The purpose of this study is to develop a finite element model that accurately describes the buckling behavior of a spacer grid. The spacer grid is the most important component of a nuclear fuel assembly and supports the fuel rod with a structurally sufficient buckling strength. Therefore, the development of a reliable spacer grid model is essential to evaluate the mechanical integrity of a nuclear fuel assembly. To achieve this objective, a three-dimensional finite element model was proposed to simulate the buckling characteristics and mechanical behavior of a PWR spacer grid. To simulate the exact mechanical properties of the spacer grid cell, the parameter values required for the model were determined by conducting a fuel rod drag test and spacer grid spring/dimple stiffness test. Finally, a spacer grid static compression test and dynamic impact test were performed according to the gap size of the spacer grid cell, and the model was verified by comparing the test and analysis results. The results obtained using the developed spacer grid finite element model agreed well with the mechanical test results, and it was confirmed that both the buckling characteristics and mechanical behaviors of the model were accurately simulated by the proposed model.

The fuel rod supporting structure of spacer grid is the key to ensure the horizontal and axial position of the fuel rod. On the purpose of improving the neutron economy and simplifying the manufacturing process of the spacer grid, the innovative curved dimple and arched spring have been designed, which could be directly punched from the strip of spacer grid. The mechanical experiments have been performed to acquire the deformation-load curves of dimples and springs on single strap and in-cell supporting structure, which could provide the load and residual deformation at 100% and 120% designed nominal deformation. It has been demonstrated that the designed in-grid cell has a relative stable load at the 100% nominal designed deformation and the test method of the single stripe is representative for the similar load-deflection curve between the spring and the in-grid cell.

Reactor pressure vessel (RPV) operates under high temperatures and pressures and is exposed to relatively high neutron radiation. RPV is considered to be irreplaceable, which is the most limiting factor for the lifetime of a nuclear power plant. As the most severe ageing degradation mechanism in RPV materials, irradiation embrittlement is a major issue affecting the integrity through the service life of a RPV. Our previous paper (ASME PVP2019-93615[1]) introduced our project for assessment of irradiation embrittlement of the materials for the Chinese RPVs to verify the 60-year design life, in which the specimens made of the RPV base material manufactured in China, the SA-508 Gr.3 Cl.1 forging, and the different types of weld metals were irradiated in the high fluence engineering test reactor (HFETR). The paper analyzed extent irradiation damage of the forging in terms of mechanical properties. As another part of the project, this paper concentrates on the evaluation of the weld metals in the same project. Tensile tests, Charpy impact tests and fracture toughness tests by master curve approach were carried out for the three types of weld metals subjected to different irradiation fluences (2.6E19n/cm 2 , 8.9E19n/cm 2 ). Comparison of the mechanical properties of the irradiated and the unirradiated materials is made. The irradiation resistance of the weld metals in our project is also compared with the data in the literatures.

A probability of rupture for WWER-1000 main piping was calculated based on the Failure Assessment Diagram (FAD), treating material properties of welds (the most likely zone for crack growing and its nucleation) and crack morphology parameters as stochastic values. In order to perform probabilistic calculations, Critical Temperature of Brittleness (CTB, WWER’s analogue of PWRs transition temperature which used to index the Fracture Toughness curve) and Yield (Ultimate) Strength as well were fitted by normal distribution, based on experimental data taken from the manufacture documentation of Ukrainian Nuclear Power Plants (NPPs). A set of calculations were conducted for Normal Operating conditions (NOC), Safe Shutdown Earthquake (SSE) and several emergency situations like: shaft jamming of a reactor cooling pump and break of the piping’s, connected to the considered ones (a set of LOCA events). Based on static and dynamic calculations, the most loaded zones were selected, where the cracks were postulated. Crack opening area was calculated according to original developed procedure, which accounts for membrane and linear stress components through the wall thickness. The Henry-Fauske flow model is used with modified parameters accounting for crack morphology as a normally distributed random variable. It is an important part of analysis, since different crack types have great differences in friction, bend protrusion and flow length parameters. The rupture probabilities for Main Circulating Piping were calculated with accounting for thermal aging effect. It was proven, that crack morphology parameters highly affect the leak rate and its distribution becomes more scattered. Among the mechanical characteristic, a Fracture toughness has more influence rather than Ultimate of Yield strength.

