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The formation and dynamic process analysis of the rockslide avalanche in mountainous areas are one of the consequences of the catastrophic accident. Such loose accumulation in upslope may saturate partially or completely when the stability of their accumulation dam is distributed or affected by rainfall. We present a case study with respect to the southern Wuyi Mountain located in eastern China, where the wet and rainy climate has led to dozens of similar rockslide hazards. The purpose of this paper was to analyse the mechanism and dynamic characteristics of the rockslide influenced by the same geological conditions and to predict the outburst susceptibility of similar landslides in the future. Detail field surveys, 3D laser scanning, and high-density electrical methods were used to collect the geotechnical information of the complex landslide, to identify the discontinuity between the landslide material and the bedrock, and to investigate the deformation characterization and dynamic process of the rockslide. Based on remote sensing interpretation and field investigation of the deformation process of a landslide in different times and different parts, the background, mechanism, and cause of the landslide were demonstrated. The landslide is controlled by the characteristics of the geological structure, including collapse, circular sliding, plane sliding, and debris flow. In addition, there are rock avalanches on the rear edge of the slope subjected to the combined action of rainfall and gravity. Moreover, there are some resistance anomaly areas of the aquifer and soil between 2 and 50 m where the resistivity is less than 120 ohm-m, and they were deduced to be full of water, confirming a "bathtub" type structure. The mechanism of the catastrophic landslide was a combination of the upper pushing deformation induced by rainfall line uplift and rotational; due to the ancient landslide reactivation in the transposition area, the velocity of the rockslide reached 40.11 m/s. [ABSTRACT FROM AUTHOR]

• The DRWM technique with the pin-by-pin response function has been proposed. • The three-dimensional pin-by-pin response function of the ex-core detector is obtained by the adjoint method with the Hydra code. • The pin-power reconstruction based on source-expansion method has been adopted to provide the pin-by-pin power distribution under steady and transient calculation. The dynamic rod worth measurement (DRWM) technique is widely applied in PWR engineering. In DRWM, the response function of the ex-core detector is very significant to calculate the spatial-correction factors which are introduced to eliminate the spatial effects due to the rapid movement of control rods. Theoretically, the peripheral pins of the peripheral fuel assemblies close to the ex-core detector have large contributions to the ex-core detector signal. However, the conventional DRWM technique only uses the nodal response function, which can't consider the pin-level contribution. Therefore, the DRWM technique based on the pin-by-pin response function has been proposed in this study. This proposed technique has been applied and verified with the measurement data of commercial PWRs operated in China. The engineering verification showed that this technique could provide accurate spatial-correction factors, and the measured control-rod worth satisfies the engineering acceptance criteria. [ABSTRACT FROM AUTHOR]

• Both full-nuclide and 29 heavy-nuclide microscopic-depletion capabilities were implemented in Bamboo-C. • The three-dimensional response function of ex-core detector has been calculated by the Hydra code. • The effects of the nuclide-number-density distribution on response function and DRWM results were quantified. The dynamic rod worth measurement (DRWM) has been widely applied in PWRs to measure the worth of RCCA (Rod Cluster Control Assembly) using the electricity signal of ex-core detector. In DRWM, the response function of ex-core detector is prerequisite for determining the spatial-correction factors of DRWM, which are vital parameters to correct the spatial effects due to the quick movement of RCCA during the measurement process. Theoretically, the response function of ex-core detector characterizes the contributions of the in-core neutron-flux distribution on the response of ex-core detector, which is dependent on the in-core nuclide-number-density distribution. Therefore, the researches have been implemented to quantify the effects of the multiple-cycle nuclide-number-density distribution on the response function of ex-core detector and further on the DRWM results. The previous microscopic-depletion model in our home-developed Bamboo-C code could only consider 29 main heavy-metal nuclides and 4 main fission-poison nuclides. In this research, the microscopic-depletion capability with full nuclides (233 nuclides) has been implemented, with which the detailed multiple-cycle nuclide-number-density distribution could be provided. For the method verification, two continuous refueling cycles of CNP1000 operated in China have been modeled, by which the effects of the multiple-cycle nuclide-number-density distribution on the response function of ex-core detector, then on the spatial-correction factors of DRWM and finally on the worth of RCCA have been quantified. Through the numerical results, it could be observed that the largest deviation of response function occurs at the fuel assembly near the ex-core detector, which is less than 3.0e-6 in absolute value. The relative deviations of spatial-correction factors are less than 1.5%, and the relative deviations of RCCA's worth are less than 0.2%. Therefore, the 29 heavy-nuclide model is accurate sufficiently for the application of DRWM technique. [ABSTRACT FROM AUTHOR]

