity distribution around fuel rods. The coolant velocity in corner channels is higher in the assembly with guide tube, exceeding that in corresponding channels of the assembly without guide tube by approximately 30%. In the assembly channel without guide tube, fluid velocity in the central channel increases along the axial direction due to interference from adjacent fuel rods, and the rate of increase gradually diminishes, while velocity with periodic fluctuations decline synchronized to the helical twisting pattern of petal-shaped fuel rods in the central channel of assembly with guide tube. The friction coefficient of the entire fuel assembly channel increases with guide tube, rising by approximately 100% within the simulated conditions compared to assembly without guide tube. In the entrance region of the assembly with guide tubes, heat transfer coefficients are higher. With the flowing, the heat transfer coefficients of both configurations gradually stabilize. The average wall temperature increases gradually along the axial direction in both assemblies (with and without guide tube). At the same axial position, the wall temperature in the non-guide-tube assembly is consistently higher, exceeding that of the assembly with guide-tube by approximately 4-5 K. In the concave regions of the fuel assembly, the wall temperature of fuel rods in the assembly without guide tubes are higher than that in the assembly with guide tubes, with a maximum temperature difference approaching 10 K. No notable difference is observed in the convex regions. The void fraction distribution shows higher in concave zones than in convex zones, with minimal impact from guide tubes. The onset location of FDB is only slightly affected by the guide tube.
This paper addresses the availability assessment of key components of pressure transmitters under severe accident conditions in nuclear power plants, with a particular focus on the irradiation damage characteristics of the bipolar operational amplifier LM108 when subjected to coupled high-temperature and high dose rate environments representative of beyond-design-basis events. Based on the NB/T 20149—2012 standard and the severe accident environmental spectrum established for the HPR1000 reactor design, irradiation tests were carefully designed to simulate the transient and harsh conditions that instrumentation may experience during severe accident scenarios. Using a 60Co γ-ray source, LM108 samples were irradiated at three elevated temperatures (50, 100, and 150 ℃) and at a high dose rate of 3.3 kGy/h until a cumulative absorbed dose of 36 kGy was reached. Emphasis was placed on monitoring the degradation behavior of the input bias current (Ib), the device parameter identified as most sensitive to ionizing dose in preliminary characterizations, and on recording time-resolved parameter evolution for detailed mechanistic analysis. The experimental results reveal a pronounced and non-monotonic suppression effect of elevated temperature on LM108 irradiation damage. At 50 ℃, taken as the reference operating temperature, Ib increases monotonically with cumulative dose, demonstrating classical damage accumulation due to trapped oxide charge and interface states induced by ionizing radiation. In stark contrast, at 100 ℃ and 150 ℃, Ib exhibits a marked increase at low-to-intermediate doses followed by a decline as the cumulative dose continued to rise. Particularly at 150 ℃, Ib reaches a distinct damage peak at intermediate doses and then partially recovers toward its initial, pre-irradiation value at higher accumulated doses. These behaviors indicate that during the early phase of severe accident conditions the high dose rate accelerates the generation of radiation-induced defects, whereas at elevated temperatures thermally activated annealing processes, especially the annealing of oxide-trapped charge density (Not) and recombination of some interface defects, become progressively dominant and effectively counteract further ionizing damage. To quantitatively capture this interplay, a dynamic competition model between trap charge capture and thermally activated annealing was developed. The model incorporates generation terms proportional to dose rate and trapping efficiency, as well as annealing terms governed by Arrhenius-type temperature dependence. Key model parameters were extracted by fitting the measured Ib-versus-dose curves across the three temperature conditions. The resulting model not only reproduces the observed non-monotonic trends and the peak-and-recovery behavior at high-temperature but also predicts continued recovery in the late stage of the severe accident scenario. These findings provide a mechanistic explanation for temperature-dependent irradiation responses and supply a theoretical basis for survivability assessment and reliability prediction of instrumentation and control components in nuclear power plants subjected to beyond-design-basis accident environments.