ThO 2 has been considered as a possible replacement for UO 2 fuel for future generation of nuclear reactors, and thorium-based mixed oxide (Th-MOX) fuel performance in a light water reactor was investigated due to better neutronics properties and proliferation resistance compared to conventional UO 2 fuel. In this study, the thermal, mechanical properties of Th 0.923 U 0.077 O 2 and Th 0.923 Pu 0.077 O 2 fuel were reviewed with updated properties and compared with UO 2 fuel, and the corresponding fuel performance in a light water reactor under normal operation conditions were also analyzed and compared by using CAMPUS code. The Th 0.923 U 0.077 O 2 fuel were found to decrease the fuel centerline temperature, while Th 0.923 Pu 0.077 O 2 fuel was found to have a bit higher fuel centerline temperature than UO 2 fuel at the beginning of fuel burnup, and then much lower fuel centerline than UO 2 fuel at high fuel burnup. The Th 0.923 U 0.077 O 2 fuel was found to have lowest fuel centerline temperature, fission gas release and plenum pressure. While the Th 0.923 Pu 0.077 O 2 fuel was found to have earliest gap closure time with much less fission gas release and much lower plenum pressure compared to UO 2 fuel. So the fuel performance could be expected to be improved by applying Th 0.923 U 0.077 O 2 and Th 0.923 Pu 0.077 O 2 fuel.

The spacer grid is a key element of the fuel assembly used in the Pressurized water reactor. Due to its structural complexity, the analysis and the design of the spacer grid structure is difficult. This paper discusses the 5 × 5 cell size partial grid analysis including the detailed grid structural elements, through which, the impact force, the rebound velocity and the time history of acceleration and as well as other mechanical properties of grid under different initial impact velocity were obtained. This paper carried out the dynamic buckling criterion studies, and determined the dynamic buckling load of the 5 × 5 cell size partial spacer grid. Based on assuming the impact process is simple harmonic vibration, the method to determine the dynamic stiffness of the spacer grid was proposed. The experiments were also performed for the comparison with the analytical results. It is found that the analytical results are in good agreement with the experimental results. As a result, we can conclude that the analysis model including detailed grid elements is able to yield accurate analytical results.

Additive Manufacturing (AM) can fabricate 3D complex functional parts, which can reduce material waste and increase manufacturing efficiency significantly. These benefits make AM technique suitable for some critical industry applications. Confident application of the AM technique requires whole understanding of AM parts’ properties. Safety and economics are essential to nuclear power plant. In this study, an innovative 316L stainless steel spent fuel storage rack with integrative structure was designed, and a small model of this rack was fabricated by selective laser melting (SLM), mechanical properties of the 316L stainless steel manufactured by SLM technique are studied and discussed. Key technical issues of application of AM to manufacturing nuclear parts are also discussed.

Impinging jet cooling technique has been widely used extensively in various industrial processes, namely, cooling and drying of films and papers, processing of metals and glasses, cooling of gas turbine blades and most recently cooling of various components of electronic devices. Due to high heat removal rate the jet impingement cooling of the hot surfaces is being used in nuclear industries. During the loss of coolant accidents (LOCA) in nuclear power plant, an emergency core cooling system (ECCS) cool the cluster of clad tubes using consisting of fuel rods. Controlled cooling, as an important procedure of thermal-mechanical control processing technology, is helpful to improve the microstructure and mechanical properties of steel. In industries for heat transfer efficiency and homogeneous cooling performance which usually requires a jet impingement with improved heat transfer capacity and controllability. It provides better cooling in comparison to air. Rapid quenching by water jet, sometimes, may lead to formation of cracks and poor ductility to the quenched surface. Spray and mist jet impingement offers an alternative method to uncontrolled rapid cooling, particularly in steel and electronics industries. Mist jet impingement cooling of downward facing hot surface has not been extensively studied in the literature. The present experimental study analyzes the heat transfer characteristics a 0.15mm thick hot horizontal stainless steel (SS-304) foil using Internal mixing full cone (spray angle 20 deg) mist nozzle from the bottom side. Experiments have been performed for the varied range of water pressure (0.7–4.0 bar) and air pressure (0.4–5.8 bar). The effect of water and air inlet pressures, on the surface heat flux has been examined in this study. The maximum surface heat flux is achieved at stagnation point and is not affected by the change in nozzle to plate distance, Air and Water flow rates.