• A data-assimilation method has been proposed for the burnup distribution. • The three-dimensional variational algorithm and artificial neutral network algorithm were applied. • The data-assimilation method has been applied to PWR operated in China. In this paper, a data-assimilation method has been proposed and applied for the burnup distribution of PWR. The burnup distribution is significant for the safety and economy of the reactor, as it is essential for the fuel-reloading design and optimization. Due to the fact that the burnup distribution cannot be measured directly during the reactor operation, the numerical simulation is widely applied in reactor engineering. However, there definitely exist differences between the numerical simulation and the actual core due to some unavoidable factors, such as component manufacturing deviation, uneven flow distribution and so on. These differences would induce the errors to the simulation values of power distributions and hence to the burnup distributions. Therefore, a data-assimilation method for the burnup distribution has been proposed, aiming at reducing the burnup-distribution errors with application of the power-distribution measurements. In our research, the three-dimensional variational (3DVAR) algorithm was applied for burnup-distribution calibration and the artificial neutral network (ANN) algorithm was applied to establish the complex relations between burnup distribution and power distribution. As engineering verification, the proposed data-assimilation method has been applied to the CNP1000 PWR operated in China. The numerical results indicated that the burnup-distribution errors can be reduced notably, as the maximum value of relative errors for power distribution can be reduced from 9.53% to 5.11%. [ABSTRACT FROM AUTHOR]

• An improved homogenization method was proposed for heavy reflectors of VVER. • The resonance effect of the main isotopes in heavy reflector has been evaluated. • The improved homogenization method has been applied to VVER operated in China. Compared with the traditional Cartesian PWR design, the heavy reflector, composed of larger amount of iron and less moderator, is an important feature of the VVER reactor. This structural feature would result in two different neutronics characteristics: first, less moderator would harden the neutron spectrum and make more adjoint reflector-node correlated; second, larger amount of iron within harden neutron spectrum may lead to notable resonance phenomenon for the structure–material isotopes. Therefore, the above neutronics characteristic would introduce corresponding challenges to the homogenization method for the heavy reflectors. In order to provide more accurate few-group homogenized parameters, an improved homogenization method has been proposed. As method verification, a two-dimensional VVER benchmark has been simulated, with the reference values provided by the Monte-Carlo code. Moreover, the improved homogenization method has been applied to the real VVER-1000 reactor operated in China, validating the simulation values of the key parameters with corresponding measurement values. [ABSTRACT FROM AUTHOR]

• A optimization method for PWR fuel-reloading pattern has been proposed. • The improved CNN was applied to evaluate fuel-reloading pattern efficiently. • The optimization method has been applied to CNP1000 PWR operated in China. In this paper, the optimization method of fuel-reloading pattern for PWR has been studied based on the improved convolutional neural network (CNN) and genetic algorithm (GA). It is very important to search out the optimized fuel-reloading pattern to guarantee the safety and economy of the nuclear power plants. During the optimization, large number of fuel-reloading patterns should be evaluated, providing the core parameters (including the cycle length, power-peak factors and so on) to the optimization algorithm to search for the optimized pattern. In our study, the CNN was improved with the advanced Inception-ResNet structure and applied to train the rapid-evaluation model, which can receive the fuel-reloading patterns and feedback corresponding core parameters with sufficient accuracy and very-high efficiency. The GA was applied as the optimization algorithm to search for the optimized fuel-reloading pattern. This proposed optimization method has been applied to the optimization of fuel-reloading pattern for the CNP1000-type PWR reactor operated in China. It can be observed that the CNN can evaluated the core parameters of one-single fuel-reloading pattern in about 0.0005 s and the averaged evaluation errors smaller than 0.6%; the GA can search the optimized fuel-reloading pattern in about 20 min. The study in this paper indicated that the combination of CNN and GA can provide the optimization of fuel-reloading pattern for PWR in very-short time, which can be applied to improve the safety and economy of the nuclear power plants. [ABSTRACT FROM AUTHOR]

