ALBERT

Description: The high-temperature gas-cooled reactor pebble-bed module (HTR-PM) nuclear power plant consists of two nuclear steam supply system modules, each of which drives the steam turbine by the superheated steam flow and is fed by the heated-up water flow. The shared steam/water system induces mutual effects on normal operation conditions and transients of the nuclear power plant, which is worthy of safety concerns and intensive study. In this paper, a coupling code package was developed with the TINTE and vPower codes to understand how the HTR-PM operated. The TINTE code was used to analyze the reactor core and primary circuit, while the vPower code simulated the steam/water flow in the conventional island. Two TINTE models were built and coupled to one vPower model through the data exchange in the steam generator models. Using this code package, two typical transients were simulated by decreasing the primary flow rate or introducing the negative reactivity of one module. Important parameters, including the reactor power, the fuel temperature, and the reactor inlet and outlet helium temperatures of two modules, had been studied. The calculation results preliminarily proved that this code package can be further used to evaluate working performance of the HTR-PM.

Description: The paper presents a conceptual design of a 10 MW multipurpose nuclear research reactor (MPRR) loaded with the low-enriched uranium (LEU) VVR-KN fuel type. Neutronics and burnup calculations have been performed using the REBUS-MCNP6 linkage system code and the ENDF/B-VII.0 data library. The core consists of 36 fuel assemblies: 27 standard fuel assemblies and 9 control fuel assemblies with the uranium density of 2.8 gU/cm3 and the 235U enrichment of 19.75 wt.%. The cycle length of the core is 86 effective full-power days with the excess reactivity of 9600 and 1039 pcm at the beginning of cycle and the end of cycle, respectively. The highest power rate and the highest discharged burnup of fuel assembly are 393.49 kW and 56.74% loss of 235U, respectively. Thermal hydraulics analysis has also been conducted using the PLTEMP4.2 code for evaluating the safety parameters at a steady state of the hottest channel. The maximum temperatures of coolant and fuel cladding are 66.0°C and 83.0°C, respectively. This value is lower than the design limit of 98°C for cladding temperature. Thermal fluxes at the vertical irradiation channels and the horizontal beam ports have been evaluated. The maximum thermal fluxes of 2.5 × 1014 and 8.9 ×1013 n·cm−2·s−1 are found at the neutron trap and the beryllium reflector, respectively.

Description: Although many types of simulated radionuclides have been widely used as a substitute for actual nuclear waste in the studies of nuclear waste solidification, the understanding of the applicability and validity of simulated radionuclides is still insufficient. In particular, the selection and use of simulated radionuclides, which can play a decisive role in the accuracy of the experimental results, still lack unified or integrated references. This paper provides a critical review on the selection, experimental methods, and applicability of the most commonly studied simulated radionuclides, followed by a careful discussion and recommendation of simulated radionuclides suitable for different solidified bodies. The main factors (e.g., temperature, pH, and atmosphere) affecting the choice of simulated radionuclides were analyzed in detail. This work helps to integrate the selection and use of simulated radionuclides, and it will be beneficial for improving the effectiveness of nuclide solidification research.

Description: The thermal hydraulic and neutronics coupling analysis is an important part of the high-fidelity simulation for nuclear reactor core. In this paper, a thermal hydraulic and neutronics coupling method was proposed for the plate type fuel reactor core based on the Fluent and Monte Carlo code. The coupling interface module was developed using the User Defined Function (UDF) in Fluent. The three-dimensional thermal hydraulic model and reactor core physics model were established using Fluent and Monte Carlo code for a typical plate type fuel assembly, respectively. Then, the thermal hydraulic and neutronics coupling analysis was performed using the developed coupling code. The simulation results with coupling and noncoupling analysis methods were compared to demonstrate the feasibility of coupling code, and it shows that the accuracy of the proposed coupling method is higher than that of the traditional method. Finally, the fuel assembly blockage accident was studied based on the coupling code. Under the inlet 30% blocked conditions, the maximum coolant temperature would increase around 20°C, while the maximum fuel temperature rises about 30°C. The developed coupling method provides an effective way for the plate type fuel reactor core high-fidelity analysis.

Description: Electric valves have significant importance in industrial applications, especially in nuclear power plants. Keeping in view the quantity and criticality of valves in any plant, it is necessary to analyze the degradation of electric valves. However, it is difficult to inspect each valve in conventional maintenance. Keeping in view the quantity and criticality of valves in any plant, it is necessary to analyze the degradation of electric valves. Thus, there exists a genuine demand for remote sensing of a valve condition through nonintrusive methods as well as prediction of its remaining useful life (RUL). In this paper, typical aging modes have been summarized. The data for sensing valve conditions were gathered during aging experiments through acoustic emission sensors. During data processing, convolution kernel integrated with LSTM is utilized for feature extraction. Subsequently, LSTM which has an excellent ability in sequential analysis is used for predicting RUL. Experiments show that the proposed method could predict RUL more accurately compared to other typical machine learning and deep learning methods. This will further enhance maintenance efficiency of any plant.