The storage tanks in nuclear facilities has a significant impact on the safety of the reactor and the radiation shielding, so its mechanical property analysis has been widely concerned in the field of engineering and scientific research. Meanwhile, the storage tank is usually filled with gas and liquid medium. In the presence of external disturbances (such as external force, displacement, earthquake etc.), the position and structure of the vessel changes, that lead to changing of the gas-liquid interface. This characteristic can make the storage tank system as a tightly fluid-structure coupling system. In this paper, a storage tank which stored radioactive gas liquid medium is choosing to study such fluid-structure coupling system phenomenon, and a typical dynamic seismic condition is assumed. A two-way fluid-structure coupling method is used with CFD (Computational Fluid Dynamics) and FEM (Finite Element Method) numerical method. The study considered interaction between structure and two phase turbulent fluid. In FEM calculation, the time history seismic acceleration load is applied to the support of tank, and the flow loading coming from fluid medium is applied to the wall of tank which is send from CFD code. Then, the structure displacement which is calculate by FEM is transferred to CFD code. In CFD calculation, multiphase fluid numerical model is applied to simulate the flow characteristics of gas-water two phase fluid, and the turbulent properties are also considered in the calculation. Mesh deformation method is used to simulate the displacement of flow passage boundary which is send by FEM code. After CFD calculation, flow loading is transferred to the tank wall of FEM code again. Such loop of FEM and CFD calculation continues to go on with the seismic time history, the response characteristics of the tank will be solved. In order to evaluate the difference between the above method and the traditional analysis method. An independent calculation used added mass approach is carrying out, in which the effect of steady state water is applied to the wall of the vessel, and this load will not change with the earthquake. All others load and constraint mode are same with the above method. According to the two-way fluid-structure coupling analysis, the detailed characteristics of liquid free surface distribution and structural response of the vessel are obtained. The results show that the response vibration amplitude of the tank structure increases with the earthquake, and the response is mainly affected by the liquid sloshing. According to comparative analysis, the advantages of coupling method are proved. The method from this study can be used for the same type of analysis.

Due to its good creep strength, molybdenum (Mo) single crystal is used for space thermionic reactor fuel element. However, stray grain formation during Mo single crystal welding leads to the degradation of mechanical properties. In this paper, Mo single crystal is welded by electron beam welding and the joint properties are evaluated by means of small punch test (SPT). The load-displacement curves of Mo single crystal matrix and fusion area are measured. The experimental data is fit to establish the relationship between yield load and standard test yield strength. The results show that the yield strength of the fusion is much lower than the other material. The micro hardness of the both materials is almost equal. Simultaneously the displacement value of the fusion area is the smallest which means the joint brittleness increases. SEM observation of rupture fracture demonstrates the embrittlement of the weld.

Control rods with low worth absorber materials may provide a mechanical means of making relatively small adjustments in core reactivity. Mo-Tb-Dy and Y-Tb-Dy alloys were developed to obtain appropriate nuclear performance for low worth absorber material. The two alloys were prepared by powder metallurgy technology and vacuum melting technology individually. To clarify the effects of Mo and Y diluents, Tb-Dy was also prepared to be compared. The microstructures were analyzed by X-ray diffraction (XRD), Scanning Electron Microscope (SEM) and Transmission Electron Microscopy (TEM). The experiment results showed that homogeneous microstructures were obtained. Out-pile properties, including mechanical properties, thermal conductivities, thermal expansion, corrosion resistance properties and ion irradiation properties were measured and analyzed. Y-Tb-Dy has similar properties with Tb-Dy. With temperature increasing, yield strengths of Tb-Dy and Y-Tb-Dy decreases largely while Mo-Tb-Dy decreases slightly. Thermal conductivities of Mo-Tb-Dy were four times more than Tb-Dy and Y-Tb-Dy. Mo element significantly increases thermal conductivity. Tb-Dy and Y-Tb-Dy showed severe corrosion and became powders in 280°C/10MPa de-ionized water while Mo-Tb-Dy had very slow corrosion rate. All three alloys were irradiated at 400∼700°C for 25 displacement per atom (dpa). No voids was observed for Tb-Dy and Y-Tb-Dy. Void diameter increases and its density decreases with temperature increasing for Mo-Tb-Dy. Maximum irradiation swelling rate with 0.5% was observed at 500°C. Irradiation swelling significantly decreased with increasing irradiation temperature.

Reactor containment of a nuclear power plant is a structure to ensure the safety of nuclear power plant. It acts as the last barrier to prevent the release of radioactive materials from NPP during accidents. Finite element models were established to simulate a 1/3 scale model of a reactor containment building under leakage test pressure. General finite element software ANSYS were applied. The nonlinear behavior of containment materials, geometric were taken into account in the analysis. The reliability of the finite element model was verified through the comparison of theoretical analysis results with experimental results. In the ANSYS finite element model, the concrete, steel bars and prestress tendons were separated and the prestress tendons were considered by the method of cooling method on the prestress tendon elements. The mechanical properties of the finite element model in the prestress tension process and the absolute internal pressure of 0.52MPa were analyzed. Transient and time dependent losses were taken into account at the same time during the calculation of prestress of tendons, so as to calculate effective prestress at different locations of tendons. Calculation results of prestress losses show that the prestress losses at the hole of equipment hatch are larger than the other areas. The results show that, the deformation of over-all structure of the containment is shrink inward under the action of prestress. And the simulation can achieve the consistent deformation effect between tendons and concrete. The maximum radial displacement of the whole containment structure is located at of 10 ° ∼ 20 °area on the right of the hole of the gate. The effect of expansion deformation of the containment caused by design internal pressure is insufficient to offset the inward shrink effect generated by tendons, and the over-all structure of the concrete containment scale model is mainly under compressive stress. The containment test model is still with a large safety margin under the action of design internal pressure. The largest tensile stress is on the up and down areas of the internal sides of the equipment hatch, dome area close to ring beam, and bottom of perimeter wall close to the base slab. There is possibility of cracking on the concrete in limited local zones. This benchmark can provide a reference for engineering design of containment.