• A new strategy for validating transient neutronic solvers has been proposed. • The ex-core detector signals and reactivity worth measured directly from the DRWM process were employed as the reference for validation. • The DRWM process was simulated by SPARK code based on the time-dependent neutron diffusion equation and the response function. • Simulated results achieved good agreement with measurements for both ex-core detector signals and reactivity worth. A new strategy was proposed in this study to validate the transient neutronic solver. This new strategy features in employing the measurement data of the ex-core detector signals and reactivity worth from the dynamic rod worth measurement (DRWM). Meanwhile, the control-rod movement during the DRWM is simulated explicitly by the transient neutronic solver based on the time-dependent neutron diffusion equation and the response function. By comparing the simulated results with measurement data, this strategy can strongly guarantee the accuracy and reliability of the transient neutronic solver in engineering applications. Additionally, different from the limited transient benchmarks and transient facilities, this new strategy has sufficient measurement data for transient validation benefiting from the widely applied DRWM technique in PWRs across the world. The application of such a new strategy has been demonstrated based on the CNP1000 reactor operated in China. The ex-core detector signals and reactivity worth with respect to each individual control rod were studied and provided. Good agreements have been achieved between measurement and simulation. This newly proposed strategy can be a feasible and reliable approach to validate the transient neutronic solver. [ABSTRACT FROM AUTHOR]

• A correction method is proposed to incorporate the resonance elastic scattering effect into the multi-group data used for the neutron transport calculation. • The proposed method is implemented in the lattice physics code LOCUST. • The impact of resonance elastic scattering on reactor-core physics simulations is quantified. NECP-Atlas and similar nuclear data processing codes have developed the function of incorporating the resonance elastic scattering effect into the multi-group cross sections and scattering matrices. However, the hyper-fine group method is commonly used in deterministic-based lattice physics codes and whole-core neutronics codes for resonance self-shielding calculations. In this method, the neutron slowing-down equation is established based on the conventional asymptotic scattering model where the neutron up-scattering caused by the resonance elastic scattering is ignored. Consequently, the deterministic codes cannot directly use the multi-group data produced by the nuclear data processing codes. In this paper, a correction factor method is proposed to correct the multi-group cross sections and scattering matrices obtained by the hyper-fine group calculations to consider the resonance elastic scattering. The proposed method has been implemented in the lattice physics code LOCUST. The few-group constants produced by the LOCUST code are provided to reactor-core analysis code SPARK to simulate a CNP-1000 type PWR reactor operated in China. Some core parameters measured during zero-power physics tests at 600 K and operations are analyzed to show the effect of the resonance elastic scattering on the core parameters. The numerical results show that the boron concentration reduces by 10 ∼ 14 ppm at 900 K during operations, which is closer to the measured value and improves the accuracy of reactor-core physics calculation. [ABSTRACT FROM AUTHOR]

• CENDL-3.2 and ENDF/B-VIII.0 have been used in the reactor physics simulations of commercial PWR. • Key parameters measured in the startup physics tests and operation in the first-three fuel cycles are calculated. • The results are compared with the previous version ENDF/B-VII.0. The latest CENDL and ENDF/B evaluated nuclear data libraries was released in 2020 and 2018, respectively. To apply CENDL-3.2 and ENDF/B-VIII.0 in the reactor physic simulations of pressurized water reactor (PWR), the CNP-1000 PWR, which is an improved GEN-II PWR and operated in Fuqing Nuclear Power Plant in China, is simulated using these two libraries. The key parameters during the startup physics tests and power operation in the first three fuel cycles of the CNP-1000 reactor have been simulated and compared with corresponding measurement values. The numerical results show that ENDF/B-VIII.0 performs better in several parameters of the startup physics tests than ENDF/B-VII.0; the cross-section data in CENDL-3.2 is competent in the engineering application of PWR. [ABSTRACT FROM AUTHOR]