Description: The geometries, adsorption energies, and electronic structures of Cs, Sr, and Ag atoms on matrix graphite surface with point defects were calculated and analyzed using the density functional theory (DFT) and the Perdew–Burke–Ernzerhof (PBE) formulation of the generalized gradient approximation (GGA). Three different types of point defects, i.e., single vacancy and “bridge” and “spiro” interstitials are considered using approximate van der Waals (vdW) correction methods. The results of adsorption energies show that the metal fission products of Cs, Sr, and Ag are more stable on single vacancy defects than “bridge” or “spiro” interstitial defects. This is further confirmed by the analysis of electronic structures, such as charge density difference (CDD) and density of state (DOS). All these results indicate that dangling bonds play an important role in the adsorption behaviors of metallic fission products on matrix graphite.

Description: Electrorefining is a key step in pyroprocessing. The solid cathode processing is necessary to separate the salt from the cathode of the electrorefiner since the uranium deposit in a solid cathode contains electrolyte salt. Moreover, it is very important to increase the throughput of the salt separation system due to the high uranium content of the spent nuclear fuel and high salt fraction of uranium dendrites. Therefore, in this study, the effect of deposit on the evaporation of the adhered salt in a uranium deposit was investigated by using the samples of salt in the uranium deposit and salt in the deposit of the surrogate material for the effective separation of the salt. It was found that the salt evaporation rate is dependent on the deposit type and bulk density in the crucible. Additionally, the evaporation rate was found to be lower when the deposit structure is complex; the rate also decreases as the bulk density of the deposit is increased owing to the retardation of the salt vapour transport process. It was concluded that the mass transfer of the salt vapour is an important parameter for the achievement of a high throughput performance in the salt distillation process.

Description: This paper focuses on fluid forces acting on a confined cylinder subjected to axial flow in application to fuel assembly dynamic behavior. From the literature, it is difficult to estimate the damping induced by the flow. Therefore, it is proposed to study numerically the damping fluid forces on a cylinder for various parameters. It has been observed that it increases with the smaller confinement and with the presence of an obstacle and decreases when the Reynolds number increases. Larger values correspond to a greater contribution of pressure forces. Dynamic simulations are compared to the steady ones and give different values, but the order of magnitude and general trend remain the same. Therefore, steady simulations are suitable to have a rough estimation of drag coefficients in dynamics.

Description: For nuclear reactor physics, uncertainties in the multigroup cross sections inevitably exist, and these uncertainties are considered as the most significant uncertainty source. Based on the home-developed 3D high-fidelity neutron transport code HNET, the perturbation theory was used to directly calculate the sensitivity coefficient of keff to the multigroup cross sections, and a reasonable relative covariance matrix with a specific energy group structure was generated directly from the evaluated covariance data by using the transforming method. Then, the “Sandwich Rule” was applied to quantify the uncertainty of keff. Based on these methods, a new SU module in HNET was developed to directly quantify the keff uncertainty with one-step deterministic transport methods. To verify the accuracy of the sensitivity and uncertainty analysis of HNET, an infinite-medium problem and the 2D pin-cell problem were used to perform SU analysis, and the numerical results demonstrate that acceptable accuracy of sensitivity and uncertainty analysis of the HNET are achievable. Finally, keff SU analysis of a 3D minicore was analyzed by using the HNET, and some important conclusions were also drawn from the numerical results.

Description: The new Structural Seismic Isolation System (SSIS) intends to provide high safety for important structures such as nuclear power plants, offshore oil platforms, and high-rise buildings against near-fault and long-period earthquakes. The presented SSIS structure foot base and foundation contact surfaces have been designed as any curved surfaces (spherical, elliptical, etc.) depending on the earthquake-soil-superstructure parameters, and these contact surfaces have been separated by using elastomeric (lead core rubber or laminated rubber bearings with up to 4-second period) seismic isolation devices. It would allow providing inverse pendulum behavior to the structure. As a result of this behavior, the natural period of the structure will possess greater intervals which are larger than the predominant period of the majority of the possible earthquakes including near-fault zones. Consequently, the structure can maintain its serviceability after the occurrence of strong and long-period earthquakes. This study has investigated the performance of the SSIS for the nuclear containment (SSIS-NC) structure. The finite element model of SISS-NC structure has been developed, and nonlinear dynamic analysis of the model has been conducted under the strong and long-period ground motions. The results have been presented in comparison with the conventional application method of the seismic base isolation devices for nuclear containment (CAMSBID-NC) and fixed base nuclear containment (FB-NC) structures. The base and top accelerations, effective stress, and critical shear stress responses of the SSIS-NC structure are 48.67%, 36.70%, and 32.60% on average lower than those of CAMSBID-NC structure, respectively. The result also confirms that the SSIS-NC structure did not cause resonant vibrations under long-period earthquakes. On the other hand, there is excessive deformation in the isolation layers of CAMSBID-NC structure.