The anti-seismic performance of fuel assembly is mainly determined by the critical crush load and the stiffness of spacer grids. To comprehensive know about the influence of fuel rods on the spacer grid, a 5×5 spacer grid FEM model which including fuel rods is established. Basing the fact that the grid spring has remarkable influence on the grid crush strength which is found in experiment, some cases are carried out, which are used to analyze effects of grid with/without fuel rod, friction between the grid spring/dimple and the fuel rod, the deflection of grid spring on the static buckling strength. Results show that grids with fuel rods will have higher crush strength than those without fuel rods; at certain range, increasing grid spring deflection at working point will do help to increase the grid crush strength; higher friction coefficient of grid spring and fuel rod can enhance the crush strength. Comparing with experimental results in literatures, results from simulations show the same tendency with the experimental results. The conclusion and the simulation method involved in this paper can provide some guidelines to optimize the performance of spacer grid assembly.

The effect of neutron irradiation damage of reactor pressure vessel (RPV) steels is a main failure mode. Accelerated neutron irradiation experiments at 292 °C were conducted on RPV steels, followed by testing of the mechanical, electrical and magnetic properties for both the unirradiated and irradiated steels in a hot laboratory. The results showed that a significant increase in the strength, an obvious decrease in toughness, a corresponding increase in resistivity, and the clockwise turn of the hysteresis loops, resulting in a slight decrease in saturation magnetization when the RPV steel irradiation damage reached 0.0409 dpa; at the same time, the variation rate of the resistivity between the irradiated and unirradiated RPV steels shows good agreement with the variation rates of the mechanical properties parameters, such as nano-indentation hardness, ultimate tensile strength, yield strength at 0.2% offset, upper shelf energy and reference nil ductility transition temperature. Thus, as a complement to destructive mechanical testing, the resistivity variation can be used as a potentially non-destructive evaluation technique for the monitoring of the RPV steel irradiation damage of operational nuclear power plants.

Two Zr-Sn-Nb alloys with minor Germanium or silicon additions were prepared by traditional manufacturing process to meet the design requirements. Transmission electron microscope (TEM) and scanning electron microscope (SEM) were utilized to characterize the detail microstructure of base alloys. Corrosion resistance was examined by the weight gain in static autoclave with different water chemistry environments. The mechanical properties at room temperature and elevated temperature were evaluated by conventional tensile testing. Thermal creep resistance was evaluated by an internally-pressurized creep test at 385 °C with hoop stresses of about 108 MPa and 150 MPa (during 24 h). It was found that SZA-6 and SZA-4 alloys consisted of partially recrystallized grain structures with uniformly distributed fine second phase particles (SPPs) located within grain interior and at grain boundaries. Both SZA-4 and SZA-6 alloys exhibited excellent corrosion resistance in two water chemistry conditions. The corrosion resistance of SZA-6 was better than the reference commercial alloy, and SZA-4 was slightly better than SZA-6. The mechanical properties of two new zirconium alloys were comparable, and both of them can meet the design criterion. Moreover, the thermal creep resistance of SZA-4 and SZA-6 alloys was equivalent to existing commercial alloy. Considering the outstanding corrosion resistance, satisfied mechanical properties and thermal creep resistance, SZA-4 and SZA-6 alloys were suggested as promising alloys used for CAP1400 fuel assembly in the future.

Nozzles of fuel assembly play an important role in pressure water reactor (PWR) fuel assembly element. For a long time, ordinary processing technologies of nozzles of fuel assembly have the problems of difficult and complicated process, the low availability of material and long the development cycles of manufacturing. However, according to the study these issues can be well settled by using the additive manufacturing technology. This paper studies a nozzle of fuel assembly prepared by this additive manufacturing technology through slow-strain-rate tension (SSRT) test and microstructure observation experiment. The results of SSRT test show that yield strength of the nozzle of fuel assembly is about 401.5MPa, the extensional rigidity is about 673.5MPa and the ductility is about 45.7%. And the SEM fracture results of the SSRT sample indicate that the fracture microstructure contains a large number of dimples, and the way of fracture belongs to plastic. And the metallographic observation consequences manifest that the microstructure of nozzle of fuel assembly prepared by the additive manufacturing technology is composite tissue of both austenite and ferrite, and the grains are settled along the way of laser scanning and there are isometric with some kind of direction. This metallographic microstructure is different from the traditional morphology of the free carbide distributed in the matrix. The dual phase microstructure of austenite and ferrite can improve the mechanical properties of the matrix effectively, and avoid the free carbides which may lead to matrix fragmentation in the tensile deformation process. Moreover, the laser power could affect the microstructure and properties of nozzles of fuel assembly observably, and the high laser power could bring about the ablation of metal. Through the analysis of mechanical properties and microstructure, we have made it possible to make the laser additive manufacturing technology to be used for the fuel assembly nozzle preparation in the nuclear power area. This work not only presents the advantages of the laser additive manufacturing technology in the fuel element processing area of the nuclear power station, but also broadens the application range of the laser additive manufacturing technology. What’s more we provide the new thoughts for the fast and effective preparation of the fuel element especially for the fuel assembly nozzle in the nuclear power station.

A novel type of flexible composite with tungsten powder filled in polymers (TPFP) homogeneously was developed as an alternative to lead (Pb) based radiation shields used in nuclear industries. TPFP had a density in the range of 4 to 11.3 g/cm 3 , which can be tailor-made according to the applications. In addition to the advantage of lower toxicity over Pb-based shielding, TPFP can be formed into various shapes, such as pipe shields, pipe wraps, safe floor shields, blankets, etc. The mechanical properties and attenuation of γ-ray was investigated for the developed TPFP.

Up to now, two kinds of filler metal with or without nickel element for submerged arc welding have been largely used in the reactor pressure vessel (RPV) manufacturing. In order to study the effect of nickel element on weld metal properties of SA-508 Gr.3 Cl.1, submerged arc welding material with nickel (AWS classification F8P4-EGN-F2N, F2 for short) and welding material without nickel (F8P4-EA3N-A3N, A3 for short) were used; and conventional mechanical properties, low-cycle fatigue test, and proton irradiation analysis of the two weld metals were studied. Results show that the mechanical properties of the two different weld metals are similar, except that the Charpy V-notch impact property of the weld metal with nickel is better than that without nickel; the micro-structures of F2 and A3 weld metals are both composed of ferrite base and granular bainite, but the columnar grain size of F2 weld metal is smaller relatively, which results in better impact property. In addition, the irradiated A3 weld metal has fewer dislocation loops than the irradiated F2 weld metal after the same proton irradiation dose; the irradiated weld metals both have higher micro-Vickers hardness than before.

Additive Manufacturing (AM) can fabricate 3D near-net-shaped functional parts using unit materials, such as powder or wire. Additive manufacturing’s computer-aided design offers superior geometrical flexibility. The near-net shaping capability also reduces materials waste and increase manufacturing efficiency significantly. These benefits make AM desirable for critical industry applications, such as art, aerospace, ground transportation, and medical. Confident utilization of the technology requires thorough understanding of the AM materials, ensuring both structural integrity and performance requirements are met or exceeded. Safety and economics are essential to nuclear power plant. In this study, mechanical properties of a ferritic steel fabricated by electric melting additive manufacturing (EMAM) technique are studied and compared with ASME SA-336 Gr.F12, which applied to nuclear main steam line penetration, the results are systematically presented and discussed. Key technical issues of application of AM to manufacturing nuclear components are also discussed.

The mechanical analysis of weld zone is a significant problem during the nuclear power plant design due to the fact that the material property of weld zone is complex. In most of the standards and codes, the weld zone is assumed as an effective homogeneous material. However, the weld zone is non-homogeneous at micro-scale, and the microstructure has great effect on the mechanical property, especially the fracture behaviors. Therefore, how to propose an accurate and convenient model to analyze fracture behavior of the weld zone of nuclear pipe or components becomes a major issue. The objective of present work is to develop an accurate method to investigate the fracture behavior of weld zone from the microstructure of weld zone. To achieve the objective, a double smoothed image based reconstruction method is developed to generate the finite element mesh from digital image of weld zone directly. Then a cohesive finite element framework is utilized to simulate the initiation and propagation of micro-cracks in weld zone. Moreover, a damping boundary condition is used to apply the complex loading case on the micro-scale model.